National Library of Energy BETA

Sample records for test reactor area

  1. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  2. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  3. F Reactor Area Cleanup Complete

    Broader source: Energy.gov [DOE]

    RICHLAND, Wash. – U.S. Department of Energy (DOE) contractors have cleaned up the F Reactor Area, the first reactor area at the Hanford Site in southeastern Washington state to be fully remediated.

  4. Summary of the 1987 soil sampling effort at the Idaho National Engineering Laboratory Test Reactor Area Paint Shop Ditch

    SciTech Connect (OSTI)

    Wood, T.R.; Knight, J.L.; Hertzler, C.L.

    1989-08-01

    Sampling of the Test Reactor Area (TRA) Paint Shop Ditch at the Idaho National Engineering Laboratory was initiated in compliance with the Interim Agreement between the Department of Energy (DOE) and the Environmental Protection Agency (EPA). Sampling of the TRA Paint Shop Ditch was done as part of the Action Plan to achieve and maintain compliance with the Resource Conservation and Recovery Act (RCRA) and applicable regulations. It is the purpose of this document to provide a summary of the July 6, 1987 sampling activities that occurred in ditch west of Building TRA-662, which housed the TRA Paint Shop in 1987. This report will give a narrative description of the field activities, locations of collected samples, discuss the sampling procedures and the chemical analyses. Also included in the scope of this report is to bring together data and reports on the TRA Paint Shop Ditch for archival purposes. 6 refs., 10 figs., 8 tabs.

  5. Massive Hanford Test Reactor Removed - Plutonium Recycle Test...

    Energy Savers [EERE]

    challenge on the U.S. Department of Energy's (DOE) Hanford Site by removing a 1,082-ton nuclear test reactor from the 300 Area. The River Corridor is a 220-square-mile section of...

  6. PIA - Advanced Test Reactor National Scientific User Facility...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor...

  7. PIA - Advanced Test Reactor National Scientific User Facility...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test...

  8. Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presenter: Bentley Harwood, Advanced Test Reactor Nuclear Safety Engineer Battelle Energy Alliance Idaho National Laboratory

  9. Testing of Gas Reactor Fuel and Materials in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2006-10-01

    The recent growth in interest for high temperature gas reactors has resulted in an increased need for materials and fuel testing for this type of reactor. The Advanced Test Reactor (ATR), located at the US Department of Energy’s Idaho National Laboratory, has long been involved in testing gas reactor fuel and materials, and has facilities and capabilities to provide the right environment for gas reactor irradiation experiments. These capabilities include both passive sealed capsule experiments, and instrumented/actively controlled experiments. The instrumented/actively controlled experiments typically contain thermocouples and control the irradiation temperature, but on-line measurements and controls for pressure and gas environment have also been performed in past irradiations. The ATR has an existing automated gas temperature control system that can maintain temperature in an irradiation experiment within very tight bounds, and has developed an on-line fission product monitoring system that is especially well suited for testing gas reactor particle fuel. The ATR’s control system, which consists primarily of vertical cylinders used to rotate neutron poisons/reflectors toward or away from the reactor core, provides a constant vertical flux profile over the duration of each operating cycle. This constant chopped cosine shaped axial flux profile, with a relatively flat peak at the vertical centre of the core, is more desirable for experiments than a constantly moving axial flux peak resulting from a control system of axially positioned control components which are vertically withdrawn from the core.

  10. Light Water Reactor Fuel Cladding Research and Testing | ornl...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The...

  11. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  12. Advanced burner test reactor preconceptual design report.

    SciTech Connect (OSTI)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

  13. Beryllium Use in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Glen R. Longhurst

    2007-12-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) began operation in 1967. It makes use of a unique serpentine fuel core design and a beryllium reflector. Reactor control is achieved with rotating beryllium cylinders to which have been fastened plates of hafnium. Over time, the beryllium develops rather high helium content because of nuclear transmutations and begins to swell. The beryllium must be replaced at nominally 10-year intervals. Determination of when the replacement is made is by visual observation using a periscope to examine the beryllium surface for cracking and swelling. Disposition of the irradiated beryllium was once accomplished in the INL’s Radioactive Waste Management Complex, but that is no longer possible. Among contributing reasons are high levels of specific radioactive contaminants including transuranics. The INL is presently considering disposition pathways for this irradiated beryllium, but presently is storing it in the canal adjacent to the reactor. Numerous issues are associated with this situation including (1) Is there a need for ultra-low uranium material? (2) Is there a need to recover tritium from irradiated beryllium either because this is a strategic material resource or in preparation for disposal? (3) Is there a need to remove activation and fission products from irradiated beryllium? (4) Will there be enough material available to meet requirements for research reactors (fission and fusion)? In this paper will be discussed the present status of considerations on these issues.

  14. Hydraulic tests of emergency cooling system: L-Area

    SciTech Connect (OSTI)

    Hinton, J H

    1988-01-01

    The delay in L-Area startup provided an opportunity to obtain valuable data on the Emergency Cooling System (ECS) which will permit reactor operation at the highest safe power level. ECS flow is a major input to the FLOOD code which calculates reactor ECS power limits. The FLOOD code assesses the effectiveness of the ECS cooling capacity by modeling the core and plenum hydraulics under accident conditions. Presently, reactor power is not limited by the ECS cooling capacity (power limit). However, the manual calculations of ECS flows had been recently updated to include piping changes (debris strainer, valve changes, pressure release systems) and update fitting losses. Both updates resulted in reduced calculated ECS flows. Upon completion of the current program to update, validate, and document, reactor power may be limited under certain situations by ECS cooling capacity for some present reactor charge designs. A series of special hydraulic tests (Reference 1, 3) were conducted in L-Area using all sources of emergency coolant including the ECS pumps (Reference 2). The tests provided empirical hydraulic data on the ECS piping. These data will be used in computer models of the system as well as manual calculations of ECS flows. The improved modeling and accuracy of the flow calculations will permit reactor operation at the highest safe power level with respect to an ECS power limit.

  15. The behavior of fission products during nuclear rocket reactor tests

    SciTech Connect (OSTI)

    Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

    1991-01-01

    The experience base regarding fission product behavior developed during the Rover program, the nuclear rocket development program of 1955--1972, will be useful in planning a renewed nuclear rocket program. During the Rover program, 20 reactors were tested at the Nuclear Rocket Development Station in Nevada. Nineteen of these discharged effluent directly into the atmosphere; the last reactor tested, a non-flight-prototypic, fuel-element-testing reactor called the Nuclear Furnace (NF-1) was connected to an effluent cleanup system that removed fission products before the hydrogen coolant (propellant) was discharged to the atmosphere. In general, we are able to increase both test duration and fuel temperature during the test series. Therefore fission product data from the later part of the program are more interesting and more applicable to future reactors. We have collected fission product retention (and release) data reported in both formal and informal publications for six of the later reactor tests; five of these were Los Alamos reactors that were firsts of a kind in configuration or operating conditions. We have also, with the cooperation of Westinghouse, included fission product data from the NRX-A6 reactor, the final member of series of developmental reactors with the same basic geometry, but with significant design and fabrication improvements as the series continued. Table 1 lists the six selected reactors and the test parameters for each.

  16. CHARACTERIZATION OF RADIOACTIVITY IN THE REACTOR VESSEL OF THE HEAVY WATER COMPONENT TEST REACTOR

    SciTech Connect (OSTI)

    Vinson, Dennis

    2010-06-01

    The Heavy Water Component Test Reactor (HWCTR) facility is a pressurized heavy water reactor that was used to test candidate fuel designs for heavy water power reactors. The reactor operated at nominal power of 50 MW{sub th}. The reactor coolant loop operated at 1200 psig and 250 C. Two isolated test loop were designed into the reactor to provide special test conditions. Fig. 1 shows a cut-away view of the reactor. The two loops are contained in four inch diameter stainless steel piping. The HWCTR was operated for only a short duration, from March 1962 to December 1964 in order to test the viability of test fuel elements and other reactor components for use in a heavy water power reactor. The reactor achieved 13,882 MWd of total power while testing 36 different fuel assemblies. In the course of operation, HWCTR experienced the cladding failures of 10 separate test fuel assemblies. In each case, the cladding was breached with some release of fuel core material into the isolated test loop, causing fission product and actinide contamination in the main coolant loop and the liquid and boiling test loops. Despite the contribution of the contamination from the failed fuel, the primary source of radioactivity in the HWCTR vessel and internals is the activation products in the thermal shields, and to a lesser degree, activation products in the reactor vessel walls and liner. A detailed facility characterization report of the HWCTR facility was completed in 1996. Many of the inputs and assumptions in the 1996 characterization report were derived from the HWCTR decommissioning plan published in 1975. The current paper provides an updated assessment of the radioisotopic characteristics of the HWCTR vessel and internals to support decommissioning activities on the facility.

  17. Advanced Test Reactor outage risk assessment

    SciTech Connect (OSTI)

    Thatcher, T.A.; Atkinson, S.A.

    1997-12-31

    Beginning in 1997, risk assessment was performed for each Advanced Test Reactor (ATR) outage aiding the coordination of plant configuration and work activities (maintenance, construction projects, etc.) to minimize the risk of reactor fuel damage and to improve defense-in-depth. The risk assessment activities move beyond simply meeting Technical Safety Requirements to increase the awareness of risk sensitive configurations, to focus increased attention on the higher risk activities, and to seek cost-effective design or operational changes that reduce risk. A detailed probabilistic risk assessment (PRA) had been performed to assess the risk of fuel damage during shutdown operations including heavy load handling. This resulted in several design changes to improve safety; however, evaluation of individual outages had not been performed previously and many risk insights were not being utilized in outage planning. The shutdown PRA provided the necessary framework for assessing relative and absolute risk levels and assessing defense-in-depth. Guidelines were written identifying combinations of equipment outages to avoid. Screening criteria were developed for the selection of work activities to receive review. Tabulation of inherent and work-related initiating events and their relative risk level versus plant mode has aided identification of the risk level the scheduled work involves. Preoutage reviews are conducted and post-outage risk assessment is documented to summarize the positive and negative aspects of the outage with regard to risk. The risk for the outage is compared to the risk level that would result from optimal scheduling of the work to be performed and to baseline or average past performance.

  18. Automated Test Coverage Measurement for Reactor Protection System Software

    E-Print Network [OSTI]

    in implementing safety critical systems such as nuclear reactor protection systems. We have defined new structural) are widely used to implement safety- critical systems such as nuclear reactor protection systems, testing implementation language. The Korea Nuclear Instrumentation and Control System R&D Center (KNICS) project, whose

  19. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect (OSTI)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  20. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    SciTech Connect (OSTI)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  1. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    SciTech Connect (OSTI)

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  2. Cameron.Hardy@rl.doe.gov F Reactor Area Cleanup Complete

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    or cocooned, the F Reactor Area is the first to have all of its associated buildings and waste sites cleaned up in addition to having its reactor sealed up. "The cleanup of the F...

  3. Technology Options for a Fast Spectrum Test Reactor

    SciTech Connect (OSTI)

    D. M. Wachs; R. W. King; I. Y. Glagolenko; Y. Shatilla

    2006-06-01

    Idaho National Laboratory in collaboration with Argonne National Laboratory has evaluated technology options for a new fast spectrum reactor to meet the fast-spectrum irradiation requirements for the USDOE Generation IV (Gen IV) and Advanced Fuel Cycle Initiative (AFCI) programs. The US currently has no capability for irradiation testing of large volumes of fuels or materials in a fast-spectrum reactor required to support the development of Gen IV fast reactor systems or to demonstrate actinide burning, a key element of the AFCI program. The technologies evaluated and the process used to select options for a fast irradiation test reactor (FITR) for further evaluation to support these programmatic objectives are outlined in this paper.

  4. Acoustic emission monitoring of hot functional testing: Watts Bar Unit 1 Nuclear Reactor

    SciTech Connect (OSTI)

    Hutton, P.H.; Dawson, J.F.; Friesel, M.A.; Harris, J.C.; Pappas, R.A.

    1984-06-01

    Acoustic emission (AE) monitoring of selected pressure boundary areas at TVA's Watts Bar, Unit 1 Nuclear Power Plant during hot functional preservice testing is described in this report. The report deals with background, methodology, and results. The work discussed here is a major milestone in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. The subject work demonstrated that anticipated problem areas can be overcome. Work is continuing toward AE monitoring during reactor operation.

  5. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  6. Advanced Test Reactor - A National Scientific User Facility

    SciTech Connect (OSTI)

    Clifford J. Stanley

    2008-05-01

    The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and energy can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ~1.0 x1015 n/cm2-sec with a maximum fast flux of ~5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

  7. In-Situ Creep Testing Capability for the Advanced Test Reactor

    SciTech Connect (OSTI)

    B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

    2012-09-01

    An instrumented creep testing capability is being developed for specimens irradiated in Pressurized Water Reactor (PWR) coolant conditions at the Advanced Test Reactor (ATR). The test rig has been developed such that samples will be subjected to stresses ranging from 92 to 350 MPa at temperatures between 290 and 370 °C up to at least 2 dpa (displacement per atom). The status of Idaho National Laboratory (INL) efforts to develop the test rig in-situ creep testing capability for the ATR is described. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper reports efforts by INL to evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory (HTTL). Initial data from autoclave tests with 304 stainless steel (304 SS) specimens are reported.

  8. Removal of the Plutonium Recycle Test Reactor - 13031

    SciTech Connect (OSTI)

    Herzog, C. Brad [CH2M HILL, Inc. (United States)] [CH2M HILL, Inc. (United States); Guercia, Rudolph [US-DOE (United States)] [US-DOE (United States); LaCome, Matt [Meier Engineering Inc (United States)] [Meier Engineering Inc (United States)

    2013-07-01

    The 309 Facility housed the Plutonium Recycle Test Reactor (PRTR), an operating test reactor in the 300 Area at Hanford, Washington. The reactor first went critical in 1960 and was originally used for experiments under the Hanford Site Plutonium Fuels Utilization Program. The facility was decontaminated and decommissioned in 1988-1989, and the facility was deactivated in 1994. The 309 facility was added to Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) response actions as established in an Interim Record of Decision (IROD) and Action Memorandum (AM). The IROD directs a remedial action for the 309 facility, associated waste sites, associated underground piping and contaminated soils resulting from past unplanned releases. The AM directs a removal action through physical demolition of the facility, including removal of the reactor. Both CERCLA actions are implemented in accordance with U.S. EPA approved Remedial Action Work Plan, and the Remedial Design Report / Remedial Action Report associated with the Hanford 300-FF-2 Operable Unit. The selected method for remedy was to conventionally demolish above grade structures including the easily distinguished containment vessel dome, remove the PRTR and a minimum of 300 mm (12 in) of shielding as a single 560 Ton unit, and conventionally demolish the below grade structure. Initial sample core drilling in the Bio-Shield for radiological surveys showed evidence that the Bio-Shield was of sound structure. Core drills for the separation process of the PRTR from the 309 structure began at the deck level and revealed substantial thermal degradation of at least the top 1.2 m (4LF) of Bio-Shield structure. The degraded structure combined with the original materials used in the Bio-Shield would not allow for a stable structure to be extracted. The water used in the core drilling process proved to erode the sand mixture of the Bio-Shield leaving the steel aggregate to act as ball bearings against the core drill bit. A redesign is being completed to extract the 309 PRTR and entire Bio-Shield structure together as one monolith weighing 1100 Ton by cutting structural concrete supports. In addition, the PRTR has hundreds of contaminated process tubes and pipes that have to be severed to allow for a uniformly flush fit with a lower lifting frame. Thirty-two 50 mm (2 in) core drills must be connected with thirty-two wire saw cuts to allow for lifting columns to be inserted. Then eight primary saw cuts must be completed to severe the PRTR from the 309 Facility. Once the weight of the PRTR is transferred to the lifting frame, then the PRTR may be lifted out of the facility. The critical lift will be executed using four 450 Ton strand jacks mounted on a 9 m (30 LF) tall mobile lifting frame that will allow the PRTR to be transported by eight 600 mm (24 in) Slide Shoes. The PRTR will then be placed on a twenty-four line, double wide, self powered Goldhofer for transfer to the onsite CERCLA Disposal Cell (ERDF Facility), approximately 33 km (20 miles) away. (authors)

  9. Materials Test-2 LOCA Simulation in the NRU Reactor

    SciTech Connect (OSTI)

    Barner, J. O.; Hesson, G. M.; King, I. L.; Marshall, R. K.; Parchen, L. J.; Pilger, J. P.; Rausch, W. N.; Russcher, G. E.; Webb, B. J.; Wildung, N. J.; Wilson, C. L.; Wismer, M. D.; Mohr, C. L.

    1982-03-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This third experiment of the program produced fuel cladding temperatures exceeding 1033 K (1400°F) for 155 s and resulted in eight ruptured fuel rods. Experiment data and initial results are presented in the form of photographs and graphical summaries.

  10. Reactor protection system with automatic self-testing and diagnostic

    DOE Patents [OSTI]

    Gaubatz, Donald C. (Cupertino, CA)

    1996-01-01

    A reactor protection system having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically "identical" values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic.

  11. Reactor protection system with automatic self-testing and diagnostic

    DOE Patents [OSTI]

    Gaubatz, D.C.

    1996-12-17

    A reactor protection system is disclosed having four divisions, with quad redundant sensors for each scram parameter providing input to four independent microprocessor-based electronic chassis. Each electronic chassis acquires the scram parameter data from its own sensor, digitizes the information, and then transmits the sensor reading to the other three electronic chassis via optical fibers. To increase system availability and reduce false scrams, the reactor protection system employs two levels of voting on a need for reactor scram. The electronic chassis perform software divisional data processing, vote 2/3 with spare based upon information from all four sensors, and send the divisional scram signals to the hardware logic panel, which performs a 2/4 division vote on whether or not to initiate a reactor scram. Each chassis makes a divisional scram decision based on data from all sensors. Automatic detection and discrimination against failed sensors allows the reactor protection system to automatically enter a known state when sensor failures occur. Cross communication of sensor readings allows comparison of four theoretically ``identical`` values. This permits identification of sensor errors such as drift or malfunction. A diagnostic request for service is issued for errant sensor data. Automated self test and diagnostic monitoring, sensor input through output relay logic, virtually eliminate the need for manual surveillance testing. This provides an ability for each division to cross-check all divisions and to sense failures of the hardware logic. 16 figs.

  12. LOCA simulation in the NRU reactor: materials test-1

    SciTech Connect (OSTI)

    Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

    1981-10-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607/sup 0/F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions.

  13. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    SciTech Connect (OSTI)

    Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley; Steven Taylor

    2012-08-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  14. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    SciTech Connect (OSTI)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; K. Condie

    2011-06-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility (NSUF) in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  15. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  16. Reactor Testing and Qualification: Prioritized High-level Criticality Testing Needs

    SciTech Connect (OSTI)

    S. Bragg-Sitton; J. Bess; J. Werner; G. Harms; S. Bailey

    2011-09-01

    Researchers at the Idaho National Laboratory (INL) were tasked with reviewing possible criticality testing needs to support development of the fission surface power system reactor design. Reactor physics testing can provide significant information to aid in development of technologies associated with small, fast spectrum reactors that could be applied for non-terrestrial power systems, leading to eventual system qualification. Several studies have been conducted in recent years to assess the data and analyses required to design and build a space fission power system with high confidence that the system will perform as designed [Marcille, 2004a, 2004b; Weaver, 2007; Parry et al., 2008]. This report will provide a summary of previous critical tests and physics measurements that are potentially applicable to the current reactor design (both those that have been benchmarked and those not yet benchmarked), summarize recent studies of potential nuclear testing needs for space reactor development and their applicability to the current baseline fission surface power (FSP) system design, and provide an overview of a suite of tests (separate effects, sub-critical or critical) that could fill in the information database to improve the accuracy of physics modeling efforts as the FSP design is refined. Some recommendations for tasks that could be completed in the near term are also included. Specific recommendations on critical test configurations will be reserved until after the sensitivity analyses being conducted by Los Alamos National Laboratory (LANL) are completed (due August 2011).

  17. Preliminary safety evaluation of the advanced burner test reactor.

    SciTech Connect (OSTI)

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  18. Vital area identification for U.S. Nuclear Regulatory Commission nuclear power reactor licensees and new reactor applicants.

    SciTech Connect (OSTI)

    Whitehead, Donnie Wayne; Varnado, G. Bruce

    2008-09-01

    U.S. Nuclear Regulatory Commission nuclear power plant licensees and new reactor applicants are required to provide protection of their plants against radiological sabotage, including the placement of vital equipment in vital areas. This document describes a systematic process for the identification of the minimum set of areas that must be designated as vital areas in order to ensure that all radiological sabotage scenarios are prevented. Vital area identification involves the use of logic models to systematically identify all of the malicious acts or combinations of malicious acts that could lead to radiological sabotage. The models available in the plant probabilistic risk assessment and other safety analyses provide a great deal of the information and basic model structure needed for the sabotage logic model. Once the sabotage logic model is developed, the events (or malicious acts) in the model are replaced with the areas in which the events can be accomplished. This sabotage area logic model is then analyzed to identify the target sets (combinations of areas the adversary must visit to cause radiological sabotage) and the candidate vital area sets (combinations of areas that must be protected against adversary access to prevent radiological sabotage). Any one of the candidate vital area sets can be selected for protection. Appropriate selection criteria will allow the licensee or new reactor applicant to minimize the impacts of vital area protection measures on plant safety, cost, operations, or other factors of concern.

  19. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect (OSTI)

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  20. Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)

    E-Print Network [OSTI]

    Rodriguez, Judy N

    2013-01-01

    The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

  1. HISTORICAL AMERICAN ENGINEERING RECORD - IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY, TEST AREA NORTH, HAER NO. ID-33-E

    SciTech Connect (OSTI)

    Susan Stacy; Hollie K. Gilbert

    2005-02-01

    Test Area North (TAN) was a site of the Aircraft Nuclear Propulsion (ANP) Project of the U.S. Air Force and the Atomic Energy Commission. Its Cold War mission was to develop a turbojet bomber propelled by nuclear power. The project was part of an arms race. Test activities took place in five areas at TAN. The Assembly & Maintenance area was a shop and hot cell complex. Nuclear tests ran at the Initial Engine Test area. Low-power test reactors operated at a third cluster. The fourth area was for Administration. A Flight Engine Test facility (hangar) was built to house the anticipated nuclear-powered aircraft. Experiments between 1955-1961 proved that a nuclear reactor could power a jet engine, but President John F. Kennedy canceled the project in March 1961. ANP facilities were adapted for new reactor projects, the most important of which were Loss of Fluid Tests (LOFT), part of an international safety program for commercial power reactors. Other projects included NASA's Systems for Nuclear Auxiliary Power and storage of Three Mile Island meltdown debris. National missions for TAN in reactor research and safety research have expired; demolition of historic TAN buildings is underway.

  2. Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    Budny, R.V.; Darrow, D.S.; Medley, S.S.; Nazikian, R.; Zweben, S.J.; et al.

    1998-12-14

    Alpha particle physics experiments were done on the Tokamak Fusion Test Reactor (TFTR) during its deuterium-tritium (DT) run from 1993-1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single-particle confinement model in magnetohydrodynamic (MHD) quiescent discharges. Also, the observed alpha particle interactions with sawteeth, toroidal Alfvén eigenmodes (TAE), and ion cyclotron resonant frequency (ICRF) waves were roughly consistent with theoretical modeling. This paper reviews what was learned and identifies what remains to be understood.

  3. Injectivity Test At Long Valley Caldera Geothermal Area (Morin...

    Open Energy Info (EERE)

    Injectivity Test At Long Valley Caldera Geothermal Area (Morin, Et Al., 1993) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Injectivity Test At...

  4. Injectivity Test At Long Valley Caldera Geothermal Area (Farrar...

    Open Energy Info (EERE)

    Injectivity Test At Long Valley Caldera Geothermal Area (Farrar, Et Al., 2003) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Injectivity Test...

  5. Flow Test At Blue Mountain Geothermal Area (Fairbank Engineering...

    Open Energy Info (EERE)

    Flow Test At Blue Mountain Geothermal Area (Fairbank Engineering Ltd, 2003) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Blue...

  6. Test Area North Pool Stabilization Project: Environmental assessment

    SciTech Connect (OSTI)

    NONE

    1996-05-01

    The Test Area North (TAN) Pool is located within the fenced TAN facility boundaries on the Idaho National Engineering Laboratory (INEL). The TAN pool stores 344 canisters of core debris from the March, 1979, Three Mile Island (TMI) Unit 2 reactor accident; fuel assemblies from Loss-of-Fluid Tests (LOFT); and Government-owned commercial fuel rods and assemblies. The LOFT and government owned commercial fuel rods and assemblies are hereafter referred to collectively as {open_quotes}commercial fuels{close_quotes} except where distinction between the two is important to the analysis. DOE proposes to remove the canisters of TMI core debris and commercial fuels from the TAN Pool and transfer them to the Idaho Chemical Processing Plant (ICPP) for interim dry storage until an alternate storage location other than at the INEL, or a permanent federal spent nuclear fuel (SNF) repository is available. The TAN Pool would be drained and placed in an industrially and radiologically safe condition for refurbishment or eventual decommissioning. This environmental assessment (EA) identifies and evaluates environmental impacts associated with (1) constructing an Interim Storage System (ISS) at ICPP; (2) removing the TMI and commercial fuels from the pool and transporting them to ICPP for placement in an ISS, and (3) draining and stabilizing the TAN Pool. Miscellaneous hardware would be removed and decontaminated or disposed of in the INEL Radioactive Waste Management Complex (RWMC). This EA also describes the environmental consequences of the no action alternative.

  7. FFTF (Fast Flux Test Facility) reactor shutdown system reliability reevaluation

    SciTech Connect (OSTI)

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations.

  8. Enhanced In-pile Instrumentation for Material Testing Reactors

    SciTech Connect (OSTI)

    Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley

    2012-07-01

    An increasing number of U.S. nuclear research programs are requesting enhanced in-pile instrumentation capable of providing real-time measurements of key parameters during irradiations. For example, fuel research and development funded by the U.S. Department of Energy now emphasize approaches that rely on first principle models to develop optimized fuel designs that offer significant improvements over current fuels. To facilitate this approach, high fidelity, real-time data are essential for characterizing the performance of new fuels during irradiation testing. Furthermore, sensors that obtain such data must be miniature, reliable and able to withstand high flux/high temperature conditions. Depending on user requirements, sensors may need to obtain data in inert gas, pressurized water, or liquid metal environments. To address these user needs, in-pile instrumentation development efforts have been initiated as part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF), the Fuel Cycle Research & Development (FCR&D), and the Nuclear Energy Enabling Technology (NEET) programs. This paper reports on recent INL achievements to support these programs. Specifically, an overview of the types of sensors currently available to support in-pile irradiations and those sensors currently available to MTR users are identified. In addition, recent results and products available from sensor research and development are detailed. Specifically, progress in deploying enhanced in-pile sensors for detecting elongation and thermal conductivity are reported. Results from research to evaluate the viability of ultrasonic and fiber optic technologies for irradiation testing are also summarized.

  9. Testing of an advanced thermochemical conversion reactor system

    SciTech Connect (OSTI)

    Not Available

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  10. Record of Cycling Operation of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon

    E-Print Network [OSTI]

    Record of Cycling Operation of the Natural Nuclear Reactor in the Oklo/Okelobondo Area in Gabon A billion yr old Oklo natural nuclear reactor. In addition to elevated abundances of fission-produced Zr, Ce nuclear chain reaction was predicted by Kuroda [1] 20 years before the remnants of the natural reactor

  11. Core design studies for advanced burner test reactor.

    SciTech Connect (OSTI)

    Yang, W. S.; Kim, T. K.; Hill, R. N.; Nuclear Engineering Division

    2008-01-01

    The U.S. government announced in February 2006 the Global Nuclear Energy Partnership (GNEP) to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. The advanced burner reactor (ABR) based on a fast spectrum is one of the three major technologies to be demonstrated in GNEP. In FY06, a pre-conceptual design study was performed to develop an advanced burner test reactor (ABTR) that supports development of a prototype full-scale ABR, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR were (1) to demonstrate reactor-based transmutation of transuranics (TRU) as part of an advanced fuel cycle, (2) to qualify the TRU-containing fuels and advanced structural materials needed for a full-scale ABR, (3) to support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. Based on these objectives, core design and fuel cycle studies were performed to develop ABTR core designs, which can accommodate the expected changes of the TRU feed and the conversion ratio. Various option and trade-off studies were performed to determine the appropriate power level and conversion ratio. Both ternary metal alloy (U-TRU-10Zr) and mixed oxide (UO{sub 2}-TRUO{sub 2}) fuel forms have been considered with TRU feeds from weapons-grade plutonium (WG-Pu) and TRU recovered from light water reactor spent fuel (LWR-SF). Reactor performances were evaluated in detail including equilibrium cycle core parameters, mass flow, power distribution, kinetic parameters, reactivity feedback coefficient, reactivity control requirements and shutdown margins, and spent fuel characteristics. Trade-off studies on power level suggested that about 250 MWt is a reasonable compromise to allow a low project cost, at the same time providing a reasonable prototypic irradiation environment for demonstrating TRU-based fuels. Preliminary design studies showed that it is feasible to design the ABTR to accommodate a wide range of conversion ratio (CR) by employing different assembly designs. The TRU enrichments required for various conversion ratios and the irradiation database suggested a phased approach with initial startup using conventional enrichment plutonium-based fuel and gradual transitioning to full core loading of transmutation fuel after its qualification phase (resulting in {approx}0.6 CR). The low CR transmutation fuel tests can be accommodated in the designated test assemblies, and if fully developed, core conversion to low CR fuel can be envisioned. Reference ABTR core designs with a rated power of 250 MWt were developed for ternary metal alloy and mixed oxide fuels based on WG-Pu feed. The reference core contains 54 driver, 6 test fuel, and 3 test material assemblies. For the startup core designs, the calculated TRU conversion ratio is 0.65 for the metal fuel core and 0.64 for the oxide fuel core. Both the metal and oxide cores show good performances. The metal fuel core requires an average TRU enrichment of 18.8% and yields a reactivity swing of 1.2 %{Delta}k over the 4-month cycle. The core average flux level is {approx}2.4 x 10{sup 15} n/cm{sup 2}s, and test assembly flux level is {approx}2.8 x 10{sup 15} n/cm{sup 2}s. Compared to the metal fuel core, the lower density oxide fuel core requires an average TRU enrichment of 21.8%, which results in a 780 kg TRU loading (as compared to 732 kg for metal) despite a {approx}9% smaller heavy metal inventory. The lower heavy metal inventory increases the burnup reactivity swing by {approx}10% and reduces the flux levels by {approx}8%. Alternative designs were also studied for a LWR-SF TRU feed and a low conversion ratio, including the recycle of the ABTR spent fuel TRU. The lower fissile contents of the LWR-SF TRU relative to the WG-Pu TRU significantly increase the required TRU enrichment of the startup cores to maintain the same cycle length. The even lower fissile fraction of the ABTR spent fuel TRU furt

  12. Recommended Underground Test Area Activity Reference Documents

    National Nuclear Security Administration (NNSA)

    Rainier MesaShoshone Mountain Rev 0, Nevada Test Site, Nevada, DOENV-1031, December 2004. (838783) o Phase II Corrective Action Investigation Plan for Corrective Action Units...

  13. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    SciTech Connect (OSTI)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  14. Safety Evaluation for Packaging for onsite Transfer of plutonium recycle test reactor ion exchange columns

    SciTech Connect (OSTI)

    Smith, R.J.

    1995-09-11

    The purpose of this Safety Evaluation for Packaging (SEP) is to authorize the use of three U.S. Department of Transportation (DOT) 7A, Type A metal boxes (Capital Industries Part No. S 0600-0600-1080- 0104) to package 12 Plutonium Recycle Test Reactor (PRTR) ion exchange columns as low-level waste (LLW). The packages will be transferred from the 309 Building in the 300 Area to low level waste burial in the 200 West Area. Revision 1 of WHC-SD-TP-SEP-035 (per ECN No. 621467) documents that the boxes containing ion exchange columns and grout will maintain the payload under normal conditions of transport if transferred without the box lids

  15. Nevada National Security Site Underground Test Area (UGTA) Flow...

    Office of Environmental Management (EM)

    and Transport Modeling - Approach and Example Nevada National Security Site Underground Test Area (UGTA) Flow and Transport Modeling - Approach and Example Bill Wilborn UGTA...

  16. Aerial Photography At Nevada Test And Training Range Area (Sabin...

    Open Energy Info (EERE)

    Aerial Photography At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Aerial...

  17. Geodetic Survey At Nevada Test And Training Range Area (Sabin...

    Open Energy Info (EERE)

    Geodetic Survey At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Geodetic Survey At...

  18. Geothermometry At Nevada Test And Training Range Area (Sabin...

    Open Energy Info (EERE)

    Geothermometry At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Geothermometry At...

  19. Testing of a 7-tube palladium membrane reactor for potential use in TEP

    SciTech Connect (OSTI)

    Carlson, Bryan J; Trujillo, Stephen; Willms, R. Scott

    2010-01-01

    A Palladium Membrane Reactor (PMR) consists of a palladium/silver membrane permeator filled with catalyst (catalyst may be inside or outside the membrane tubes). The PMR is designed to recover tritium from the methane, water, and other impurities present in fusion reactor effluent. A key feature of a PMR is that the total hydrogen isotope content of a stream is significantly reduced as (1) methane-steam reforming and/or water-gas shift reactions proceed on the catalyst bed and (2) hydrogen isotopes are removed via permeation through the membrane. With a PMR design matched to processing requirements, nearly complete hydrogen isotope removals can be achieved. A 3-tube PMR study was recently completed. From the results presented in this study, it was possible to conclude that a PMR is appropriate for TEP, perforated metal tube protectors function well, platinum on aluminum (PtA) catalyst performs the best, conditioning with air is probably required to properly condition the Pd/Ag tubes, and that CO/CO{sub 2} ratios maybe an indicator of coking. The 3-tube PMR had a permeator membrane area of 0.0247 m{sup 2} and a catalyst volume to membrane area ratio of 4.63 cc/cm{sup 2} (with the catalyst on the outside of the membrane tubes and the catalyst only covering the membrane tube length). A PMR for TEP will require a larger membrane area (perhaps 0.35 m{sup 2}). With this in mind, an intermediate sized PMR was constructed. This PMR has 7 permeator tubes and a total membrane area of 0.0851 m{sup 2}. The catalyst volume to membrane area ratio for the 7-tube PMR was 5.18 cc/cm{sup 2}. The total membrane area of the 7-tube PMR (0.0851 m{sup 2}) is 3.45 times larger than total membrane area of the 3-tube PMR (0.0247 m{sup 2}). The following objectives were identified for the 7-tube PMR tests: (1) Refine test measurements, especially humidity and flow; (2) Refine maintenance procedures for Pd/Ag tube conditioning; (3) Evaluate baseline PMR operating conditions; (4) Determine PMR scaling method; (5) Evaluate PMR with realistic feed compositions; (6) Evaluate PMR performance with varying permeate pressures; (7) Study coking-related issues; and (8) Identify any unexpected behavior that may require further investigation (used to study transient behavior). This report presents the tests results defined by these objectives.

  20. Identification of Selected Areas to Support Federal Clean Energy Goals Using Small Modular Reactors

    SciTech Connect (OSTI)

    Belles, Randy; Mays, Gary T; Omitaomu, Olufemi A; Poore III, Willis P

    2013-12-01

    This analysis identifies candidate locations, in a broad sense, where there are high concentrations of federal government agency use of electricity, which are also suitable areas for near-term SMRs. Near-term SMRs are based on light-water reactor (LWR) technology with compact design features that are expected to offer a host of safety, siting, construction, and economic benefits. These smaller plants are ideally suited for small electric grids and for locations that cannot support large reactors, thus providing utilities or governement entities with the flexibility to scale power production as demand changes by adding additional power by deploying more modules or reactors in phases. This research project is aimed at providing methodologies, information, and insights to assist the federal government in meeting federal clean energy goals.

  1. Adaptation of Crack Growth Detection Techniques to US Material Test Reactors

    SciTech Connect (OSTI)

    A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis; Joy L. Rempe; Gordon Kohse; Yakov Ostrovsky; David M. Carpenter

    2014-04-01

    A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some materials testing reactors (MTRs) outside the U.S., such as the Halden Boiling Water Reactor (HBWR), have deployed a technique to measure crack growth propagation during irradiation. This technique incorporates a compact loading mechanism to stress the specimen during irradiation. A crack in the specimen is monitored using the Direct Current Potential Drop (DCPD) method. A project is underway to develop and demonstrate the performance of a similar type of test rig for use in U.S. MTRs. The first year of this three year project was devoted to designing, analyzing, fabricating, and bench top testing a mechanism capable of applying a controlled stress to specimens while they are irradiated in a pressurized water loop (simulating PWR reactor conditions). During the second year, the mechanism will be tested in autoclaves containing high pressure, high temperature water with representative water chemistries. In addition, necessary documentation and safety reviews for testing in a reactor environment will be completed. In the third year, the assembly will be tested in the Massachusetts Institute of Technology Reactor (MITR) and Post Irradiation Examinations (PIE) will be performed.

  2. Observations of Fallout from the Fukushima Reactor Accident in San Francisco Bay Area Rainwater

    E-Print Network [OSTI]

    Eric B. Norman; Christopher T. Angell; Perry A. Chodash

    2011-03-30

    We have observed fallout from the recent Fukushima Dai-ichi reactor accident in samples of rainwater collected in the San Francisco Bay area. Gamma ray spectra measured from these samples show clear evidence of fission products - 131,132I, 132Te, and 134,137Cs. The activity levels we have measured for these isotopes are very low and pose no health risk to the public.

  3. Population Sensitivity Evaluation of Two Proposed Hampton Roads Area Sites for a Possible Small Modular Reactor

    SciTech Connect (OSTI)

    Belles, R. J.; Omitaomu, O. A.

    2014-08-01

    The overall objective of this research project is to use the OR-SAGE tool to support the US Department of Energy (DOE) Office of Nuclear Energy (NE) in evaluating future electrical generation deployment options for small modular reactors (SMRs) in areas with significant energy demand from the federal sector. Deployment of SMRs in zones with high federal energy use can provide a means of meeting federal clean energy goals.

  4. Mixed oxide fuels testing in the advanced test reactor to support plutonium disposition

    SciTech Connect (OSTI)

    Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S.

    1995-09-01

    An intense worldwide effort is now under way to find means of reducing the stockpile of weapons-grade plutonium. One of the most attractive solutions would be to use WGPu as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PUO{sub 2}) mixed with urania (UO{sub 2}). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification, (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania, (3) The effects of WGPu isotopic composition, (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight, (5) The effects of americium and gallium in WGPu, (6) Fission gas release from MOX fuel pellets made from WGPu, (7) Fuel/cladding gap closure, (8) The effects of power cycling and off-normal events on fuel integrity, (9) Development of radial distributions of burnup and fission products, (10) Power spiking near the interfaces of MOX and urania fuel assemblies, and (11) Fuel performance code validation. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified.

  5. Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

    SciTech Connect (OSTI)

    Romano, T.

    1997-09-29

    This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

  6. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect (OSTI)

    D. L. Knudson; J. L. Rempe

    2012-02-01

    New materials are being considered for fuel, cladding, and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine, and return irradiated samples for each measurement makes this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated in pressurized water reactor (PWR) coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL's High Temperature Test Laboratory (HTTL).

  7. Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

    SciTech Connect (OSTI)

    D. L. Knudson; J. L. Rempe

    2012-02-01

    New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.

  8. Advanced Test Reactor Design Basis Reconstitution Project Issue Resolution Process

    SciTech Connect (OSTI)

    Steven D. Winter; Gregg L. Sharp; William E. Kohn; Richard T. McCracken

    2007-05-01

    The Advanced Test Reactor (ATR) Design Basis Reconstitution Program (DBRP) is a structured assessment and reconstitution of the design basis for the ATR. The DBRP is designed to establish and document the ties between the Document Safety Analysis (DSA), design basis, and actual system configurations. Where the DBRP assessment team cannot establish a link between these three major elements, a gap is identified. Resolutions to identified gaps represent configuration management and design basis recovery actions. The proposed paper discusses the process being applied to define, evaluate, report, and address gaps that are identified through the ATR DBRP. Design basis verification may be performed or required for a nuclear facility safety basis on various levels. The process is applicable to large-scale design basis reconstitution efforts, such as the ATR DBRP, or may be scaled for application on smaller projects. The concepts are applicable to long-term maintenance of a nuclear facility safety basis and recovery of degraded safety basis components. The ATR DBRP assessment team has observed numerous examples where a clear and accurate link between the DSA, design basis, and actual system configuration was not immediately identifiable in supporting documentation. As a result, a systematic approach to effectively document, prioritize, and evaluate each observation is required. The DBRP issue resolution process provides direction for consistent identification, documentation, categorization, and evaluation, and where applicable, entry into the determination process for a potential inadequacy in the safety analysis (PISA). The issue resolution process is a key element for execution of the DBRP. Application of the process facilitates collection, assessment, and reporting of issues identified by the DBRP team. Application of the process results in an organized database of safety basis gaps and prioritized corrective action planning and resolution. The DBRP team follows the ATR DBRP issue resolution process which provides a method for the team to promptly sort and prioritize questions and issues between those that can be addressed as a normal part of the reconstitution project and those that are to be handle as PISAs. Presentation of the DBRP issue resolution process provides an example for similar activities that may be required at other facilities within the Department of Energy complex.

  9. Standard Guide for Conducting Supplemental Surveillance Tests for Nuclear Power Reactor Vessels, E 706 (IH)

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide discusses test procedures that can be used in conjunction with, but not as alternatives to, those required by Practices E185 and E2215 for the surveillance of nuclear reactor vessels. The supplemental mechanical property tests outlined permit the acquisition of additional information on radiation-induced changes in fracture toughness, notch ductility, and yield strength properties of the reactor vessel steels. 1.2 This guide provides recommendations for the preparation of test specimens for irradiation, and identifies special precautions and requirements for reactor surveillance operations and postirradiation test planning. Guidance on data reduction and computational procedures is also given. Reference is made to other ASTM test methods for the physical conduct of specimen tests and for raw data acquisition.

  10. Second test of base hydrolysate decomposition in a 0.04 gallon per minute scale reactor

    SciTech Connect (OSTI)

    Cena, R.J.; Thorsness, C.B.; Coburn, T.T.; Watkins, B.E.

    1994-10-11

    LLNL has built and operated a pilot plant for processing oil shale using recirculating hot solids. This pilot plant, was adapted in 1993 to demonstrate the feasibility of decomposing base hydrolysate, a mixture of sodium nitrite, sodium formate and other constituents. This material is the waste stream from the base hydrolysis process for destruction of energetic materials. In the Livermore process, the waste feed is thermally treated in a moving packed bed of ceramic spheres, where constituents in the waste decompose, in the presence of carbon dioxide, to form solid sodium carbonate and a suite of gases including: methane, carbon monoxide, oxygen, nitrogen oxides, ammonia and possibly molecular nitrogen. The ceramic spheres are circulated and heated, providing the energy required for thermal decomposition. The spheres provide a large surface area for evaporation and decomposition to occur, avoiding sticking and agglomeration of the waste. We performed a 2.5 hour test of the solids recirculation system, with continuous injection of approximately 0.04 gal/min of waste. Gasses from the packed bed reactor were directed through the lift pipe and water was not condensed. Potassium carbonate (0.356 M) was added to the hydrolysate prior to its introduction to the retort. Continuous on-line gas analysis was invaluable in tracking the progress of the experiment and quantifying the decomposition products. Analyses showed the primary solid product, collected in the lift exit cyclone, was indeed sodium carbonate, as expected. For the reactor condition studied in this test, N{sub 2}O was found to be the primary nitrogen bearing gas species. In the test, approximately equal quantities of ammonia and nitrogen bearing oxide gases were produced. Under proper conditions, this ammonia and NO{sub x} can be recombined downstream to form N{sub 2} and O{sub 2} as the primary effluent gases.

  11. Opportunities for mixed oxide fuel testing in the advanced test reactor to support plutonium disposition

    SciTech Connect (OSTI)

    Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W.

    1995-08-01

    Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. These issues include the following: (1) MOX fuel fabrication process verification; (2) Whether and how to use burnable poisons to depress MOX fuel initial reactivity, which is higher than that of urania; (3) The effects of WGPu isotopic composition; (4) The feasibility of loading MOX fuel with plutonia content up to 7% by weight; (5) The effects of americium and gallium in WGPu; (6) Fission gas release from MOX fuel pellets made from WGPu; (7) Fuel/cladding gap closure; (8) The effects of power cycling and off-normal events on fuel integrity; (9) Development of radial distributions of burnup and fission products; (10) Power spiking near the interfaces of MOX and urania fuel assemblies; and (11) Fuel performance code validation. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory possesses many advantages for performing tests to resolve most of the issues identified above. We have performed calculations to show that the use of hafnium shrouds can produce spectrum adjustments that will bring the flux spectrum in ATR test loops into a good approximation to the spectrum anticipated in a commercial LWR containing MOX fuel while allowing operation of the test fuel assemblies near their optimum values of linear heat generation rate. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. The requirements for planning and implementing a test program in the ATR have been identified. The facilities at Argonne National Laboratory-West can meet all potential needs for pre- and post-irradiation examination that might arise in a MOX fuel qualification program.

  12. 10 CFR 830 Major Modification Determination for Advanced Test Reactor LEU Fuel Conversion

    SciTech Connect (OSTI)

    Boyd D. Christensen; Michael A. Lehto; Noel R. Duckwitz

    2012-05-01

    The Advanced Test Reactor (ATR), located in the ATR Complex of the Idaho National Laboratory (INL), was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. The ATR is fueled with high-enriched uranium (HEU) matrix (UAlx) in an aluminum sandwich plate cladding. The National Nuclear Security Administration Global Threat Reduction Initiative (GTRI) strategic mission includes efforts to reduce and protect vulnerable nuclear and radiological material at civilian sites around the world. Converting research reactors from using HEU to low-enriched uranium (LEU) was originally started in 1978 as the Reduced Enrichment for Research and Test Reactors (RERTR) Program under the U.S. Department of Energy (DOE) Office of Science. Within this strategic mission, GTRI has three goals that provide a comprehensive approach to achieving this mission: The first goal, the driver for the modification that is the subject of this determination, is to convert research reactors from using HEU to LEU. Thus the mission of the ATR LEU Fuel Conversion Project is to convert the ATR and Advanced Test Reactor Critical facility (ATRC) (two of the six U.S. High-Performance Research Reactors [HPRR]) to LEU fuel by 2017. The major modification criteria evaluation of the project pre-conceptual design identified several issues that lead to the conclusion that the project is a major modification.

  13. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    SciTech Connect (OSTI)

    Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

    2011-10-31

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with â??warm boreâ?ť diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged â??spiderâ?ť design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project â??Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limitersâ?ť was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZPâ??s product development program, the amount of HTS wire employed per FCL and its cost as a percentage of the total FCL product content had not dropped substantially from an unsustainable level of more than 50% of the total cost of the FCL, nor had the availability increased (today the availability of 2G wire for commercial applications outside of specific partnerships with the leading 2G wire manufacturers is extremely limited). ZP had projected a very significant commercial potential for FCLs with higher performance and lower costs compared to the initial models built with 1G wire, which would come about from the widespread availability of low-cost, high-performance 2G HTS wire. The potential for 2G wires at greatly reduced performance-based prices compared to 1G HTS conductor held out the potential for the commercial production of FCLs at price and performance levels attractive to the utility industry. However, the price of HTS wire did not drop as expected and today the available quantities of 2G wire are limited, and the price is higher than the currently available supplies of 1G wire. The commercial option for ZP to provide a reliable and reasonably priced FCL to the utility industry is to employ conventional resistive conductor DC electromagnets to bias the FCL. Since the premise of the original funding was to stimulate the HTS wire industry and ZP concluded that copper-based magnets were more economical for the foreseeable future, DOE and ZP decided to mutually terminate the project.

  14. Hydrogen loops in existing reactors for testing fuel elements for nuclear propulsion

    SciTech Connect (OSTI)

    Olsen, C.S.; Welland, H.; Abraschoff, J. (Idaho National Engineering Laboratory, EG G Idaho Inc., P.O. Box 1625, Idaho Falls, Idaho 83415 (United States)); Thoms, K. (Oak Ridge National Laboratory, P.O. Box, Oak Ridge, Tennessee 37831-8087 (United States))

    1993-01-15

    The Space Exploration Initiative (SEI) has revitalized interest in adapting nuclear energy for power and propulsion. Prior to the selection of a nuclear thermal propulsion (NTP) system, extensive testing of the various proposed concepts will be required. In today's environmental, safety and health culture, full size rocket engine tests as were done under the Rover/NERVA program will be extremely difficult and expensive to perform and meet NASA's schedules. A different test strategy uses a hydrogen loop in an existing reactor to test a wide variety of single elements or clusters of elements for fuel qualification. This approach is expected to reduce operating and capital costs and expedite the testing schedule. This paper examines the potential of performing subscale tests in a hydrogen loop in an existing reactor such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. The HFIR is expected to achieve power densities comparable to those achieved in ATR because of the 85 MWt power level and the high thermal and fast flux levels. The available length and diameter of the test region of FHIR are 60 cm and 10 cm whereas the available length and diameter of the test region of ATR are 120 cm and 12 cm respectively.

  15. MODELING ASSUMPTIONS FOR THE ADVANCED TEST REACTOR FRESH FUEL SHIPPING CONTAINER

    SciTech Connect (OSTI)

    Rick J. Migliore

    2009-09-01

    The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is currently licensed per 10 CFR 71 to transport a fresh fuel element for either the Advanced Test Reactor, the University of Missouri Research Reactor (MURR), or the Massachusetts Institute of Technology Research Reactor (MITR-II). During the licensing process, the Nuclear Regulatory Commission (NRC) raised a number of issues relating to the criticality analysis, namely (1) lack of a tolerance study on the fuel and packaging, (2) moderation conditions during normal conditions of transport (NCT), (3) treatment of minor hydrogenous packaging materials, and (4) treatment of potential fuel damage under hypothetical accident conditions (HAC). These concerns were adequately addressed by modifying the criticality analysis. A tolerance study was added for both the packaging and fuel elements, full-moderation was included in the NCT models, minor hydrogenous packaging materials were included, and fuel element damage was considered for the MURR and MITR-II fuel types.

  16. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    SciTech Connect (OSTI)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these experiments were of particular importance because they provide extensive information which can be directly applied to the design of large LMFBR’s. It should be recognized that the data presented in the initial report were evaluated only to the extent necessary to ensure that adequate data were obtained. Later reports provided further interpretation and detailed comparisons with prediction techniques. The conclusion of the isothermal physics measurements was that the FFTF nuclear characteristics were essentially as designed and all safety requirements were satisfied. From a nuclear point of view, the FFTF was qualified to proceed into power operation mode. The FFTF was completed in 1978 and first achieved criticality on February 9, 1980. Upon completion of the isothermal physics and reactor characterization programs, the FFTF operated for ten years from April 1982 to April 1992. Reactor operations of the FFTF were terminated and the reactor facility was then defueled, deactivated, and placed into cold standby condition. Deactivation of the reactor was put on hold from 1996 to 2000 while the U.S. Department of Energy examined alternative uses for the FFTF but then announced the permanent deactivation of the FFTF in December 2001. Its core support basket was later drilled in May 2005, so as to remove all remaining sodium coolant. On April 17, 2006, the American Nuclear Society designated the FFTF as a “National Nuclear Historic Landmark”.

  17. Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information

    SciTech Connect (OSTI)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

  18. 100 Area soil washing bench-scale test procedures

    SciTech Connect (OSTI)

    Freeman, H.D.; Gerber, M.A.; Mattigod, S.V.; Serne, R.J.

    1993-03-01

    This document describes methodologies and procedures for conducting soil washing treatability tests in accordance with the 100 Area Soil Washing Treatability Test Plan (DOE-RL 1992, Draft A). The objective of this treatability study is to evaluate the use of physical separation systems and chemical extraction methods as a means of separating chemically and radioactively contaminated soil fractions from uncontaminated soil fractions. These data will be primarily used for determining feasibility of the individual unit operations and defining the requirements for a system, or systems, for pilot-scale testing.

  19. Activated carbon testing for the 200 area effluent treatment facility

    SciTech Connect (OSTI)

    Wagner, R.N.

    1997-01-17

    This report documents pilot and laboratory scale testing of activated carbon for use in the 200 Area Effluent Treatment Facility peroxide decomposer columns. Recommendations are made concerning column operating conditions and hardware design, the optimum type of carbon for use in the plant, and possible further studies.

  20. Testing of a Transport Cask for Research Reactor Spent Fuel - 13003

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

    2013-07-01

    Since the beginning of the last decade three Latin American countries that operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the TRIGA and MTR reactors operated in the region. A main drive in this initiative, sponsored by the International Atomic Energy Agency, is the fact that no definite solution regarding the back end of the research reactor fuel cycle has been taken by any of the participating country. However, any long-term solution - either disposition in a repository or storage away from reactor - will involve at some stage the transportation of the spent fuel through public roads. Therefore, a licensed cask that provides adequate shielding, assurance of subcriticality, and conformance to internationally accepted safety, security and safeguards regimes is considered a strategic part of any future solution to be adopted at a regional level. As a step in this direction, a packaging for the transport of irradiated fuel for MTR and TRIGA research reactors was designed by the tri-national team and a half-scale model equipped with the MTR version of the internal basket was constructed in Argentina and Brazil and tested in Brazil. Three test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. After failing the tests in the first two test series, the specimen successfully underwent the last test sequence. A second specimen, incorporating the structural improvements in view of the previous tests results, will be tested in the near future. Numerical simulations of the free drop and thermal tests are being carried out in parallel, in order to validate the computational modeling that is going to be used as a support for the package certification. (authors)

  1. NEUTRON ACTIVATION COOLDOWN OF THE TOKAMAK FUSION TEST REACTOR

    E-Print Network [OSTI]

    maintenance, reliability, and performance requirements. Fig. 1 shows a partial schematic plan view of the TFTR system pump ducts, and the activation measurements. The characteristics of the Test Cell shielding

  2. Continuous-flow stirred-tank reactor 20-L demonstration test: Final report

    SciTech Connect (OSTI)

    Lee, D.D.; Collins, J.L.

    2000-02-01

    One of the proposed methods of removing the cesium, strontium, and transuranics from the radioactive waste storage tanks at Savannah River is the small-tank tetraphenylborate (TPB) precipitation process. A two-reactor-in-series (15-L working volume each) continuous-flow stirred-tank reactor (CSTR) system was designed, constructed, and installed in a hot cell to test the Savannah River process. The system also includes two cross-flow filtration systems to concentrate and wash the slurry produced in the process, which contains the bulk of radioactivity from the supernatant processed through the system. Installation, operational readiness reviews, and system preparation and testing were completed. The first test using the filtration systems, two CSTRs, and the slurry concentration system was conducted over a 61-h period with design removal of Cs, Sr, and U achieved. With the successful completion of Test 1a, the following tests, 1b and 1c, were not required.

  3. Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests

    SciTech Connect (OSTI)

    Braley, Jenifer C.; Sinkov, Sergey I.; Delegard, Calvin H.; Schmidt, Andrew J.

    2011-06-15

    Scoping tests to evaluate the effects of agitation and pH adjustment on simulant sludge agglomeration and uranium metal oxidation at {approx}95 C were performed under Test Instructions(a,b) and as per sections 5.1 and 5.2 of this Test Plan prepared by AREVA. (c) The thermal testing occurred during the week of October 4-9, 2010. The results are reported here. For this testing, two uranium-containing simulant sludge types were evaluated: (1) a full uranium-containing K West (KW) container sludge simulant consisting of nine predominant sludge components; (2) a 50:50 uranium-mole basis mixture of uraninite [U(IV)] and metaschoepite [U(VI)]. This scoping study was conducted in support of the Sludge Treatment Project (STP) Phase 2 technology evaluation for the treatment and packaging of K-Basin sludge. The STP is managed by CH2M Hill Plateau Remediation Company (CHPRC) for the U.S. Department of Energy. Warm water ({approx}95 C) oxidation of sludge, followed by immobilization, has been proposed by AREVA and is one of the alternative flowsheets being considered to convert uranium metal to UO{sub 2} and eliminate H{sub 2} generation during final sludge disposition. Preliminary assessments of warm water oxidation have been conducted, and several issues have been identified that can best be evaluated through laboratory testing. The scoping evaluation documented here was specifically focused on the issue of the potential formation of high strength sludge agglomerates at the proposed 95 C process operating temperature. Prior hydrothermal tests conducted at 185 C produced significant physiochemical changes to genuine sludge, including the formation of monolithic concretions/agglomerates that exhibited shear strengths in excess of 100 kPa (Delegard et al. 2007).

  4. Technical bases to consider for performance and demonstration testing of space fission reactors

    SciTech Connect (OSTI)

    Hixson, L. L. (Laurie L.); Houts, M. G. (Michael G.); Clement, S. D. (Steven D.)

    2004-01-01

    Performance and demonstration testing are critical to the success of a space fission reactor program. However, the type and extent to which testing of space reactors should be performed has been a point of discussion within the industry for many years. With regard to full power ground nuclear tests, questions such as: (1) Do the benefits outweigh the risks; (2) Are there equivalent alternatives; (3) Can a test facility be constructed (or modified) in a reasonable amount of time; (4) Will the test article accurately represent the flight system; and (5) Are the costs too restrictive, have been debated for decades. There are obvious benefits of full power ground nuclear testing such as obtaining systems integrated reliability data on a full-scale, complete end-to-end system. But these benefits come at some programmatic risk. In addition, this type of testing does not address safety related issues. This paper will discuss and assess these and other technical considerations essential in deciding which type of performance and demonstration testing to conduct on space fission reactor systems.

  5. Evaluation of Fluid Conduction and Mixing within a Subassembly of the Actinide Burner Test Reactor

    SciTech Connect (OSTI)

    Cliff B. Davis

    2007-09-01

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of the Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid, including axial and radial heat conduction and subchannel mixing, that are not currently represented with internal code models. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor.

  6. In-situ Creep Testing Capability Development for Advanced Test Reactor

    SciTech Connect (OSTI)

    B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

    2010-08-01

    Creep is the slow, time-dependent strain that occurs in a material under a constant strees (or load) at high temperature. High temperature is a relative term, dependent on the materials being evaluated. A typical creep curve is shown in Figure 1-1. In a creep test, a constant load is applied to a tensile specimen maintained at a constant temperature. Strain is then measured over a period of time. The slope of the curve, identified in the figure below, is the strain rate of the test during Stage II or the creep rate of the material. Primary creep, Stage I, is a period of decreasing creep rate due to work hardening of the material. Primary creep is a period of primarily transient creep. During this period, deformation takes place and the resistance to creep increases until Stage II, Secondary creep. Stage II creep is a period with a roughly constant creep rate. Stage II is referred to as steady-state creep because a balance is achieved between the work hardening and annealing (thermal softening) processes. Tertiary creep, Stage III, occurs when there is a reduction in cross sectional area due to necking or effective reduction in area due to internal void formation; that is, the creep rate increases due to necking of the specimen and the associated increase in local stress.

  7. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    SciTech Connect (OSTI)

    Durbin, Samuel; Lindgren, Eric R.

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below-ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

  8. 2010 Radiological Monitoring Results Associated with the Advance Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    mike lewis

    2011-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  9. 2013 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2014-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  10. 2012 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2013-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  11. 2011 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2012-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  12. Comment on Li pellet conditioning in tokamak fusion test reactor R. V. Budny

    E-Print Network [OSTI]

    Budny, Robert

    Comment on Li pellet conditioning in tokamak fusion test reactor R. V. Budny Princeton Plasma; published online 9 September 2011) Li pellet conditioning in TFTR results in a reduction of the edge technique for introducing Li is via pellet injection. This was pioneered in ALCATOR- CMOD where it was first

  13. Analytical Study of High Concentration PCB Paint at the Heavy Water Components Test Reactor

    SciTech Connect (OSTI)

    Lowry, N.J.

    1998-10-21

    This report provides results of an analytical study of high concentration PCB paint in a shutdown nuclear test reactor located at the US Department of Energy's Savannah River Site (SRS). The study was designed to obtain data relevant for an evaluation of potential hazards associated with the use of and exposure to such paints.

  14. Development of a universal diagnostic probe system for Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    Mastronardi, R.; Cabral, R.; Manos, D.

    1982-05-01

    The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

  15. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The first experiment was inserted in the ATR in August 2009 and started its irradiation in September 2009. It is anticipated to complete its irradiation in early calendar 2011. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and the irradiation experience to date.

  16. The possibility of high amplitude driven contained modes during ion Bernstein wave experiments in the tokamak fusion test reactor

    E-Print Network [OSTI]

    in the Tokamak Fusion Test Reactor TFTR ,1 substantial evidence of interaction of IBWs with fast par- ticles in interacting with the IBW albeit that the beam ions were being heated as opposed to cooled , while wall in the tokamak fusion test reactor Daniel S. Clarka) and Nathaniel J. Fisch Princeton Plasma Physics Laboratory

  17. Neoclassical simulations of fusion alpha particles in pellet charge exchange experiments on the Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    on the Tokamak Fusion Test Reactor M. H. Redi a , S. H. Batha, M. G. Bell, R. V. Budny, D. S. Darrow, F. M on the Tokamak Fusion Test Reactor (TFTR) [Phys. Plas. 5 , 1577 (1998)] are found to be in good agreement code which includes the neoclassical transport processes, a recent first­principles model

  18. Neoclassical simulations of fusion alpha particles in pellet charge exchange experiments on the Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    on the Tokamak Fusion Test Reactor M. H. Redia , S. H. Batha, M. G. Bell, R. V. Budny, D. S. Darrow, F. M on the Tokamak Fusion Test Reactor (TFTR) [Phys. Plas. 5, 1577 (1998)] are found to be in good agreement code which includes the neoclassical transport processes, a recent first-principles model

  19. Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine

    SciTech Connect (OSTI)

    Reilly, Raymond W.

    2012-07-30

    This project, Development and Testing of a High Capacity Plasma Chemical Reactor in the Ukraine was established at the Kharkiv Institute of Physics and Technology (KIPT). The associated CRADA was established with Campbell Applied Physics (CAP) located in El Dorado Hills, California. This project extends an earlier project involving both CAP and KIPT conducted under a separate CRADA. The initial project developed the basic Plasma Chemical Reactor (PCR) for generation of ozone gas. This project built upon the technology developed in the first project, greatly enhancing the output of the PCR while also improving reliability and system control.

  20. Design and Testing of Vacuum Breaker Check Valve for Simplified Boiling Water Reactor

    SciTech Connect (OSTI)

    Ishii, M.; Xu, Y.; Revankar, S.T.

    2002-07-01

    A new design of the vacuum breaker check valve was developed to replace the mechanical valve in a simplified boiling water reactor. Scaling and design calculations were performed to obtain the geometry of new passive hydraulic vacuum breaker check valve. In order to check the valve performance, a RELAP5 model of the simplified boiling water reactor system with the new valve was developed. The valve was implemented in an integral facility, PUMA and was tested for large break loss of coolant accident. (authors)

  1. Nevada Test 1999 Waste Management Monitoring Report, Area 3 and Area 5 radioactive waste management sites

    SciTech Connect (OSTI)

    Yvonne Townsend

    2000-05-01

    Environmental monitoring data were collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site (NTS). These monitoring data include radiation exposure, air, groundwater, meteorology, vadose zone, and biota data. Although some of these media (radiation exposure, air, and groundwater) are reported in detail in other Bechtel Nevada reports (Annual Site Environmental Report [ASER], the National Emissions Standard for Hazardous Air Pollutants [NESHAP] report, and the Annual Groundwater Monitoring Report), they are also summarized in this report to provide an overall evaluation of RWMS performance and environmental compliance. Direct radiation monitoring data indicate that exposure at and around the RWMSs is not above background levels. Air monitoring data indicate that tritium concentrations are slightly above background levels, whereas radon concentrations are not above background levels. Groundwater monitoring data indicate that the groundwater in the alluvial aquifer beneath the Area 5 RWMS has not been affected by the facility. Meteorology data indicate that 1999 was a dry year: rainfall totaled 3.9 inches at the Area 3 RWMS (61 percent of average) and 3.8 inches at the Area 5 RWMS (75 percent of average). Vadose zone monitoring data indicate that 1999 rainfall infiltrated less than one foot before being returned to the atmosphere by evaporation. Soil-gas tritium data indicate very slow migration, and tritium concentrations in biota were insignificant. All 1999 monitoring data indicate that the Area 3 and Area 5 RWMSs are performing as expected at isolating buried waste.

  2. An X-Band Gun Test Area at SLAC

    SciTech Connect (OSTI)

    Limborg-Deprey, C.; Adolphsen, C.; Chu, T.S.; Dunning, M.P.; Jobe, R.K.; Jongewaard, E.N.; Hast, C.; Vlieks, A.E.; Wang, F.; Walz, D.R.; Marsh, R.A.; Anderson, S.G.; Hartemann, F.V.; Houck, T.L.; /LLNL, Livermore

    2012-09-07

    The X-Band Test Area (XTA) is being assembled in the NLCTA tunnel at SLAC to serve as a test facility for new RF guns. The first gun to be tested will be an upgraded version of the 5.6 cell, 200 MV/m peak field X-band gun designed at SLAC in 2003 for the Compton Scattering experiment run in ASTA. This new version includes some features implemented in 2006 on the LCLS gun such as racetrack couplers, increased mode separation and elliptical irises. These upgrades were developed in collaboration with LLNL since the same gun will be used in an injector for a LLNL Gamma-ray Source. Our beamline includes an X-band acceleration section which takes the electron beam up to 100 MeV and an electron beam measurement station. Other X-Band guns such as the UCLA Hybrid gun will be characterized at our facility.

  3. Fuel and core testing plan for a target fueled isotope production reactor.

    SciTech Connect (OSTI)

    Coats, Richard Lee; Dahl, James J.; Parma, Edward J., Jr.

    2010-12-01

    In recent years there has been an unstable supply of the critical diagnostic medical isotope 99Tc. Several concepts and designs have been proposed to produce 99Mo the parent nuclide of 99Tc, at a commercial scale sufficient to stabilize the world supply. This work lays out a testing and experiment plan for a proposed 2 MW open pool reactor fueled by Low Enriched Uranium (LEU) 99Mo targets. The experiments and tests necessary to support licensing of the reactor design are described and how these experiments and tests will help establish the safe operating envelop for a medical isotope production reactor is discussed. The experiments and tests will facilitate a focused and efficient licensing process in order to bring on line a needed production reactor dedicated to supplying medical isotopes. The Target Fuel Isotope Reactor (TFIR) design calls for an active core region that is approximately 40 cm in diameter and 40 cm in fuel height. It contains up to 150 cylindrical, 1-cm diameter, LEU oxide fuel pins clad with Zircaloy (zirconium alloy), in an annular hexagonal array on a {approx}2.0 cm pitch surrounded, radially, by a graphite or a Be reflector. The reactor is similar to U.S. university reactors in power, hardware, and safety/control systems. Fuel/target pin fabrication is based on existing light water reactor fuel fabrication processes. However, as part of licensing process, experiments must be conducted to confirm analytical predictions of steady-state power and accident conditions. The experiment and test plan will be conducted in phases and will utilize existing facilities at the U.S. Department of Energy's Sandia National Laboratories. The first phase is to validate the predicted reactor core neutronics at delayed critical, zero power and very low power. This will be accomplished by using the Sandia Critical Experiment (CX) platform. A full scale TFIR core will be built in the CX and delayed critical measurements will be taken. For low power experiments, fuel pins can be removed after the experiment and using Sandia's metrology lab, relative power profiles (radially and axially) can be determined. In addition to validating neutronic analyses, confirming heat transfer properties of the target/fuel pins and core will be conducted. Fuel/target pin power limits can be verified with out-of-pile (electrical heating) thermal-hydraulic experiments. This will yield data on the heat flux across the Zircaloy clad and establish safety margin and operating limits. Using Sandia's Annular Core Research Reactor (ACRR) a 4 MW TRIGA type research reactor, target/fuel pins can be driven to desired fission power levels for long durations. Post experiment inspection of the pins can be conducted in the Auxiliary Hot Cell Facility to observe changes in the mechanical properties of the LEU matrix and burn-up effects. Transient tests can also be conducted at the ACRR to observe target/fuel pin performance during accident conditions. Target/fuel pins will be placed in double experiment containment and driven by pulsing the ACRR until target/fuel failure is observed. This will allow for extrapolation of analytical work to confirm safety margins.

  4. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    SciTech Connect (OSTI)

    Agarwal, Vivek; Smith, James A.; Jewell, James Keith

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  5. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    SciTech Connect (OSTI)

    Lv, Quiping; Sun, Xiaodong; Chtistensen, Richard; Blue, Thomas; Yoder, Graydon; Wilson, Dane

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  6. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect (OSTI)

    Sheryl Morton; Carl Baily; Tom Hill; Jim Werner

    2006-02-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  7. Proceedings of the 1990 International Meeting on Reduced Enrichment for Research and Test Reactors

    SciTech Connect (OSTI)

    Not Available

    1993-07-01

    The global effort to reduce, and possibly, eliminate the international traffic in highly-enriched uranium caused by its use in research reactors requires extensive cooperation and free exchange of information among all participants. To foster this free exchange of information, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the thirteenth of a series which began in 1978. The common effort brought together, past, a large number of specialists from many countries. On hundred twenty-three participants from 26 countries, including scientists, reactor operators, and personnel from commercial fuel suppliers, research centers, and government organizations, convened in Newport, Rhode Island to discuss their results, their activities, and their plans relative to converting research reactors to low-enriched fuels. As more and more reactors convert to the use of low-enriched uranium, the emphasis of our effort has begun to shift from research and development to tasks more directly related to implementation of the new fuels and technologies that have been developed, and to refinements of those fuels and technologies. It is appropriate, for this reason, that the emphasis of this meeting was placed on safety and on conversion experiences. This individual papers in this report have been cataloged separately.

  8. Hanford 100-D Area Biostimulation Treatability Test Results

    SciTech Connect (OSTI)

    Truex, Michael J.; Vermeul, Vincent R.; Fritz, Brad G.; Mackley, Rob D.; Mendoza, Donaldo P.; Elmore, Rebecca P.; Mitroshkov, Alexandre V.; Sklarew, Deborah S.; Johnson, Christian D.; Oostrom, Martinus; Newcomer, Darrell R.; Brockman, Fred J.; Bilskis, Christina L.; Hubbard, Susan S.; Peterson, John E.; Williams, Kenneth H.; Gasperikova, E.; Ajo-Franklin, J.

    2009-09-30

    Pacific Northwest National Laboratory conducted a treatability test designed to demonstrate that in situ biostimulation can be applied to help meet cleanup goals in the Hanford Site 100-D Area. In situ biostimulation has been extensively researched and applied for aquifer remediation over the last 20 years for various contaminants. In situ biostimulation, in the context of this project, is the process of amending an aquifer with a substrate that induces growth and/or activity of indigenous bacteria for the purpose of inducing a desired reaction. For application at the 100-D Area, the purpose of biostimulation is to induce reduction of chromate, nitrate, and oxygen to remove these compounds from the groundwater. The in situ biostimulation technology is intended to provide supplemental treatment upgradient of the In Situ Redox Manipulation (ISRM) barrier previously installed in the Hanford 100-D Area and thereby increase the longevity of the ISRM barrier. Substrates for the treatability test were selected to provide information about two general approaches for establishing and maintaining an in situ permeable reactive barrier based on biological reactions, i.e., a biobarrier. These approaches included 1) use of a soluble (miscible) substrate that is relatively easy to distribute over a large areal extent, is inexpensive, and is expected to have moderate longevity; and 2) use of an immiscible substrate that can be distributed over a reasonable areal extent at a moderate cost and is expected to have increased longevity.

  9. INDEPENDENT CONFIRMATORY SURVEY REPORT FOR THE REACTOR BUILDING, HOT LABORATORY, PRIMARY PUMP HOUSE, AND LAND AREAS AT THE PLUM BROOK REACTOR FACILITY, SANDUSKY, OHIO

    SciTech Connect (OSTI)

    Erika N. Bailey

    2011-10-10

    In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics later called the National Aeronautics and Space Administration (NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventually built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities

  10. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental results show similar trends as the computational fluid dynamics (CFD) results presented in this report; however, some differences exist that will need to be assessed in future studies. The results of this testing will be used to improve the diode design to be tested in the liquid salt loop system.

  11. PEROXIDE DESTRUCTION TESTING FOR THE 200 AREA EFFLUENT TREATMENT FACILITY

    SciTech Connect (OSTI)

    HALGREN DL

    2010-03-12

    The hydrogen peroxide decomposer columns at the 200 Area Effluent Treatment Facility (ETF) have been taken out of service due to ongoing problems with particulate fines and poor destruction performance from the granular activated carbon (GAC) used in the columns. An alternative search was initiated and led to bench scale testing and then pilot scale testing. Based on the bench scale testing three manganese dioxide based catalysts were evaluated in the peroxide destruction pilot column installed at the 300 Area Treated Effluent Disposal Facility. The ten inch diameter, nine foot tall, clear polyvinyl chloride (PVC) column allowed for the same six foot catalyst bed depth as is in the existing ETF system. The flow rate to the column was controlled to evaluate the performance at the same superficial velocity (gpm/ft{sup 2}) as the full scale design flow and normal process flow. Each catalyst was evaluated on peroxide destruction performance and particulate fines capacity and carryover. Peroxide destruction was measured by hydrogen peroxide concentration analysis of samples taken before and after the column. The presence of fines in the column headspace and the discharge from carryover was generally assessed by visual observation. All three catalysts met the peroxide destruction criteria by achieving hydrogen peroxide discharge concentrations of less than 0.5 mg/L at the design flow with inlet peroxide concentrations greater than 100 mg/L. The Sud-Chemie T-2525 catalyst was markedly better in the minimization of fines and particle carryover. It is anticipated the T-2525 can be installed as a direct replacement for the GAC in the peroxide decomposer columns. Based on the results of the peroxide method development work the recommendation is to purchase the T-2525 catalyst and initially load one of the ETF decomposer columns for full scale testing.

  12. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    SciTech Connect (OSTI)

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  13. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Main report

    SciTech Connect (OSTI)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). The study results are presented in two volumes. Volume 1 (Main Report) contains the results in summary form.

  14. Closure Report for Corrective Action Unit 254: Area 25, R-MAD Decontamination Facility, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    G. N. Doyle

    2002-02-01

    Corrective Action Unit (CAU) 254 is located in Area 25 of the Nevada Test Site (NTS), approximately 100 kilometers (km) (62 miles) northwest of Las Vegas, Nevada. The site is located within the Reactor Maintenance, Assembly and Disassembly (R-MAD) compound and consists of Building 3126, two outdoor decontamination pads, and surrounding areas within an existing fenced area measuring approximately 50 x 37 meters (160 x 120 feet). The site was used from the early 1960s to the early 1970s as part of the Nuclear Rocket Development Station program to decontaminate test-car hardware and tooling. The site was reactivated in the early 1980s to decontaminate a radiologically contaminated military tank. This Closure Report (CR) describes the closure activities performed to allow un-restricted release of the R-MAD Decontamination Facility.

  15. Identification of Selected Areas to Support Federal Clean Energy Goals Using Small Modular Reactors

    SciTech Connect (OSTI)

    Belles, R. J. [ORNL; Mays, G. T. [ORNL; Omitaomu, O. A. [ORNL; Poore, W. P. [ORNL

    2013-12-30

    Beginning in late 2008, Oak Ridge National Laboratory (ORNL) responded to ongoing internal and external studies addressing key questions related to our national electrical energy supply. This effort has led to the development and refinement of Oak Ridge Siting Analysis for power Generation Expansion (OR-SAGE), a tool to support power plant siting evaluations. The objective in developing OR-SAGE was to use industry-accepted approaches and/or develop appropriate criteria for screening sites and employ an array of geographic information systems (GIS) data sources at ORNL to identify candidate areas for a power generation technology application. The basic premise requires the development of exclusionary, avoidance, and suitability criteria for evaluating sites for a given siting application, such as siting small modular reactors (SMRs). For specific applications of the tool, it is necessary to develop site selection and evaluation criteria (SSEC) that encompass a number of key benchmarks that essentially form the site environmental characterization for that application. These SSEC might include population density, seismic activity, proximity to water sources, proximity to hazardous facilities, avoidance of protected lands and floodplains, susceptibility to landslide hazards, and others.

  16. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    SciTech Connect (OSTI)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  17. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    SciTech Connect (OSTI)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace several obsolete components of the current analytical tool set used for ATR neutronics support. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). It will also greatly facilitate the LEU conversion effort, since the upgraded computational capabilities are now at a stage where they can be, and in fact have been, used for the required physics analysis from the beginning. In this context, extensive scoping neutronics analyses were completed for six preconceptual candidate LEU fuel element designs for the ATR (and for its companion critical facility, ATRC). Of these, four exhibited neutronics performance in what is believed to be an acceptable range. However, there are currently some concerns with regard to fabricability and mechanical performance that have emerged for one of the four latter concepts. Thus three concepts have been selected for more comprehensive conceptual design analysis during the upcoming fiscal year.

  18. Testing of Passive Safety System Performance for Higher Power Advanced Reactors

    SciTech Connect (OSTI)

    brian G. Woods; Jose Reyes, Jr.; John Woods; John Groome; Richard Wright

    2004-12-31

    This report describes the results of NERI research on the testing of advanced passive safety performance for the Westinghouse AP1000 design. The objectives of this research were: (a) to assess the AP1000 passive safety system core cooling performance under high decay power conditions for a spectrum of breaks located at a variety of locations, (b) to compare advanced thermal hydraulic computer code predictions to the APEX high decay power test data and (c) to develop new passive safety system concepts that could be used for Generation IV higher power reactors.

  19. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    Lumia, M E

    2002-01-01

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  20. Industrial Hygiene Concerns during the Decontamination and Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    M.E. Lumia; C.A. Gentile

    2002-01-18

    A significant industrial hygiene concern during the Decontamination and Decommissioning (D and D) of the Tokamak Fusion Test Reactor (TFTR) was the oxidation of the lead bricks' surface, which were utilized for radiation shielding. This presented both airborne exposure and surface contamination issues for the workers in the field removing this material. This paper will detail the various protection and control methods tested and implemented to protect the workers, including those technologies deployed to decontaminate the work surfaces. In addition, those techniques employed to recycle the lead for additional use at the site will be discussed.

  1. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L; Aaron, Adam M; Cunningham, Richard Burns; Fugate, David L; Holcomb, David Eugene; Kisner, Roger A; Peretz, Fred J; Robb, Kevin R; Wilgen, John B; Wilson, Dane F

    2014-01-01

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

  2. System Upgrades at the Advanced Test Reactor Help Ensure that Nuclear Energy Research Continues at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Craig Wise

    2011-12-01

    Fully operational in 1967, the Advanced Test Reactor (ATR) is a first-of-its-kind materials test reactor. Located on the Idaho National Laboratory’s desert site, this reactor remains at the forefront of nuclear science, producing extremely high neutron irradiation in a relatively short time span. The Advanced Test Reactor is also the only U.S. reactor that can replicate multiple reactor environments concurrently. The Idaho National Laboratory and the Department of Energy recently invested over 13 million dollars to replace three of ATR’s instrumentation and control systems. The new systems offer the latest software and technology advancements, ensuring the availability of the reactor for future energy research. Engineers and project managers successfully completed the four year project in March while the ATR was in a scheduled maintenance outage. “These new systems represent state-of-the-art monitoring and annunciation capabilities,” said Don Feldman, ATR Station Manager. “They are comparable to systems currently used for advanced reactor designs planned for construction in the U.S. and in operation in some foreign countries.”

  3. NUMERICAL SIMULATION FOR MECHANICAL BEHAVIOR OF U10MO MONOLITHIC MINIPLATES FOR RESEARCH AND TEST REACTORS

    SciTech Connect (OSTI)

    Hakan Ozaltun & Herman Shen

    2011-11-01

    This article presents assessment of the mechanical behavior of U-10wt% Mo (U10Mo) alloy based monolithic fuel plates subject to irradiation. Monolithic, plate-type fuel is a new fuel form being developed for research and test reactors to achieve higher uranium densities within the reactor core to allow the use of low-enriched uranium fuel in high-performance reactors. Identification of the stress/strain characteristics is important for understanding the in-reactor performance of these plate-type fuels. For this work, three distinct cases were considered: (1) fabrication induced residual stresses (2) thermal cycling of fabricated plates; and finally (3) transient mechanical behavior under actual operating conditions. Because the temperatures approach the melting temperature of the cladding during the fabrication and thermal cycling, high temperature material properties were incorporated to improve the accuracy. Once residual stress fields due to fabrication process were identified, solution was used as initial state for the subsequent simulations. For thermal cycling simulation, elasto-plastic material model with thermal creep was constructed and residual stresses caused by the fabrication process were included. For in-service simulation, coupled fluid-thermal-structural interaction was considered. First, temperature field on the plates was calculated and this field was used to compute the thermal stresses. For time dependent mechanical behavior, thermal creep of cladding, volumetric swelling and fission induced creep of the fuel foil were considered. The analysis showed that the stresses evolve very rapidly in the reactor. While swelling of the foil increases the stress of the foil, irradiation induced creep causes stress relaxation.

  4. Assessment of Feasibility of the Beneficial Use of Waste Heat from the Advanced Test Reactor

    SciTech Connect (OSTI)

    Donna P. Guillen

    2012-07-01

    This report investigates the feasibility of using waste heat from the Advanced Test Reactor (ATR). A proposed glycol waste heat recovery system was assessed for technical and economic feasibility. The system under consideration would use waste heat from the ATR secondary coolant system to preheat air for space heating of TRA-670. A tertiary coolant stream would be extracted from the secondary coolant system loop and pumped to a new plate and frame heat exchanger, where heat would be transferred to a glycol loop for preheating outdoor air in the heating and ventilation system. Historical data from Advanced Test Reactor operations over the past 10 years indicates that heat from the reactor coolant was available (when needed for heating) for 43.5% of the year on average. Potential energy cost savings by using the waste heat to preheat intake air is $242K/yr. Technical, safety, and logistics considerations of the glycol waste heat recovery system are outlined. Other opportunities for using waste heat and reducing water usage at ATR are considered.

  5. Structural analysis of fuel assembly clads for the Upgraded Transient Reactor Test Facility (TREAT Upgrade)

    SciTech Connect (OSTI)

    Ewing, T.F.; Wu, T.S.

    1986-01-01

    The Upgraded Transient Reactor Test Facility (TREAT Upgrade) is designed to test full-length, pre-irradiated fuel pins of the type used in large LMFBRs under accident conditions, such as severe transient overpower and loss-of-coolant accidents. In TREAT Upgrade, the central core region is to contain new fuel assemblies of higher fissile loadings to maximize the energy deposition to the test fuel. These fuel assemblies must withstand normal peak clad temperatures of 850/sup 0/C for hundreds of test transients. Due to high temperatures and gradients predicted in the clad, creep and plastic strain effects are significant, and the clad structural behavior cannot be analyzed by conventional linear techniques. Instead, the detailed elastic-plastic-creep behavior must be followed along the time-dependent load history. This paper presents details of the structural evaluations of the conceptual TREAT Upgrade fuel assembly clads.

  6. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  7. Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

  8. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  9. Vitrification testing of soil fines from contaminated Hanford 100 Area and 300 Area soils

    SciTech Connect (OSTI)

    Ludowise, J.D.

    1994-05-01

    The suitability of Hanford soil for vitrification is well known and has been demonstrated extensively in other work. The tests reported here were carried out to confirm the applicability of vitrification to the soil fines (a subset of the Hanford soil potentially different in composition from the bulk soil) and to provide data on the performance of actual, vitrified soil fines. It was determined that the soil fines were generally similar in composition to the bulk Hanford soil, although the fraction <0.25 mm in the 100 Area soil sample appears to differ somewhat from the bulk soil composition. The soil fines are readily melted into a homogeneous glass with the simple additions of CaO and/or Na{sub 2}O. The vitrified waste (plus additives) occupies only 60% of the volume of the initial untreated waste. Leach testing has shown the glasses made from the soil fines to be very durable relative to natural and man-made glasses and has demonstrated the ability of the vitrified waste to greatly reduce the release of radionuclides to the environment. Viscosity and electrical conductivity measurements indicate that the soil fines will be readily processable, although with levels of additives slightly greater than used in the radioactive melts. These tests demonstrate the applicability of vitrification to the contaminated soil fines and the exceptional performance of the waste form resulting from the vitrification of contaminated Hanford soils.

  10. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    SciTech Connect (OSTI)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2010-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850şC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 şC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm core characteristics, thermal-hydraulic properties, and radiation shielding. The high-temperature test operation at 950 şC represented the fifth and final phase of the rise-to-power tests. The safety tests demonstrated inherent safety features of the HTTR such as slow temperature response during abnormal events due to the large heat capacity of the core and the negative reactivity feedback. The experimental benchmark performed and currently evaluated in this report pertains to the data available for the annular core criticals from the initial six isothermal, annular and fully-loaded, core critical measurements performed at the HTTR. Evaluation of the start-up core physics tests specific to the fully-loaded core is compiled elsewhere (HTTR-GCR-RESR-001).

  11. CLOSURE REPORT FOR CORRECTIVE ACTION UNIT 115: AREA 25 TEST CELL A FACILITY, NEVADA TEST SITE, NEVADA

    SciTech Connect (OSTI)

    NA

    2006-03-01

    This Closure Report (CR) describes the activities performed to close CAU 115, Area 25 Test Cell A Facility, as presented in the NDEP-approved SAFER Plan (NNSA/NSO, 2004). The SAFER Plan includes a summary of the site history, process knowledge, and closure standards. This CR provides a summary of the completed closure activities, documentation of waste disposal, and analytical and radiological data to confirm that the remediation goals were met and to document final site conditions. The approved closure alternative as presented in the SAFER Plan for CAU 115 (NNSA/NSO, 2004) was clean closure; however, closure in place was implemented under a Record of Technical Change (ROTC) to the SAFER Plan when radiological surveys indicated that the concrete reactor pad was radiologically activated and could not be decontaminated to meet free release levels. The ROTC is included as Appendix G of this report. The objectives of closure were to remove any trapped residual liquids and gases, dispose regulated and hazardous waste, decontaminate removable radiological contamination, demolish and dispose aboveground structures, remove the dewar as a best management practice (BMP), and characterize and restrict access to all remaining radiological contamination. Radiological contaminants of concern (COCs) included cobalt-60, cesium-137, strontium-90, uranium-234/235/236/238, and plutonium-239/240. Additional COCs included Resource Conservation and Recovery Act (RCRA) metals, polychlorinated biphenyls (PCBs), and asbestos.

  12. Improved computational neutronics methods and validation protocols for the advanced test reactor

    SciTech Connect (OSTI)

    Nigg, D. W.; Nielsen, J. W.; Chase, B. M.; Murray, R. K.; Steuhm, K. A.; Unruh, T. [Idaho National Laboratory, 2525 Fremont Street, Idaho Falls, ID 83415-3870 (United States)

    2012-07-01

    The Idaho National Laboratory (INL) is in the process of updating the various reactor physics modeling and simulation tools used to support operation and safety assurance of the Advanced Test Reactor (ATR). Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purposes. On the experimental side of the project, new hardware was fabricated, measurement protocols were finalized, and the first four of six planned physics code validation experiments based on neutron activation spectrometry have been conducted at the ATRC facility. Data analysis for the first three experiments, focused on characterization of the neutron spectrum in one of the ATR flux traps, has been completed. The six experiments will ultimately form the basis for flexible and repeatable ATR physics code validation protocols that are consistent with applicable national standards. (authors)

  13. Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

    E-Print Network [OSTI]

    M. Osipenko; F. Pompili; M. Ripani; M. Pillon; G. Ricco; B. Caiffi; R. Cardarelli; G. Verona-Rinati; S. Argiro

    2015-05-23

    A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a $^6$Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based on conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of $10^8$ n/cm$^2$s and at the 3 MeV D-D monochromatic neutron source named FNG (ENEA, Rome) with neutron fluxes of $10^6$ n/cm$^2$s. The neutron spectrum measurement was performed at the TAPIRO fast research reactor (ENEA, Casaccia) with fluxes of 10$^9$ n/cm$^2$s. The obtained spectra were compared to Monte Carlo simulations, modeling detector response with MCNP and Geant4.

  14. Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop

    SciTech Connect (OSTI)

    Donna Post Guillen

    2012-11-01

    This study investigated the feasibility of using a waste heat recovery system (WHRS) to recover heat from the Advanced Test Reactor (ATR) secondary coolant system (SCS). This heat would be used to preheat air for space heating of the reactor building, thus reducing energy consumption, carbon footprint, and energy costs. Currently, the waste heat from the reactor is rejected to the atmosphere via a four-cell, induced-draft cooling tower. Potential energy and cost savings are 929 kW and $285K/yr. The WHRS would extract a tertiary coolant stream from the SCS loop and pump it to a new plate and frame heat exchanger, from which the heat would be transferred to a glycol loop for preheating outdoor air supplied to the heating and ventilation system. The use of glycol was proposed to avoid the freezing issues that plagued and ultimately caused the failure of a WHRS installed at the ATR in the 1980s. This study assessed the potential installation of a new WHRS for technical, logistical, and economic feasibility.

  15. Technology, safety, and costs of decommissioning reference nuclear research and test reactors. Appendices

    SciTech Connect (OSTI)

    Konzek, G.J.; Ludwick, J.D.; Kennedy, W.E. Jr.; Smith, R.I.

    1982-03-01

    Safety and Cost Information is developed for the conceptual decommissioning of two representative licensed nuclear research and test reactors. Three decommissioning alternatives are studied to obtain comparisons between costs (in 1981 dollars), occupational radiation doses, potential radiation dose to the public, and other safety impacts. The alternatives considered are: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and EMTOMB (entombment). The study results are presented in two volumes. Volume 2 (Appendices) contains the detailed data that support the results given in Volume 1, including unit-component data.

  16. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are extremely similar. The design of the experiment will be discussed followed by its progress and status to date.

  17. Monte-Carlo Code (MCNP) Modeling of the Advanced Test Reactor Applicable to the Mixed Oxide (MOX) Test Irradiation

    SciTech Connect (OSTI)

    G. S. Chang; R. C. Pederson

    2005-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, and 40 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). The fuel burnup analyses presented in this study were performed using MCWO, a welldeveloped tool that couples the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements.

  18. HWMA/RCRA CLOSURE PLAN FOR THE MATERIALS TEST REACTOR WING (TRA-604) LABORATORY COMPONENTS VOLUNTARY CONSENT ORDER ACTION PLAN VCO-5.8 D REVISION2

    SciTech Connect (OSTI)

    KIRK WINTERHOLLER

    2008-02-25

    This Hazardous Waste Management Act/Resource Conservation and Recovery Act closure plan was developed for the laboratory components of the Test Reactor Area Catch Tank System (TRA-630) that are located in the Materials Test Reactor Wing (TRA-604) at the Reactor Technology Complex, Idaho National Laboratory Site, to meet a further milestone established under Voluntary Consent Order Action Plan VCO-5.8.d. The TRA-604 laboratory components addressed in this closure plan were deferred from the TRA-630 Catch Tank System closure plan due to ongoing laboratory operations in the areas requiring closure actions. The TRA-604 laboratory components include the TRA-604 laboratory warm wastewater drain piping, undersink drains, subheaders, and the east TRA-604 laboratory drain header. Potentially contaminated surfaces located beneath the TRA-604 laboratory warm wastewater drain piping and beneath the island sinks located in Laboratories 126 and 128 (located in TRA-661) are also addressed in this closure plan. The TRA-604 laboratory components will be closed in accordance with the interim status requirements of the Hazardous Waste Management Act/Resource Conservation and Recovery Act as implemented by the Idaho Administrative Procedures Act 58.01.05.009 and 40 Code of Federal Regulations 265, Subparts G and J. This closure plan presents the closure performance standards and the methods for achieving those standards.

  19. Impact of new K Area geotechnical parameters on K Reactor restart response spectra

    SciTech Connect (OSTI)

    Amin, J.A.; Chen, R.C.; Mulliken, J.S.

    1991-08-01

    This report provides a description of a study performed to evaluate the impact of recently obtained soils parameters on the 105-K Reactor using Soil-Structure Interaction (SSI) analysis. This study includes: the frequency check of the reactor building based on the fixed base analysis; the modification of the computer model to reflect actual building properties and building frequencies; The live load impact on seismic analysis; The comparison of soil spectrum at elev. {minus}50 ft with 60% of RG1.60 curve; comparison of spectral results to the previous FREDA and SASSI results; and implications to the stack building analysis.

  20. Impact of new K Area geotechnical parameters on K Reactor restart response spectra. Seismic Structural Engineering

    SciTech Connect (OSTI)

    Amin, J.A.; Chen, R.C.; Mulliken, J.S.

    1991-08-01

    This report provides a description of a study performed to evaluate the impact of recently obtained soils parameters on the 105-K Reactor using Soil-Structure Interaction (SSI) analysis. This study includes: the frequency check of the reactor building based on the fixed base analysis; the modification of the computer model to reflect actual building properties and building frequencies; The live load impact on seismic analysis; The comparison of soil spectrum at elev. {minus}50 ft with 60% of RG1.60 curve; comparison of spectral results to the previous FREDA and SASSI results; and implications to the stack building analysis.

  1. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  2. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  3. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    SciTech Connect (OSTI)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  4. Recent palladium membrane reactor development at the tritium systems test assembly

    SciTech Connect (OSTI)

    Willms, R.S.; Birdsell, S.A.; Wilhelm, R.C.

    1995-07-01

    The palladium membrane reactor (PMR) is proving to be a simple and effective means for recovering hydrogen isotopes from fusion fuel impurities such as methane and water. This device directly combines two techniques which have long been utilized for hydrogen processing, namely catalytic shift reactions and palladium/silver permeators. A proof-of-principle (PMR) has been constructed and tested at the Tritium Systems Test Assembly of Los Alamos National Laboratory. The first tests with this device showed that is was effective for the proposed purpose. Initial work concluded that a nickel catalyst was an appropriate choice for use in a PMR. More detailed testing of the PMR with such a catalyst was performed and reported in other works. It was shown that a nickel catalyst-packed PMR did, indeed, recover hydrogen from water and methane with efficiencies approaching 100% in a single processing pass. These experiments were conducted over an extended period of time and no failure or need for regeneration was encountered. These positive results have prompted further PMR development. Topics addressed include alternate PMR geometries and initial testing of the PMR with tritium. These are the subjects of this paper.

  5. Nevada Test Site 2008 Waste Management Monitoring Report Area 3 and Area 5 Radioactive Waste Management Sites

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2009-06-23

    Environmental monitoring data were collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site. These data are associated with radiation exposure, air, groundwater, meteorology, vadose zone, subsidence, and biota. This report summarizes the 2008 environmental data to provide an overall evaluation of RWMS performance and to support environmental compliance and performance assessment (PA) activities.

  6. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    SciTech Connect (OSTI)

    Connaway, H. M.; Lee, C. H.

    2015-11-30

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  7. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    SciTech Connect (OSTI)

    David W. Nigg

    2013-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  8. Deuterium-Tritium Simulations of the Enhanced Reversed Shear Mode in the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    Mikkelsen, D.R.; Manickam, J.; Scott, S.D.; Zarnstorff

    1997-04-01

    The potential performance, in deuterium-tritium plasmas, of a new enhanced con nement regime with reversed magnetic shear (ERS mode) is assessed. The equilibrium conditions for an ERS mode plasma are estimated by solving the plasma transport equations using the thermal and particle dif- fusivities measured in a short duration ERS mode discharge in the Tokamak Fusion Test Reactor [F. M. Levinton, et al., Phys. Rev. Letters, 75, 4417, (1995)]. The plasma performance depends strongly on Zeff and neutral beam penetration to the core. The steady state projections typically have a central electron density of {approx}2:5x10 20 m{sup -3} and nearly equal central electron and ion temperatures of {approx}10 keV. In time dependent simulations the peak fusion power, {approx} 25 MW, is twice the steady state level. Peak performance occurs during the density rise when the central ion temperature is close to the optimal value of {approx} 15 keV. The simulated pressure profiles can be stable to ideal MHD instabilities with toroidal mode number n = 1, 2, 3, 4 and {infinity} for {beta}{sub norm} up to 2.5; the simulations have {beta}{sub norm} {le} 2.1. The enhanced reversed shear mode may thus provide an opportunity to conduct alpha physics experiments in conditions imilar to those proposed for advanced tokamak reactors.

  9. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    SciTech Connect (OSTI)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  10. Turkey Flat Site Effects Test Area The Turkey-Flat strong motion "blind"

    E-Print Network [OSTI]

    Oprsal, Ivo

    Turkey Flat Site Effects Test Area B B' A A' C C' The Turkey-Flat strong motion "blind" prediction Geological Survey Turkey Flat, USA Site Effects Test Area: "Blind" Test of Predicted Ground Response of a Shallow Stiff-Soil Site to the September 28, 2004 M6.0 Parkfield Earthquake Turkey Flat Working Group Stay

  11. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 113: Reactor Maintenance, Assembly, and Disassembly Building Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    J. L. Smith

    2001-01-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the action necessary for the closure in place of Corrective Action Unit (CAU) 113 Area 25 Reactor Maintenance, Assembly, and Disassembly Facility (R-MAD). CAU 113 is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) (NDEP, 1996). The CAU is located in Area 25 of the Nevada Test Site (NTS) and consists of Corrective Action Site (CAS) 25-04-01, R-MAD Facility (Figures 1-2). This plan provides the methodology for closure in place of CAU 113. The site contains radiologically impacted and hazardous material. Based on preassessment field work, there is sufficient process knowledge to close in place CAU 113 using the SAFER process. At a future date when funding becomes available, the R-MAD Building (25-3110) will be demolished and inaccessible radiologic waste will be properly disposed in the Area 3 Radiological Waste Management Site (RWMS).

  12. Recent palladium membrane reactor development at the tritium systems test assembly

    SciTech Connect (OSTI)

    Scott, W.R.; Birdsell, S.A.; Wilhelm, R.C. [Los Alamos National Lab., NM (United States)

    1995-10-01

    The palladium membrane reactor (PMR) is being investigated as a means for recovering hydrogen isotopes (including tritium) from compounds such as water and methane. Previous work with protiated water and methane showed that this device can be used to obtain high hydrogen recovery efficiencies using a single processing pass and with essentially no waste production. With these successful proof-of-principle results completed, recent work has focused on PMR development. This included studies of various geometries and testing with tritium. The results, which are reported here, have led to a better understanding of the PMR and will lead to the ultimate goal of building a production PMR and putting it into practical tritium processing service. 3 refs., 5 figs., 1 tab.

  13. Corrective Action Plan for Corrective Action Unit 262: Area 25 Septic Systems and Underground Discharge Point, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    K. B. Campbell

    2002-06-01

    This Corrective Action Plan (CAP) provides selected corrective action alternatives and proposes the closure methodology for Corrective Action Unit (CAU) 262, Area 25 Septic Systems and Underground Discharge Point. CAU 262 is identified in the Federal Facility Agreement and Consent Order (FFACO) of 1996. Remediation of CAU 262 is required under the FFACO. CAU 262 is located in Area 25 of the Nevada Test Site (NTS), approximately 100 kilometers (km) (62 miles [mi]) northwest of Las Vegas, Nevada. The nine Corrective Action Sites (CASs) within CAU 262 are located in the Nuclear Rocket Development Station complex. Individual CASs are located in the vicinity of the Reactor Maintenance, Assembly, and Disassembly (R-MAD); Engine Maintenance, Assembly, and Disassembly (E-MAD); and Test Cell C compounds. CAU 262 includes the following CASs as provided in the FFACO (1996); CAS 25-02-06, Underground Storage Tank; CAS 25-04-06, Septic Systems A and B; CAS 25-04-07, Septic System; CAS 25-05-03, Leachfield; CAS 25-05-05, Leachfield; CAS 25-05-06, Leachfield; CAS 25-05-08, Radioactive Leachfield; CAS 25-05-12, Leachfield; and CAS 25-51-01, Dry Well. Figures 2, 3, and 4 show the locations of the R-MAD, the E-MAD, and the Test Cell C CASs, respectively. The facilities within CAU 262 supported nuclear rocket reactor engine testing. Activities associated with the program were performed between 1958 and 1973. However, several other projects used the facilities after 1973. A significant quantity of radioactive and sanitary waste was produced during routine operations. Most of the radioactive waste was managed by disposal in the posted leachfields. Sanitary wastes were disposed in sanitary leachfields. Septic tanks, present at sanitary leachfields (i.e., CAS 25-02-06,2504-06 [Septic Systems A and B], 25-04-07, 25-05-05,25-05-12) allowed solids to settle out of suspension prior to entering the leachfield. Posted leachfields do not contain septic tanks. All CASs located in CAU 262 are inactive or abandoned. However, some leachfields may still receive liquids from runoff during storm events. Results from the 2000-2001 site characterization activities conducted by International Technology (IT) Corporation, Las Vegas Office are documented in the Corrective Action Investigation Report for Corrective Action Unit 262: Area 25 Septic Systems and Underground Discharge Point, Nevada Test Site, Nevada. This document is located in Appendix A of the Corrective Action Decision Document for CAU 262. Area 25 Septic Systems and Underground Discharge Point, Nevada Test Site, Nevada. (DOE/NV, 2001).

  14. Recent Progress of RF Cavity Study at Mucool Test Area

    SciTech Connect (OSTI)

    Yonehara, Katsuya; /Fermilab

    2011-12-02

    Summar of presentation is: (1) MTA is a multi task working space to investigate RF cavities for R&D of muon beam cooling channel - (a) Intense 400 MeV H{sup -} beam, (b) Handle hydrogen (flammable) gas, (c) 5 Tesla SC solenoid magnet, (d) He cryogenic/recycling system; (2) Pillbox cavity has been refurbished to search better RF material - Beryllium button test will be happened soon; (3) E x B effect has been tested in a box cavity - Under study (result seems not to be desirable); (4) 201 MHz RF cavity with SRF cavity treatment has been tested at low magnetic field - (a) Observed some B field effect on maximum field gradient and (b) Further study is needed (large bore SC magnet will be delivered end of 2011); and (5) HPRF cavity beam test has started - (a) No RF breakdown observed and (b) Design a new HPRF cavity to investigate more plasma loading effect.

  15. Corrective Action Plan for Corrective Action Unit 143: Area 25 Contaminated Waste Dumps, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    D. L. Gustafason

    2001-02-01

    This Corrective Action Plan (CAP) has been prepared for Corrective Action Unit (CAU) 143: Area 25 Contaminated Waste Dumps, Nevada Test Site, Nevada, in accordance with the Federal Facility Agreement and Consent Order of 1996. This CAP provides the methodology for implementing the approved corrective action alternative as listed in the Corrective Action Decision Document (U.S. Department of Energy, Nevada Operations Office, 2000). The CAU includes two Corrective Action Sites (CASs): 25-23-09, Contaminated Waste Dump Number 1; and 25-23-03, Contaminated Waste Dump Number 2. Investigation of CAU 143 was conducted in 1999. Analytes detected during the corrective action investigation were evaluated against preliminary action levels to determine constituents of concern for CAU 143. Radionuclide concentrations in disposal pit soil samples associated with the Reactor Maintenance, Assembly, and Disassembly Facility West Trenches, the Reactor Maintenance, Assembly, and Disassembly Facility East Trestle Pit, and the Engine Maintenance, Assembly, and Disassembly Facility Trench are greater than normal background concentrations. These constituents are identified as constituents of concern for their respective CASs. Closure-in-place with administrative controls involves use restrictions to minimize access and prevent unauthorized intrusive activities, earthwork to fill depressions to original grade, placing additional clean cover material over the previously filled portion of some of the trenches, and placing secondary or diversion berm around pertinent areas to divert storm water run-on potential.

  16. Weapons-Grade MOX Fuel Burnup Characteristics in Advanced Test Reactor Irradiation

    SciTech Connect (OSTI)

    G. S. Chang

    2006-07-01

    Mixed oxide (MOX) test capsules prepared with weapons-derived plutonium have been irradiated to a burnup of 50 GWd/t. The MOX fuel was fabricated at Los Alamos National Laboratory (LANL) by a master-mix process and has been irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Previous withdrawals of the same fuel have occurred at 9, 21, 30, 40, and 50 GWd/t. Oak Ridge National Laboratory (ORNL) manages this test series for the Department of Energy’s Fissile Materials Disposition Program (FMDP). A UNIX BASH (Bourne Again SHell) script CMO has been written and validated at the Idaho National Laboratory (INL) to couple the Monte Carlo transport code MCNP with the depletion and buildup code ORIGEN-2 (CMO). The new Monte Carlo burnup analysis methodology in this paper consists of MCNP coupling through CMO with ORIGEN-2(MCWO). MCWO is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code ORIGEN-2. The fuel burnup analyses presented in this study were performed using MCWO. MCWO analysis yields time-dependent and neutron-spectrum-dependent minor actinide and Pu concentrations for the ATR small I-irradiation test position. The purpose of this report is to validate both the Weapons-Grade Mixed Oxide (WG-MOX) test assembly model and the new fuel burnup analysis methodology by comparing the computed results against the neutron monitor measurements and the irradiated WG-MOX post irradiation examination (PIE) data.

  17. Marysville Test Well Geothermal Area | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIXsource HistoryScenariosMarysville Mt Geothermal Area Jump to: navigation,

  18. Infiltration Testing of Homes in the Houston Gulf Coast Area 

    E-Print Network [OSTI]

    Howel, E. S.

    1990-01-01

    Air Changes per Hour (ACH) are determined by construction and weather conditions. In this test, it was found that the ACH is equal to .59 plus .07 for every year of age the home has. The higher ACH values for age are due to deterioration...

  19. Effects of K-Reactor pre-operational cold flow testing on total suspended solids in Pen Branch

    SciTech Connect (OSTI)

    Wilde, E.W.

    1991-12-01

    Total suspended solids (TSS) levels were monitored by SRL Environmental Sciences personnel at two locations in the Pen Branch Creek system in conjunction with K Reactor cold flow (pump) testing required as part of the reactor restart effort. The TSS data were compared with flow and rainfall data collected simultaneously in an effort to obtain insight on the suspension and movement for particulate material in the Pen Branch system in response to natural and operational causes. Pump testing clearly caused higher TSS levels at the two sampling locations. The artificially elevated TSS levels were more pronounced at a sampling location near the reactor than at a sampling location farther downstream. Although the environmental data provided by this study were obtained and used exclusively for process control and research purposes, rather than for formal regulatory compliance (i.e. NPDES monitoring), the TSS levels determined by the comprehensive testing were compared with NPDES limits required at various SRS outfalls. TSS values in Pen Branch were seldom in excess of these limits. Because of the relatively few times that TSS values at the two sampling locations exceeded typical'' NPDES limits, and the fact that occasional relatively high TSS values could clearly be solely attributed to rainfall, it was concluded that no major adverse environmental impacts were caused to the Pen Branch system as a result of the K-Reactor pre-operational pump testing.

  20. Effects of K-Reactor pre-operational cold flow testing on total suspended solids in Pen Branch

    SciTech Connect (OSTI)

    Wilde, E.W.

    1991-12-01

    Total suspended solids (TSS) levels were monitored by SRL Environmental Sciences personnel at two locations in the Pen Branch Creek system in conjunction with K Reactor cold flow (pump) testing required as part of the reactor restart effort. The TSS data were compared with flow and rainfall data collected simultaneously in an effort to obtain insight on the suspension and movement for particulate material in the Pen Branch system in response to natural and operational causes. Pump testing clearly caused higher TSS levels at the two sampling locations. The artificially elevated TSS levels were more pronounced at a sampling location near the reactor than at a sampling location farther downstream. Although the environmental data provided by this study were obtained and used exclusively for process control and research purposes, rather than for formal regulatory compliance (i.e. NPDES monitoring), the TSS levels determined by the comprehensive testing were compared with NPDES limits required at various SRS outfalls. TSS values in Pen Branch were seldom in excess of these limits. Because of the relatively few times that TSS values at the two sampling locations exceeded ``typical`` NPDES limits, and the fact that occasional relatively high TSS values could clearly be solely attributed to rainfall, it was concluded that no major adverse environmental impacts were caused to the Pen Branch system as a result of the K-Reactor pre-operational pump testing.

  1. Flow Test At Alum Area (DOE GTP) | Open Energy Information

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    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX ECoopButtePowerEdistoWhiskey flatsInformation 7thFlorin, California: EnergyAlum Area

  2. Flow Test At Maui Area (DOE GTP) | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX ECoopButtePowerEdistoWhiskey flatsInformationFlint Geothermal AreaOpen EnergyMaui

  3. Flow Test At Wister Area (DOE GTP) | Open Energy Information

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    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX ECoopButtePowerEdistoWhiskey flatsInformationFlint GeothermalSilver PeakWister Area

  4. TEMPERATURE MONITORING OPTIONS AVAILABLE AT THE IDAHO NATIONAL LABORATORY ADVANCED TEST REACTOR

    SciTech Connect (OSTI)

    J.E. Daw; J.L. Rempe; D.L. Knudson; T. Unruh; B.M. Chase; K.L Davis

    2012-03-01

    As part of the Advanced Test Reactor National Scientific User Facility (ATR NSUF) program, the Idaho National Laboratory (INL) has developed in-house capabilities to fabricate, test, and qualify new and enhanced sensors for irradiation testing. To meet recent customer requests, an array of temperature monitoring options is now available to ATR users. The method selected is determined by test requirements and budget. Melt wires are the simplest and least expensive option for monitoring temperature. INL has recently verified the melting temperature of a collection of materials with melt temperatures ranging from 100 to 1000 C with a differential scanning calorimeter installed at INL’s High Temperature Test Laboratory (HTTL). INL encapsulates these melt wires in quartz or metal tubes. In the case of quartz tubes, multiple wires can be encapsulated in a single 1.6 mm diameter tube. The second option available to ATR users is a silicon carbide temperature monitor. The benefit of this option is that a single small monitor (typically 1 mm x 1 mm x 10 mm or 1 mm diameter x 10 mm length) can be used to detect peak irradiation temperatures ranging from 200 to 800 C. Equipment has been installed at INL’s HTTL to complete post-irradiation resistivity measurements on SiC monitors, a technique that has been found to yield the most accurate temperatures from these monitors. For instrumented tests, thermocouples may be used. In addition to Type-K and Type-N thermocouples, a High Temperature Irradiation Resistant ThermoCouple (HTIR-TC) was developed at the HTTL that contains commercially-available doped molybdenum paired with a niobium alloy thermoelements. Long duration high temperature tests, in furnaces and in the ATR and other MTRs, demonstrate that the HTIR-TC is accurate up to 1800 C and insensitive to thermal neutron interactions. Thus, degradation observed at temperatures above 1100 C with Type K and N thermocouples and decalibration due to transmutation with tungsten-rhenium and platinum rhodium thermocouples can be avoided. INL is also developing an Ultrasonic Thermometry (UT) capability. In addition to small size, UT’s offer several potential advantages over other temperature sensors. Measurements may be made near the melting point of the sensor material, potentially allowing monitoring of temperatures up to 3000 C. In addition, because no electrical insulation is required, shunting effects are avoided. Most attractive, however, is the ability to introduce acoustic discontinuities to the sensor, as this enables temperature measurements at several points along the sensor length. As discussed in this paper, the suite of temperature monitors offered by INL is not only available to ATR users, but also to users at other MTRs.

  5. Concepts and Tests for the Remote-Controlled Dismantling of the Biological Shield and Form work of the KNK Reactor - 13425

    SciTech Connect (OSTI)

    Neff, Sylvia; Graf, Anja; Petrick, Holger; Rothschmitt, Stefan; Klute, Stefan

    2013-07-01

    The compact sodium-cooled nuclear reactor facility Karlsruhe (KNK), a prototype Fast Breeder, is currently in an advanced stage of dismantling. Complete dismantling is based on 10 partial licensing steps. In the frame of the 9. decommissioning permit, which is currently ongoing, the dismantling of the biological shield is foreseen. The biological shield consists of heavy reinforced concrete with built-in steel fitments, such as form-work of the reactor tank, pipe sleeves, ventilation channels, and measuring devices. Due to the activation of the inner part of the biological shield, dismantling has to be done remote-controlled. During a comprehensive basic design phase a practical dismantling strategy was developed. Necessary equipment and tools were defined. Preliminary tests revealed that hot wire plasma cutting is the most favorable cutting technology due to the geometrical boundary conditions, the varying distance between cutter and material, and the heavy concrete behind the steel form-work. The cutting devices will be operated remotely via a carrier system with an industrial manipulator. The carrier system has expandable claws to adjust to the varying diameter of the reactor shaft during dismantling progress. For design approval of this prototype development, interaction between manipulator and hot wire plasma cutting was tested in a real configuration. For the demolition of the concrete structure, an excavator with appropriate tools, such as a hydraulic hammer, was selected. Other mechanical cutting devices, such as a grinder or rope saw, were eliminated because of concrete containing steel spheres added to increase the shielding factor of the heavy concrete. Dismantling of the biological shield will be done in a ring-wise manner due to static reasons. During the demolition process, the excavator is positioned on its tripod in three concrete recesses made prior to the dismantling of the separate concrete rings. The excavator and the manipulator carrier system will be operated alternately. Main boundary condition for all the newly designed equipment is the decommissioning housing of limited space within the reactor building containment. To allow for a continuous removal of the concrete rubble, an additional opening on the lowest level of the reactor shaft will be made. All equipment and the interaction of the tools have to be tested before use in the controlled area. Therefore a full-scale model of the biological shield will be provided in a mock-up. The tests will be performed in early 2014. The dismantling of the biological shield is scheduled for 2015. (authors)

  6. Nevada Test And Training Range Geothermal Area | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QAsource History ViewMayo, Maryland:NPI Ventures Ltd JumpNesjavellirInformationCertified TankNevada Test And

  7. Stress Test At Coso Geothermal Area (2004) | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION JEnvironmental Jump to:EA EIS ReportEurope GmbHSoloPageBeforeCreek Wind FarmStratton Middle-Test At

  8. Test study on the RPS of TMSR-SF1 reactor

    E-Print Network [OSTI]

    Liu, Zhenbao; Liu, Guimin; Hou, Jie

    2015-01-01

    The reactor protection system (RPS), as a 1E-level safety system, should be designed and developed following a series of nuclear laws and technical disciplines.

  9. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR). Metallic Fuels Program

    SciTech Connect (OSTI)

    Batte, G.L.; Pahl, R.G. [Argonne National Lab., Idaho Falls, ID (United States); Hofman, G.L. [Argonne National Lab., IL (United States)

    1993-09-01

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II.

  10. EVALUATION OF ZERO-POWER, ELEVATED-TEMPERATURE MEASUREMENTS AT JAPAN’S HIGH TEMPERATURE ENGINEERING TEST REACTOR

    SciTech Connect (OSTI)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2011-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The experimental benchmark performed and currently evaluated in this report pertains to the data available for two zero-power, warm-critical measurements with the fully-loaded HTTR core. Six isothermal temperature coefficients for the fully-loaded core from approximately 340 to 740 K have also been evaluated. These experiments were performed as part of the power-up tests (References 1 and 2). Evaluation of the start-up core physics tests specific to the fully-loaded core (HTTR-GCR-RESR-001) and annular start-up core loadings (HTTR-GCR-RESR-002) have been previously evaluated.

  11. Beam Test of a Large Area nonn Silicon Strip Detector with Fast Binary Readout Electronics

    E-Print Network [OSTI]

    Beam Test of a Large Area n­on­n Silicon Strip Detector with Fast Binary Readout Electronics Y test was carried out for the non­irradiated and the irradiated detector modules. Efficiency, noise occupancy and performance in the edge regions were analyzed using the beam test data. High efficiency

  12. Hydrologic resources management program and underground test area operable unit fy 1997

    SciTech Connect (OSTI)

    Smith, D. F., LLNL

    1998-05-01

    This report present the results of FY 1997 technical studies conducted by the Lawrence Livermore National Laboratory (LLNL) as part of the Hydrology and Radionuclide Migration Program (HRMP) and Underground Test Area Operable Unit (UGTA). The HRMP is sponsored by the US Department of Energy to assess the environmental (radiochemical and hydrologic) consequences of underground nuclear weapons testing at the Nevada Test Site.

  13. Beam Test of a Large Area nonn Silicon Strip Detector with Fast Binary Readout Electronics

    E-Print Network [OSTI]

    Beam Test of a Large Area n­on­n Silicon Strip Detector with Fast Binary Readout Electronics Y modules was irradiated with protons to a fluence of 1.2 × 10 14 p/cm 2 . A beam test was carried out in the edge regions were analyzed using the beam test data. High efficiency both for the non

  14. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014

    SciTech Connect (OSTI)

    Dan Ogden

    2014-10-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report October 2014 Highlights • Rory Kennedy, Dan Ogden and Brenden Heidrich traveled to Germantown October 6-7, for a review of the Infrastructure Management mission with Shane Johnson, Mike Worley, Bradley Williams and Alison Hahn from NE-4 and Mary McCune from NE-3. Heidrich briefed the group on the project progress from July to October 2014 as well as the planned path forward for FY15. • Jim Cole gave two invited university seminars at Ohio State University and University of Florida, providing an overview of NSUF including available capabilities and the process for accessing facilities through the peer reviewed proposal process. • Jim Cole and Rory Kennedy co-chaired the NuMat meeting with Todd Allen. The meeting, sponsored by Elsevier publishing, was held in Clearwater, Florida, and is considered one of the premier nuclear fuels and materials conferences. Over 340 delegates attended with 160 oral and over 200 posters presented over 4 days. • Thirty-one pre-applications were submitted for NSUF access through the NE-4 Combined Innovative Nuclear Research Funding Opportunity Announcement. • Fourteen proposals were received for the NSUF Rapid Turnaround Experiment Summer 2014 call. Proposal evaluations are underway. • John Jackson and Rory Kennedy attended the Nuclear Fuels Industry Research meeting. Jackson presented an overview of ongoing NSUF industry research.

  15. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  16. Safety Design Strategy for the Advanced Test Reactor Emergency Firewater Injection System Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  17. Safety Design Strategy for the Advanced Test Reactor Diesel Bus (E-3) and Switchgear Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  18. Safety Design Strategy for the Advanced Test Reactor Primary Coolant Pump and Motor Replacement Project

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01

    In accordance with the requirements of U.S. Department of Energy (DOE) Order 413.3B, “Program and Project Management for the Acquisition of Capital Assets,” safety must be integrated into the design process for new or major modifications to DOE Hazard Category 1, 2, and 3 nuclear facilities. The intended purpose of this requirement involves the handling of hazardous materials, both radiological and chemical, in a way that provides adequate protection to the public, workers, and the environment. Requirements provided in DOE Order 413.3B and DOE Order 420.1B, “Facility Safety,” and the expectations of DOE-STD-1189-2008, “Integration of Safety into the Design Process,” provide for identification of hazards early in the project and use of an integrated team approach to design safety into the facility. This safety design strategy provides the basic safety-in-design principles and concepts that will be used for the Advanced Test Reactor Reliability Sustainment Project. While this project does not introduce new hazards to the ATR, it has the potential for significant impacts to safety-related systems, structures, and components that are credited in the ATR safety basis and are being replaced. Thus the project has been determined to meet the definition of a major modification and is being managed accordingly.

  19. Nevada Test Site, 2006 Waste Management Monitoring Report, Area 3 and Area 5 Radioactive Waste Management Sites

    SciTech Connect (OSTI)

    David B. Hudson

    2007-06-30

    Environmental monitoring data were collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site. These data are associated with radiation exposure, air, groundwater, meteorology, vadose zone, subsidence, and biota. This report summarizes the 2006 environmental data to provide an overall evaluation of RWMS performance and to support environmental compliance and performance assessment (PA) activities. Some of these data (e.g., radiation exposure, air, and groundwater) are presented in other reports (U.S. Department of Energy, 2006; Warren and Grossman, 2007; National Security Technologies, LLC, 2007). Direct radiation monitoring data indicate that exposure levels around the RWMSs are at or below background levels. Air monitoring data at the Area 3 and Area 5 RWMSs indicate that tritium concentrations are slightly above background levels. There is no detectable man-made radioactivity by gamma spectroscopy, and concentrations of americium and plutonium are only slightly above detection limits at the Area 3 RWMS. Measurements at the Area 5 RWMS show that radon flux from waste covers is no higher than natural radon flux from undisturbed soil in Area 5. Groundwater monitoring data indicate that the groundwater in the uppermost aquifer beneath the Area 5 RWMS is not impacted by facility operations. Precipitation during 2006 totaled 98.6 millimeters (mm) (3.9 inches [in.]) at the Area 3 RWMS and 80.7 mm (3.2 in.) at the Area 5 RWMS. Soil-gas tritium monitoring continues to show slow subsurface migration consistent with previous results. Moisture from precipitation at Area 5 remains at the bottom of the bare-soil weighing lysimeter, but this same moisture has been removed from the vegetated weighing lysimeter by evapotranspiration. Vadose zone data from the operational waste pit covers show that evaporation continues to slowly remove soil moisture that came from the heavy precipitation in the fall of 2004 and the spring of 2005. The vegetated final cover at U-3ax/bl continues to remove moisture by evapotranspiration. There was no drainage through 2.4 meters (8 feet) of soil from the Area 3 drainage lysimeters that received only natural precipitation or were vegetated. Water drained from the bare-soil Area 3 drainage lysimeter that received three times natural precipitation. All 2006 monitoring data indicate that the Area 3 and Area 5 RWMSs are performing within expectations of the model and parameter assumptions for the facility PAs.

  20. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

    SciTech Connect (OSTI)

    Soelberg, Renae

    2014-11-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of the INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at CAES by Boise State PI Janelle Wharry and Cory Dolph. PI Corey Dolph returned in early November to complete their research by performing nanoindentation on unirradiated specimens that will be used as a baseline for their research.

  1. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    SciTech Connect (OSTI)

    David W. Nigg; Devin A. Steuhm

    2011-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system is being implemented and initial computational results have been obtained. This capability will have many applications in 2011 and beyond as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation. Finally we note that although full implementation of the new computational models and protocols will extend over a period 3-4 years as noted above, interim applications in the much nearer term have already been demonstrated. In particular, these demonstrations included an analysis that was useful for understanding the cause of some issues in December 2009 that were triggered by a larger than acceptable discrepancy between the measured excess core reactivity and a calculated value that was based on the legacy computational methods. As the Modeling Update project proceeds we anticipate further such interim, informal, applications in parallel with formal qualification of the system under the applicable INL Quality Assurance procedures and standards.

  2. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    SciTech Connect (OSTI)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009, Cycle 145A through Cycle 151B, was successfully completed during 2012. This major effort supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR Core Safety Analysis Package (CSAP) preparation process, in parallel with the established PDQ-based methodology, beginning late in Fiscal Year 2012. Acquisition of the advanced SERPENT (VTT-Finland) and MC21 (DOE-NR) Monte Carlo stochastic neutronics simulation codes was also initiated during the year and some initial applications of SERPENT to ATRC experiment analysis were demonstrated. These two new codes will offer significant additional capability, including the possibility of full-3D Monte Carlo fuel management support capabilities for the ATR at some point in the future. Finally, a capability for rigorous sensitivity analysis and uncertainty quantification based on the TSUNAMI system has been implemented and initial computational results have been obtained. This capability will have many applications as a tool for understanding the margins of uncertainty in the new models as well as for validation experiment design and interpretation.

  3. Results of Electric Survey in the Area of Hawaii Geothermal Test...

    Open Energy Info (EERE)

    Results of Electric Survey in the Area of Hawaii Geothermal Test Well HGP-A Jump to: navigation, search OpenEI Reference LibraryAdd to library Journal Article: Results of Electric...

  4. Nevada Test Site 2005 Waste Management Monitoring Report Area 3 and Area 5 Radioactive Waste Management Sites

    SciTech Connect (OSTI)

    David B. Hudson, Cathy A. Wills

    2006-08-01

    Environmental monitoring data were collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site. These data are associated with radiation exposure, air, groundwater, meteorology, vadose zone, subsidence, and biota. This report summarizes the 2005 environmental data to provide an overall evaluation of RWMS performance and to support environmental compliance and performance assessment activities. Some of these data (e.g., radiation exposure, air, and groundwater) are presented in other reports (U.S. Department of Energy, 2005; Grossman, 2005; Bechtel Nevada, 2006). Direct radiation monitoring data indicate that exposure levels around the RWMSs are at or below background levels. Air monitoring data at the Area 3 and Area 5 RWMSs indicate that tritium concentrations are slightly above background levels. There is no detectable man-made radioactivity by gamma spectroscopy, and concentrations of americium and plutonium are only slightly above detection limits at the Area 3 RWMS. Measurements at the Area 5 RWMS show that radon flux from waste covers is no higher than natural radon flux from undisturbed soil in Area 5. Groundwater monitoring data indicate that the groundwater in the uppermost aquifer beneath the Area 5 RWMS is not impacted by facility operations. Precipitation during 2005 totaled 219.1 millimeters (mm) (8.63 inches [in.]) at the Area 3 RWMS and 201.4 mm (7.93 in.) at the Area 5 RWMS. Soil-gas tritium monitoring continues to show slow subsurface migration consistent with previous results. Moisture from precipitation at Area 5 has percolated to the bottom of the bare-soil weighing lysimeter, but this same moisture has been removed from the vegetated weighing lysimeter by evapotranspiration. Vadose zone data from the operational waste pit covers show that precipitation from the fall of 2004 and the spring of 2005 infiltrated past the deepest sensors at 188 centimeters (6.2 feet) and remains in the pit cover. Precipitation did not infiltrate to the deepest sensor on the vegetated final cover at U-3ax/bl. Water drained from all Area 3 drainage lysimeters that received three times natural precipitation, but there was no drainage from the lysimeters that received only natural precipitation. Biota monitoring data show that tritium is the primary radionuclide accessible to plants and animals. Other human-produced radionuclides in the tissues of plant and animal samples from both RWMSs were not found at concentrations higher than in biota samples collected at control locations. This suggests that sampled animals did not intrude into the waste and that waste did not move to where it is accessible to plants or animals.

  5. Full-length U-xPu-10Zr (x=0, 8, 19 wt%) Fast Reactor Fuel Test in FFTF

    SciTech Connect (OSTI)

    D. L. Porter; H.C. Tsai

    2012-08-01

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt%) metallic fast reactor test with commercial-length (91.4 cm active fuel column length) conducted to date. With few remaining test reactors there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning of life (BOL) peak cladding temperature of the hottest pin was 608?C, cooling to 522?C at end of life (EOL). Selected fuel pins were examined non destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3 cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ~0.7 X/L axial location along the fuel column. This resulted from a lower production of rare earth fission products higher in the fuel column as well as a much smaller delta-T between fuel center and cladding, and therefore less FCCI, despite the higher cladding temperature. This behavior could actually help extend the life of a fuel pin in a “long pin” reactor design to a higher peak fuel burnup.

  6. Corrective Action Plan for Corrective Action Unit 424: Area 3 Landfill Complex, Tonopah Test Range, Nevada

    SciTech Connect (OSTI)

    Bechtel Nevada

    1998-08-31

    This corrective action plan provides the closure implementation methods for the Area 3 Landfill Complex, Corrective Action Unit (CAU) 424, located at the Tonopah Test Range. The Area 3 Landfill Complex consists of 8 landfill sites, each designated as a separate corrective action site.

  7. Joint Minimization of Power and Area in Scan Testing by Scan Cell Reordering Shalini Ghosh1

    E-Print Network [OSTI]

    Touba, Nur A.

    Joint Minimization of Power and Area in Scan Testing by Scan Cell Reordering Shalini Ghosh1 This paper describes a technique for re-ordering of scan cells to minimize power dissipation that is also capable of reducing the area overhead of the circuit compared to a random ordering of the scan cells

  8. Status of the NGNP graphite creep experiments AGC-1 and AGC-2 irradiated in the advanced test reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2014-05-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the next generation nuclear plant (NGNP) very high temperature gas reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have three different compressive loads applied to the top half of three diametrically opposite pairs of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment.

  9. Proposed Design and Operation of a Heat Pipe Reactor using the Sandia National Laboratories Annular Core Test Facility and Existing UZrH Fuel Pins

    SciTech Connect (OSTI)

    Wright, Steven A.; Lipinski, Ronald J.; Pandya, Tara; Peters, Curtis [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2005-02-06

    Heat Pipe Reactors (HPR) for space power conversion systems offer a number of advantages not easily provided by other systems. They require no pumping, their design easily deals with freezing and thawing of the liquid metal, and they can provide substantial levels of redundancy. Nevertheless, no reactor has ever been operated and cooled with heat pipes, and the startup and other operational characteristics of these systems remain largely unknown. Signification deviations from normal reactor heat removal mechanisms exist, because the heat pipes have fundamental heat removal limits due to sonic flow issues at low temperatures. This paper proposes an early prototypic test of a Heat Pipe Reactor (using existing 20% enriched nuclear fuel pins) to determine the operational characteristics of the HPR. The proposed design is similar in design to the HOMER and SAFE-300 HPR designs (Elliot, Lipinski, and Poston, 2003; Houts, et. al, 2003). However, this reactor uses existing UZrH fuel pins that are coupled to potassium heat pipes modules. The prototype reactor would be located in the Sandia Annular Core Research Reactor Facility where the fuel pins currently reside. The proposed reactor would use the heat pipes to transport the heat from the UZrH fuel pins to a water pool above the core, and the heat transport to the water pool would be controlled by adjusting the pressure and gas type within a small annulus around each heat pipe. The reactor would operate as a self-critical assembly at power levels up to 200 kWth. Because the nuclear heated HPR test uses existing fuel and because it would be performed in an existing facility with the appropriate safety authorization basis, the test could be performed rapidly and inexpensively. This approach makes it possible to validate the operation of a HPR and also measure the feedback mechanisms for a typical HPR design. A test of this nature would be the world's first operating Heat Pipe Reactor. This reactor is therefore called 'HPR-1'.

  10. Characterization Report for the 92-Acre Area of the Area 5 Radioactive Waste Management Site, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    Bechtel Nevada; U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office

    2006-06-01

    The U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office manages two low-level Radioactive Waste Management Sites at the Nevada Test Site. The Area 5 RWMS uses engineered shallow-land burial cells to dispose of packaged waste. This report summarizes characterization and monitoring work pertinent to the 92-Acre Area in the southeast part of the Area 5 Radioactive Waste Management Sites. The decades of characterization and assessment work at the Area 5 RWMS indicate that the access controls, waste operation practices, site design, final cover design, site setting, and arid natural environment contribute to a containment system that meets regulatory requirements and performance objectives for the short- and long-term protection of the environment and public. The available characterization and Performance Assessment information is adequate to support design of the final cover and development of closure plans. No further characterization is warranted to demonstrate regulatory compliance. U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office is proceeding with the development of closure plans for the six closure units of the 92-Acre Area.

  11. Nevada Test Site 2001 Waste Management Monitoring Report Area 3 and Area 5 Radioactive Waste Management Sites

    SciTech Connect (OSTI)

    Y. E. Townsend

    2002-06-01

    Environmental monitoring data, subsidence monitoring data, and meteorology monitoring data were collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site (NTS) (refer to Figure 1). These monitoring data include radiation exposure, air, groundwater, meteorology, vadose zone, subsidence, and biota data. Although some of these media (radiation exposure, air, and groundwater) are reported in detail in other Bechtel Nevada (BN) reports (Annual Site Environmental Report [ASER], the National Emissions Standard for Hazardous Air Pollutants [NESHAP] report, and the Annual Groundwater Monitoring Report), they are also summarized in this report to provide an overall evaluation of RWMS performance and environmental compliance. Direct radiation monitoring data indicate that exposure at and around the RWMSs is not above background levels. Air monitoring data indicate that tritium concentrations are slightly above background levels. Groundwater monitoring data indicate that the groundwater in the uppermost aquifer beneath the Area 5 RWMS has not been affected by the facility. Meteorology data indicate that 2001 was an average rainfall year: rainfall totaled 150 mm (5.9 in) at the Area 3 RWMS and 120 mm (4.7 in) at the Area 5 RWMS. Vadose zone monitoring data indicate that 2001 rainfall infiltrated less than one meter (3 ft) before being returned to the atmosphere by evaporation. Soil-gas tritium monitoring data indicate slow subsurface migration, and tritium concentrations in biota were lower than in previous years. All 2001 monitoring data indicate that the Area 3 and Area 5 RWMSs are performing within expectations of the model and parameter assumptions for the facility performance assessments.

  12. Nevada Test Site 2000 Waste Management Monitoring Report Area 3 and Area 5 Radioactive Waste Management Sites

    SciTech Connect (OSTI)

    Yvonne Townsend

    2001-06-01

    Environmental monitoring data, subsidence monitoring data, and meteorology monitoring data were collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site (NTS) (refer to Figure 1). These monitoring data include radiation exposure, air, groundwater, meteorology, vadose zone, subsidence, and biota data. Although some of these media (radiation exposure, air, and groundwater) are reported in detail in other Bechtel Nevada reports (Annual Site Environmental Report [ASER], the National Emissions Standard for Hazardous Air Pollutants [NESHAP] report, and the Annual Groundwater Monitoring Report), they are also summarized in this report to provide an overall evaluation of RWMS performance and environmental compliance. Direct radiation monitoring data indicate that exposure at and around the RWMSs is not above background levels. Air monitoring data indicate that tritium concentrations are slightly above background levels, whereas radon concentrations are not above background levels. Groundwater monitoring data indicate that the groundwater in the uppermost aquifer beneath the Area 5 RWMS has not been affected by the facility. Meteorology data indicate that 2000 was an average rainfall year: rainfall totaled 167 mm (6.6 in) at the Area 3 RWMS (annual average is 156 mm [6.5 in]) and 123 mm (4.8 in) at the Area 5 RWMS (annual average is 127 mm [5.0 in]). Vadose zone monitoring data indicate that 2000 rainfall infiltrated less than one meter (3 ft) before being returned to the atmosphere by evaporation. Soil-gas tritium monitoring data indicate slow subsurface migration, and tritium concentrations in biota were lower than in previous years. All 2000 monitoring data indicate that the Area 3 and Area 5 RWMSs are performing well at isolating buried waste.

  13. Private Water Well Testing in Areas Impacted by Marcellus Shale Gas Drilling

    E-Print Network [OSTI]

    Walter, M.Todd

    Private Water Well Testing in Areas Impacted by Marcellus Shale Gas Drilling (Updated May 10th in the absence of shale-gas drilling, well owners are strongly encouraged to evaluate their water on a regular testing in order to more specifically document potential impacts of Marcellus Shale gas development

  14. Nevada Test Site 2009 Waste Management Monitoring Report Area 3 and Area 5 Radioactive Waste Management Sites

    SciTech Connect (OSTI)

    NSTec Radioactive Waste

    2010-06-23

    Environmental monitoring data were collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site (NTS). These data are associated with radiation exposure, air, groundwater, meteorology, vadose zone, subsidence, and biota. This report summarizes the 2009 environmental data to provide an overall evaluation of RWMS performance and to support environmental compliance and performance assessment (PA) activities. Some of these data (e.g., radiation exposure, air, and groundwater) are presented in other reports. Direct radiation monitoring data indicate exposure levels at the RWMSs are within the range of background levels measured at the NTS. Air monitoring data at the Area 3 and Area 5 RWMSs indicate that tritium concentrations are slightly above background levels. All gamma spectroscopy results for air particulates collected at the Area 3 and Area 5 RWMS were below the minimum detectable concentrations, and concentrations of americium and plutonium are only slightly above detection limits. The measured levels of radionuclides in air particulates and moisture are below derived concentration guides for these radionuclides. Radon flux from waste covers is well below regulatory limits. Groundwater monitoring data indicate that the groundwater in the uppermost aquifer beneath the Area 5 RWMS is not impacted by facility operations. The 87.6 millimeters (mm) (3.45 inches [in.]) of precipitation at the Area 3 RWMS during 2009 is 43 percent below the average of 152.4 mm (6.00 in.), and the 62.7 mm (2.47 in.) of precipitation at the Area 5 RWMS during 2009 is 49 percent below the average of 122.5 mm (4.82 in.). Soil-gas tritium monitoring at borehole GCD-05 continues to show slow subsurface migration consistent with previous results. Water balance measurements indicate that evapotranspiration from the vegetated weighing lysimeter dries the soil and prevents downward percolation of precipitation more effectively than evaporation from the bare-soil weighing lysimeter. Data from the automated vadose zone monitoring system for the operational waste pit covers show that moisture from precipitation did not percolate below 90 centimeters (cm) (3 feet [ft]) before being removed by evaporation. Moisture from precipitation did not percolate below 30 cm (1 ft) in the vegetated final mono-layer cover on the U-3ax/bl disposal unit at the Area 3 RWMS before being removed by evapotranspiration. During 2009, there was no drainage through 2.4 meters (8 ft) of soil from the Area 3 drainage lysimeters that received only natural precipitation or were vegetated, but water drained from the bare-soil Area 3 drainage lysimeter that received 3 times natural precipitation. Elevated tritium levels in plants and animals sampled from the Area 3 and Area 5 RWMSs show tritium uptake by the biota, but the low levels of other radionuclides do not indicate that there has been biota intrusion into the waste. All 2009 monitoring data indicate that the Area 3 and Area 5 RWMSs are performing within expectations of the model and parameter assumptions for the facility PAs.

  15. Nevada Test Site 2007 Waste Management Monitoring Report Area 3 and Area 5 Radioactive Waste Management Sites

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2008-06-01

    Environmental monitoring data were collected at and around the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site. These data are associated with radiation exposure, air, groundwater, meteorology, vadose zone, subsidence, and biota. This report summarizes the 2007 environmental data to provide an overall evaluation of RWMS performance and to support environmental compliance and performance assessment (PA) activities. Some of these data (e.g., radiation exposure, air, and groundwater) are presented in other reports (National Security Technologies, LLC, 2007a; 2008; Warren and Grossman, 2008). Direct radiation monitoring data indicate exposure levels at the RWMSs are at background levels. Air monitoring data at the Area 3 and Area 5 RWMSs indicate that tritium concentrations are slightly above background levels. A single gamma spectroscopy measurement for cesium was slightly above the minimum detectable concentration, and concentrations of americium and plutonium are only slightly above detection limits at the Area 3 RWMS. The measured levels of radionuclides in air particulates are below derived concentration guides for these radionuclides. Radon flux from waste covers is well below regulatory limits. Groundwater monitoring data indicate that the groundwater in the uppermost aquifer beneath the Area 5 RWMS is not impacted by facility operations. The 136.8 millimeters (mm) (5.39 inches [in.]) of precipitation at the Area 3 RWMS during 2007 is 13 percent below the average of 158.1 mm (6.22 in.), and the 123.8 mm (4.87 in.) of precipitation at the Area 5 RWMS during 2007 is 6 percent below the average of 130.7 mm (5.15 in.). Soil-gas tritium monitoring at borehole GCD-05U continues to show slow subsurface migration consistent with previous results. Water balance measurements indicate that evapotranspiration from the vegetated weighing lysimeter dries the soil and prevents downward movement percolation of precipitation more effectively than evaporation from the bare-soil weighing lysimeter. Data from the automated vadose zone monitoring system for the operational waste pit covers show that evaporation continues to slowly remove soil moisture that came from the heavy precipitation in the fall of 2004 and the spring of 2005. The vegetated final mono-layer cover on the U-3ax/bl disposal unit at the Area 3 RWMS effectively removes moisture from the cover by evapotranspiration. During 2007, there was no drainage through 2.4 meters (8 feet) of soil from the Area 3 drainage lysimeters that received only natural precipitation or were vegetated but water drained from the bare-soil Area 3 drainage lysimeter that received 3 times natural precipitation. Elevated tritium levels in plants and animals sampled from the Area 3 and Area 5 RWMSs show tritium uptake by the biota, but the low levels of other radionuclides do not suggest that there has been intrusion into the waste. All 2007 monitoring data indicate that the Area 3 and Area 5 RWMSs are performing within expectations of the model and parameter assumptions for the facility PAs.

  16. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect (OSTI)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

  17. MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11

    The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

  18. AGR-2: The first irradiation of French HTR fuel in Advanced Test Reactor

    SciTech Connect (OSTI)

    T. Lambert; B. Grover; P. Guillermier; D. Moulinier; F. Imbault Huart

    2012-10-01

    AGR-2, the second irradiation of the US program for qualification of the NGNP fuel, is open to international participation within the scope of the Generation IV International Forum. In this frame, it includes in its multi-capsule irradiation rig an irradiation of French HTR fuel manufactured in the CAPRI line (GAIA facility at CEA/Cadarache and AREVA/CERCA compacting line at Romans). The AGR-2 irradiation is designed to place our first fabrications of HTR particles under operating conditions that are representative of ANTARES project while keeping close to the test range of the German fuel as much as possible, which is the reference in terms of irradiation behavior. A few batches of particles and 12 fuel compacts were produced and characterized in 2009 by CEA and CERCA. The fuel main characteristics are in conformity with our specifications and in compliance with INL requirements. The AGR-2 experiment is based on the design and devices used in the first experiment of the AGR program. The design makes it possible to monitor the irradiation conditions and in particular, the temperature, the power and the fission products released from fuel particles. The in pile equipment consists of a multi-capsule device designed to simultaneously irradiate six independent capsules with temperature control. The out-of-core part consists of the equipment for actively controlling temperature and measuring the fission products release on-line. The target conditions for the irradiation experiment were defined with the aim of comparing the results obtained under irradiation with German particles along with the objectives of reaching burn-up and fluence targets to validate the behavior of our fuel in a significant range (15% FIMA – 5 × 1025 n/m2 at 600 EFPD with centerline fuel temperature about 1100 degrees C). These conditions have to be representative of ANTARES project characteristics. These target conditions were compared with final results from neutron and thermal design studies performed by INL team, and preliminary thermal mechanical ATLAS calculations were carried out by CEA from this pre-design. Despite the mean burn-up achieved in approximately 600 EFPD being a little high (16.3% FIMA max. associated with a low fluence up to 2.85 × 1025 n/m2), this irradiation will nevertheless encompass the range of irradiation effects covered in our experimental objectives (maximum stress peak at start of irradiation then sign inversion of the stress in the SiC layer). In addition, the fluence and burn-up acceleration factors are very similar to those of the German reference experiments. This experimental irradiation began in July 2010 in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) and first results have been acquired.

  19. 10 CFR 830 Major Modification Determination for Advanced Test Reactor RDAS and LPCIS Replacement

    SciTech Connect (OSTI)

    David E. Korns

    2012-05-01

    The replacement of the ATR Control Complex's obsolete computer based Reactor Data Acquisition System (RDAS) and its safety-related Lobe Power Calculation and Indication System (LPCIS) software application is vitally important to ensure the ATR remains available to support this national mission. The RDAS supports safe operation of the reactor by providing 'real-time' plant status information (indications and alarms) for use by the reactor operators via the Console Display System (CDS). The RDAS is a computer support system that acquires analog and digital information from various reactor and reactor support systems. The RDAS information is used to display quadrant and lobe powers via a display interface more user friendly than that provided by the recorders and the Control Room upright panels. RDAS provides input to the Nuclear Engineering ATR Surveillance Data System (ASUDAS) for fuel burn-up analysis and the production of cycle data for experiment sponsors and the generation of the Core Safety Assurance Package (CSAP). RDAS also archives and provides for retrieval of historical plant data which may be used for event reconstruction, data analysis, training and safety analysis. The RDAS, LPCIS and ASUDAS need to be replaced with state-of-the-art technology in order to eliminate problems of aged computer systems, and difficulty in obtaining software upgrades, spare parts, and technical support. The major modification criteria evaluation of the project design did not lead to the conclusion that the project is a major modification. The negative major modification determination is driven by the fact that the project requires a one-for-one equivalent replacement of existing systems that protects and maintains functional and operational requirements as credited in the safety basis.

  20. June 2012 Groundwater Sampling at the Central Nevada Test Area (Data Validation Package)

    SciTech Connect (OSTI)

    2013-03-01

    The U.S. Department of Energy Office of Legacy Management conducted annual sampling at the Central Nevada Test Area (CNTA) on June 26-27, 2012, in accordance with the 2004 Correction Action Decision Document/Corrective Action Plan for Corrective Action Unit 443: Central Nevada Test Area (CNTA)-Subsurface and the addendum to the "Corrective Action Decision Document/Corrective Action Plan" completed in 2008. Sampling and analysis were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PLN/S04351), continually updated).

  1. Underground Test Area Subproject Phase I Data Analysis Task. Volume VII - Tritium Transport Model Documentation Package

    SciTech Connect (OSTI)

    1996-12-01

    Volume VII of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the tritium transport model documentation. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  2. May 2011 Groundwater Sampling at the Central Nevada Test Area (Data Validation Package)

    SciTech Connect (OSTI)

    None

    2011-11-01

    The U.S. Department of Energy Office of Legacy Management conducted annual sampling at the Central Nevada Test Area (CNTA) on May 10-11, 2011, in accordance with the 2004 Correction Action Decision Document/Corrective Action Plan for Corrective Action Unit 443: Central Nevada Test Area (CNTA)-Subsurface and the addendum to the "Corrective Action Decision Document/Corrective Action Plan" completed in 2008. Sampling and analysis were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PLN/S04351), continually updated).

  3. May 2010 Groundwater Sampling at the Central Nevada Test Area (Data Validation Package)

    SciTech Connect (OSTI)

    None

    2011-02-01

    The U.S. Department of Energy Office of Legacy Management conducted annual sampling at the Central Nevada Test Area (CNTA) on June 7-9, 2010, in accordance with the 2004 Correction Action Decision Document/Corrective Action Plan for Corrective Action Unit 443: Central Nevada Test Area (CNTA)-Subsurface. Sampling and analysis were conducted as specified in the Sampling and Analysis Plan for U.S. Department of Energy Office of Legacy Management Sites (LMS/PLN/S04351), continually updated).

  4. Underground Test Area Subproject Phase I Data Analysis Task. Volume II - Potentiometric Data Document Package

    SciTech Connect (OSTI)

    1996-12-01

    Volume II of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the potentiometric data. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  5. Leakage Tests of the Stainless Steel Vessels of the Antineutrino Detectors in the Daya Bay Reactor Neutrino Experiment

    E-Print Network [OSTI]

    Xiaohui Chen; Xiaolan Luo; Yuekun Heng; Lingshu Wang; Xiao Tang; Xiaoyan Ma; Honglin Zhuang; Henry Band; Jeff Cherwinka; Qiang Xiao; Karsten M. Heeger

    2012-03-02

    The antineutrino detectors in the Daya Bay reactor neutrino experiment are liquid scintillator detectors designed to detect low energy particles from antineutrino interactions with high efficiency and low backgrounds. Since the antineutrino detector will be installed in a water Cherenkov cosmic ray veto detector and will run for 3 to 5 years, ensuring water tightness is critical to the successful operation of the antineutrino detectors. We choose a special method to seal the detector. Three leak checking methods have been employed to ensure the seal quality. This paper will describe the sealing method and leak testing results.

  6. A New Interpretation of Alpha-particle-driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    R. Nazikian; G.J. Kramer; C.Z. Cheng; N.N. Gorelenkov; H.L. Berk; S.E. Sharapov

    2003-03-26

    The original description of alpha-particle-driven instabilities in the Tokamak Fusion Test Reactor (TFTR) in terms of Toroidal Alfvin Eigenmodes (TAEs) remained inconsistent with three fundamental characteristics of the observations: (i) the variation of the mode frequency with toroidal mode number, (ii) the chirping of the mode frequency for a given toroidal mode number, and (iii) the anti-ballooning density perturbation of the modes. It is now shown that these characteristics can be explained by observing that cylindrical-like modes can exist in the weak magnetic shear region of the plasma that then make a transition to TAEs as the central safety factor decreases in time.

  7. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are very similar. The purpose and design of this experiment will be discussed followed by its progress and status to date.

  8. Status of the NGNP Graphite Creep Experiments AGC-1 and AGC-2 Irradiated in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six nuclear graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six peripheral stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six peripheral stacks will have different compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during irradiation of the experiment. The first experiment, AGC-1, started its irradiation in September 2009, and the irradiation was completed in January 2011. The second experiment, AGC-2, started its irradiation in April 2011 and completed its irradiation in May 2012. This paper will briefly discuss the design of the experiment and control systems, and then present the irradiation results for each experiment to date.

  9. Radiation shielding issues for MuCool test area at Fermilab

    SciTech Connect (OSTI)

    Rakhno, I.; Johnstone, C.; /Fermilab

    2005-03-01

    The MuCool Test Area (MTA) is an intense primary beam facility derived directly from the Fermilab Linac to test heat deposition and other technical concerns associated with the liquid hydrogen targets being developed for cooling intense muon beams. In this study the origin of the outgoing collimated neutron beam is examined. An alternative shielding option for MTA is investigated as well as the hypothetical worst case of experimental setup is considered.

  10. Frequency response testing at Experimental Breeder Reactor II using discrete-level periodic signals

    SciTech Connect (OSTI)

    Rhodes, W.D.; Larson, H.A. (Idaho State Univ., Pocatello, ID (USA). Coll. of Engineering); Dean, E.M. (Argonne National Lab., Idaho Falls, ID (USA))

    1990-01-01

    The Experimental Breeder Reactor 2 (EBR-2) reactivity-to-power frequency-response function was measured with pseudo-random, discrete-level, periodic signals. The reactor power deviation was small with insignificant perturbation of normal operation and in-place irradiation experiments. Comparison of results with measured rod oscillator data and with theoretical predictions show good agreement. Moreover, measures of input signal quality (autocorrelation function and energy spectra) confirm the ability to enable this type of frequency response determination at EBR-2. Measurements were made with the pseudo-random binary sequence, quadratic residue binary sequence, pseudo-random ternary sequence, and the multifrequency binary sequence. 10 refs., 7 figs., 3 tabs.

  11. Corrective action investigation plan for Central Nevada Test Area, CAU No. 417

    SciTech Connect (OSTI)

    NONE

    1997-04-01

    This Corrective Action Investigation Plan (CAIP) is part of a US Department of Energy (DOE)-funded environmental investigation of the Central Nevada Test Area (CNTA). This CAIP addresses the surface investigation and characterization of 15 identified Corrective Action Sites (CASs). In addition, several other areas of the CNTA project area have surface expressions that may warrant investigation. These suspect areas will be characterized, if necessary, in subsequent CAIPs or addendums to this CAIP prepared to address these sites. This CAIP addresses only the 15 identified CASs as shown in Table 2-1 that are associated with the drilling and construction of a number of testing wells designed as part of an underground nuclear testing program. The purpose of the wells at the time of construction was to provide subsurface access for the emplacement, testing, and post detonation evaluations of underground nuclear devices. If contamination is found at any of the 15-surface CASs, the extent of contamination will be determined in order to develop an appropriate corrective action.

  12. Geology Report: Area 3 Radioactive Waste Management Site DOE/Nevada Test Site, Nye County, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2006-07-01

    Surficial geologic studies near the Area 3 Radioactive Waste Management Site (RWMS) were conducted as part of a site characterization program. Studies included evaluation of the potential for future volcanism and Area 3 fault activity that could impact waste disposal operations at the Area 3 RWMS. Future volcanic activity could lead to disruption of the Area 3 RWMS. Local and regional studies of volcanic risk indicate that major changes in regional volcanic activity within the next 1,000 years are not likely. Mapped basalts of Paiute Ridge, Nye Canyon, and nearby Scarp Canyon are Miocene in age. There is a lack of evidence for post-Miocene volcanism in the subsurface of Yucca Flat, and the hazard of basaltic volcanism at the Area 3 RWMS, within the 1,000-year regulatory period, is very low and not a forseeable future event. Studies included a literature review and data analysis to evaluate unclassified published and unpublished information regarding the Area 3 and East Branch Area 3 faults mapped in Area 3 and southern Area 7. Two trenches were excavated along the Area 3 fault to search for evidence of near-surface movement prior to nuclear testing. Allostratigraphic units and fractures were mapped in Trenches ST02 and ST03. The Area 3 fault is a plane of weakness that has undergone strain resulting from stress imposed by natural events and underground nuclear testing. No major vertical displacement on the Area 3 fault since the Early Holocene, and probably since the Middle Pleistocene, can be demonstrated. The lack of major displacement within this time frame and minimal vertical extent of minor fractures suggest that waste disposal operations at the Area 3 RWMS will not be impacted substantially by the Area 3 fault, within the regulatory compliance period. A geomorphic surface map of Yucca Flat utilizes the recent geomorphology and soil characterization work done in adjacent northern Frenchman Flat. The approach taken was to adopt the map unit boundaries (line work) of Swadley and Hoover (1990) and re-label these with map unit designations like those in northern Frenchman Flat (Huckins-Gang et al, 1995a,b,c; Snyder et al, 1995a,b,c,d).

  13. Underground Test Area Quality Assurance Project Plan Nevada National Security Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Irene Farnham

    2011-05-01

    This Quality Assurance Project Plan (QAPP) provides the overall quality assurance (QA) program requirements and general quality practices to be applied to the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Underground Test Area (UGTA) Sub-Project (hereafter the Sub-Project) activities. The requirements in this QAPP are consistent with DOE Order 414.1C, Quality Assurance (DOE, 2005); U.S. Environmental Protection Agency (EPA) Guidance for Quality Assurance Project Plans for Modeling (EPA, 2002); and EPA Guidance on the Development, Evaluation, and Application of Environmental Models (EPA, 2009). The QAPP Revision 0 supersedes DOE--341, Underground Test Area Quality Assurance Project Plan, Nevada Test Site, Nevada, Revision 4.

  14. Halden In-Reactor Test to Exhibit PWR Axial Offset Anomaly

    SciTech Connect (OSTI)

    P.Bennett, B. Beverskog, R.Suther

    2004-12-01

    Many PWRs have encountered the axial offset anomaly (AOA) since the early 1990s, and these experiences have been reported widely. AOA is a phenomenon associated with localized boron hideout in corrosion product deposits (crud) on fuel surfaces. Several mitigation approaches have been developed or are underway to either delay the onset of AOA or avoid it entirely. This study describes the first phase of an experimental program designed to investigate whether the use of enriched boric acid (EBA) in the reactor coolant can mitigate AOA.

  15. Site characterization and monitoring data from Area 5 Pilot Wells, Nevada Test Site, Nye County, Nevada

    SciTech Connect (OSTI)

    NONE

    1994-02-01

    The Special Projects Section (SPS) of Reynolds Electrical & Engineering Co., Inc. (REECO) is responsible for characterizing the subsurface geology and hydrology of the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site (NTS) for the US Department of Energy, Nevada Operations Office (DOE/NV), Environmental Restoration and Waste Management Division, Waste Operations Branch. The three Pilot Wells that comprise the Pilot Well Project are an important part of the Area 5 Site Characterization Program designed to determine the suitability of the Area 5 RWMS for disposal of low-level waste (LLW), mixed waste (MW), and transuranic waste (TRU). The primary purpose of the Pilot Well Project is two-fold: first, to characterize important water quality and hydrologic properties of the uppermost aquifer; and second, to characterize the lithologic, stratigraphic, and hydrologic conditions which influence infiltration, redistribution, and percolation, and chemical transport through the thick vadose zone in the vicinity of the Area 5 RWMS. This report describes Pilot Well drilling and coring, geophysical logging, instrumentation and stemming, laboratory testing, and in situ testing and monitoring activities.

  16. Closure Strategy Nevada Test Site Area 5 Radioactive Waste Management Site

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2007-03-01

    This paper presents an overview of the strategy for closure of part of the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site (NTS), which is about 65 miles northwest of Las Vegas, Nevada (Figure 1). The Area 5 RWMS is in the northern part of Frenchman Flat, approximately 14 miles north of Mercury. The Area 5 RWMS encompasses 732 acres subdivided into quadrants, and is bounded by a 1,000-foot (ft)-wide buffer zone. The northwest and southwest quadrants have not been developed. The northeast and southeast quadrants have been used for disposal of unclassified low-level radioactive waste (LLW) and indefinite storage of classified materials. This paper focuses on closure of the 38 waste disposal and classified material storage units within the southeast quadrant of the Area 5 RWMS, called the ''92-Acre Area''. The U.S Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) is currently planning to close the 92-Acre Area by 2011. Closure planning for this site must take into account the regulatory requirements for a diversity of waste streams, disposal and storage configurations, disposal history, and site conditions. For ease of discussion, the 92-Acre Area has been subdivided into six closure units defined by waste type, location, and similarity in regulatory requirements. Each of the closure units contains one or more waste disposal units; waste disposal units are also called waste disposal cells. The paper provides a brief background of the Area 5 RWMS, identifies key closure issues for the 92-Acre Area, recommends actions to address the issues, and provides the National Security Technologies, LLC (NSTec), schedule for closure.

  17. Direct Test of the Time-Independence of Fundamental Nuclear Constants Using the Oklo Natural Reactor

    E-Print Network [OSTI]

    Alexander I. Shlyakhter

    2003-08-06

    [NOTE: This 1983 preprint is being uploaded to arXiv.org after the death of its author, who supported online distribution of his work. Contact info of the submitter is at http://ilya.cc .] The positions of neutron resonances have been shown to be highly sensitive to the variation of fundamental nuclear constants. The analysis of the measured isotopic shifts in the natural fossil reactor at Oklo gives the following restrictions on the possible rates of the interaction constants variation: strong ~2x10^-19 yr^-1, electromagnetic ~5x10^-18 yr^-1, weak ~10^-12 yr^-1. These limits permit to exclude all the versions of nuclear constants contemporary variation discussed in the literature. URL: http://alexonline.info >. For more recent analyses see hep-ph/9606486, hep-ph/0205206 and astro-ph/0204069 .

  18. Long term out-of-pile thermocouple tests in conditions representative for nuclear gas-cooled high temperature reactors

    SciTech Connect (OSTI)

    Laurie, M.; Fourrez, S.; Fuetterer, M. A.; Lapetite, J. M.

    2011-07-01

    During irradiation tests at high temperature, failure of commercial Inconel 600 sheathed thermocouples is commonly encountered. To understand and remedy this problem, out-of-pile tests were performed with thermocouples in carburizing atmospheres which can be assumed to be at least locally representative for High Temperature Reactors. The objective was to screen those thermocouples which would consecutively be used under irradiation. Two such screening tests have been performed with a set of thermocouples embedded in graphite (mainly conventional Type N thermocouples and thermocouples with innovative sheaths) in a dedicated furnace with helium flushing. Performance indicators such as thermal drift, insulation and loop resistance were monitored and compared to those from conventional Type N thermocouples. Several parameters were investigated: niobium sleeves, bending, thickness, sheath composition, temperature as well as the chemical environment. After the tests, Scanning Electron Microscopy (SEM) examinations were performed to analyze possible local damage in wires and in the sheath. The present paper describes the two experiments, summarizes results and outlines further work, in particular to further analyze the findings and to select suitable thermocouples for qualification under irradiation. (authors)

  19. Facility Closure Report for T-Tunnel (U12t), Area 12, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2008-08-01

    This Facility Closure Report (FCR) has been prepared to document the actions taken to permanently close the remaining accessible areas of U12t-Tunnel (T-Tunnel) in Area 12 of the Nevada Test Site (NTS). The closure of T-Tunnel was a prerequisite to transfer facility ownership from the Defense Threat Reduction Agency (DTRA) to the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO). Closure of the facility was accomplished with the cooperation and concurrence of both NNSA/NSO and the Nevada Division of Environmental Protection (NDEP). The purpose of this FCR is to document that the closure of T-Tunnel complied with the closure requirements specified in the Facility Closure Plan for N- and T-Tunnels Area 12, Nevada Test Site (Appendix D) and that the facility is ready for transfer to NNSA/NSO. The Facility Closure Plan (FCP) is provided in Appendix D. T-Tunnel is located approximately 42 miles north of Mercury in Area 12 of the NTS (Figure 1). Between 1970 and 1987, T-Tunnel was used for six Nuclear Weapons Effects Tests (NWETs). The tunnel was excavated horizontally into the volcanic tuffs of Rainier Mesa. The T-Tunnel complex consists of a main access drift with two NWET containment structures, a Gas Seal Plug (GSP), and a Gas Seal Door (GSD) (Figure 2). The T-Tunnel complex was mothballed in 1993 to preserve the tunnel for resumption of testing, should it happen in the future, to stop the discharge of tunnel effluent, and to prevent unauthorized access. This was accomplished by sealing the main drift GSD.

  20. Advanced Test Reactor Complex Facilities Radioactive Waste Management Basis and DOE Manual 435.1-1 Compliance Tables

    SciTech Connect (OSTI)

    Lisa Harvego; Brion Bennett

    2011-11-01

    U.S. Department of Energy Order 435.1, 'Radioactive Waste Management,' along with its associated manual and guidance, requires development and maintenance of a radioactive waste management basis for each radioactive waste management facility, operation, and activity. This document presents a radioactive waste management basis for Idaho National Laboratory's Advanced Test Reactor Complex facilities that manage radioactive waste. The radioactive waste management basis for a facility comprises existing laboratory-wide and facility-specific documents. U.S. Department of Energy Manual 435.1-1, 'Radioactive Waste Management Manual,' facility compliance tables also are presented for the facilities. The tables serve as a tool to develop the radioactive waste management basis.

  1. Modeling of divertor geometry effects in China fusion engineering testing reactor by SOLPS/B2-Eirene

    SciTech Connect (OSTI)

    Zhao, M. L., E-mail: zml812@mail.ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Chen, Y. P.; Li, G. Q.; Luo, Z. P. [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China)] [Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Guo, H. Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China) [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); General Atomics, P.O. Box 85608, San Diego, California 92186 (United States); Ye, M. Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China) [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027 (China); Institute of Plasma Physics, Chinese Academy of Science, Hefei 230031 (China); Tendler, M. [Alfven Laboratory, Royal Institute of Technology, Stockholm (Sweden)] [Alfven Laboratory, Royal Institute of Technology, Stockholm (Sweden)

    2014-05-15

    The China Fusion Engineering Testing Reactor (CFETR) is currently under design. The SOLPS/B2-Eirene code package is utilized for the design and optimization of the divertor geometry for CFETR. Detailed modeling is carried out for an ITER-like divertor configuration and one with relatively open inner divertor structure, to assess, in particular, peak power loading on the divertor target, which is a key issue for the operation of a next-step fusion machine, such as ITER and CFETR. As expected, the divertor peak heat flux greatly exceeds the maximum steady-state heat load of 10?MW/m{sup 2}, which is a limit dictated by engineering, for both divertor configurations with a wide range of edge plasma conditions. Ar puffing is effective at reducing divertor peak heat fluxes below 10?MW/m{sup 2} even at relatively low densities for both cases, favoring the divertor configuration with more open inner divertor structure.

  2. Integral and Separate Effects Tests for Thermal Hydraulics Code Validation for Liquid-Salt Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Peterson, Per

    2012-10-30

    The objective of the 3-year project was to collect integral effects test (IET) data to validate the RELAP5-3D code and other thermal hydraulics codes for use in predicting the transient thermal hydraulics response of liquid salt cooled reactor systems, including integral transient response for forced and natural circulation operation. The reference system for the project is a modular, 900-MWth Pebble Bed Advanced High Temperature Reactor (PB-AHTR), a specific type of Fluoride salt-cooled High temperature Reactor (FHR). Two experimental facilities were developed for thermal-hydraulic integral effects tests (IETs) and separate effects tests (SETs). The facilities use simulant fluids for the liquid fluoride salts, with very little distortion to the heat transfer and fluid dynamics behavior. The CIET Test Bay facility was designed, built, and operated. IET data for steady state and transient natural circulation was collected. SET data for convective heat transfer in pebble beds and straight channel geometries was collected. The facility continues to be operational and will be used for future experiments, and for component development. The CIET 2 facility is larger in scope, and its construction and operation has a longer timeline than the duration of this grant. The design for the CIET 2 facility has drawn heavily on the experience and data collected on the CIET Test Bay, and it was completed in parallel with operation of the CIET Test Bay. CIET 2 will demonstrate start-up and shut-down transients and control logic, in addition to LOFC and LOHS transients, and buoyant shut down rod operation during transients. Design of the CIET 2 Facility is complete, and engineering drawings have been submitted to an external vendor for outsourced quality controlled construction. CIET 2 construction and operation continue under another NEUP grant. IET data from both CIET facilities is to be used for validation of system codes used for FHR modeling, such as RELAP5-3D. A set of numerical models were developed in parallel to the experimental work. RELAP5-3D models were developed for the salt-cooled PB-AHTR, and for the simulat fluid CIET natural circulation experimental loop. These models are to be validated by the data collected from CIET. COMSOL finite element models were used to predict the temperature and fluid flow distribution in the annular pebble bed core; they were instrumental for design of SETs, and they can be used for code-to-code comparisons with RELAP5-3D. A number of other small SETs, and numerical models were constructed, as needed, in support of this work. The experiments were designed, constructed and performed to meet CAES quality assurance requirements for test planning, implementation, and documentation; equipment calibration and documentation, procurement document control; training and personnel qualification; analysis/modeling software verification and validation; data acquisition/collection and analysis; and peer review.

  3. SUMMARY OF ‘AFIP’ FULL SIZED PLATE IRRADIATIONS IN THE ADVANCED TEST REACTOR

    SciTech Connect (OSTI)

    Robinson, Adam B; Wachs, Daniel M

    2010-03-01

    Recent testing at the Idaho National Laboratory has included four AFIP (ATR Full Size plate In center flux trap Position) experiments. These experiments included both dispersion plates and monolithic plates fabricated by both hot isostatic pressing and friction bonding utilizing both thermally sprayed inter-layers and zirconium barriers. These plates were tested between 100 and 350 w/cm2 at low temperatures and high burn-ups. The post irradiation exams performed have indicated good performance under the conditions tested and a summary of the findings and irradiation history are included herein.

  4. Testing and evaluation of large-area heliostats for solar thermal applications

    SciTech Connect (OSTI)

    Strachan, J.W.; Houser, R.M.

    1993-02-01

    Two heliostats representing the state-of-the-art in glass-metal designs for central receiver (and photovoltaic tracking) applications were tested and evaluated at the National Solar Thermal Test Facility in Albuquerque, New Mexico from 1986 to 1992. These heliostats have collection areas of 148 and 200 m{sup 2} and represent low-cost designs for heliostats that employ glass-metal mirrors. The evaluation encompassed the performance and operational characteristics of the heliostats, and examined heliostat beam quality, the effect of elevated winds on beam quality, heliostat drives and controls, mirror module reflectance and durability, and the overall operational and maintenance characteristics of the two heliostats. A comprehensive presentation of the results of these and other tests is presented. The results are prefaced by a review of the development (in the United States) of heliostat technology.

  5. NEUTRON ACTIVATION COOL-DOWN OF THE TOKAMAK FUSION TEST REACTOR

    E-Print Network [OSTI]

    maintenance, reliability, and performance requirements. Fig. 1 shows a partial schematic plan view of the TFTR system pump ducts, and the activation measurements. The characteristics of the Test Cell shielding

  6. Offsite environmental monitoring report; radiation monitoring around United States nuclear test areas, Calendar Year 1996

    SciTech Connect (OSTI)

    Davis, M.G.; Flotard, R.D.; Fontana, C.A.; Huff, P.A.; Maunu, H.K.; Mouck, T.L.; Mullen, A.A.; Sells, M.D.

    1997-08-01

    This report describes the Offsite Radiation Safety Program. This laboratory operated an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling milk, water, and air; by deploying thermoluminescent dosimeters (TLDs); and using pressurized ionization chambers (PICs). No nuclear weapons testing was conducted in 1996 due to the continuing nuclear test moratorium. During this period, R and IE personnel maintained readiness capability to provide direct monitoring support if testing were to be resumed and ascertained compliance with applicable EPA, DOE, state, and federal regulations and guidelines. Comparison of the measurements and sample analysis results with background levels and with appropriate standards and regulations indicated that there was no airborne radioactivity from diffusion or resuspension detected by the various EPA monitoring networks surrounding the NTS. There was no indication of potential migration of radioactivity to the offsite area through groundwater and no radiation exposure above natural background was received by the offsite population. All evaluated data were consistent with previous data history.

  7. A FEASIBILITY AND OPTIMIZATION STUDY TO DETERMINE COOLING TIME AND BURNUP OF ADVANCED TEST REACTOR FUELS USING A NONDESTRUCTIVE TECHNIQUE

    SciTech Connect (OSTI)

    Jorge Navarro

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method for ATR applications the technique was tested using one-isotope, multi-isotope and fuel simulated sources. Burnup calibrations were perfomed using convoluted and deconvoluted data. The calibrations results showed burnup prediction by this method improves using deconvolution. The final stage of the deconvolution method development was to perform an irradiation experiment in order to create a surrogate fuel source to test the deconvolution method using experimental data. A conceptual design of the fuel scan system is path forward using the rugged LaBr3 detector in an above the water configuration and deconvolution algorithms.

  8. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  9. Closure Report for Corrective Action Unit 398: Area 25 Spill Sites, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    K. B. Campbell

    2003-04-01

    This Closure Report (CR) documents the activities performed to close Corrective Action Unit (CAU) 398: Area 25 Spill Sites, in accordance with the Federal Facility Agreement and Consent Order (FFACO) of 1996, and the Nevada Division of Environmental Protection (NDEP)-approved Streamlined Approach for Environmental Restoration (SA4FER) Plan for CAU 398: Area 25 Spill Sites, Nevada Test Site, Nevada (U.S. Department of Energy, Nevada Operations Office [DOEN], 2001). CAU 398 consists of the following thirteen Corrective Action Sites (CASs) all located in Area 25 of the Nevada Test Site (NTS) (Figure 1): CAS 25-25-02, Oil Spills, CAS 25-25-03, Oil Spills, CAS 25-25-04, Oil Spills, CAS 25-25-05, Oil Spills, CAS 25-25-06, Oil Spills, CAS 25-25-07, Hydraulic Oil Spill(s), CAS 25-25-08, Hydraulic Oil Spill(s), CAS 25-25-16, Diesel Spill (from CAS 25-01-02), CAS 25-25-17, Subsurface Hydraulic Oil Spill, CAS 25-44-0 1, Fuel Spill, CAS 25-44-04, Acid Spill (from CAS 25-01-01), CAS 25-44-02, Spill, and CAS 25-44-03, Spill. Copies of the analytical results for the site verification samples are included in Appendix B. Copies of the CAU Use Restriction Information forms are included in Appendix C.

  10. Transferability of Data Related to the Underground Test Area Project, Nevada Test Site, Nye County, Nevada: Revision 0

    SciTech Connect (OSTI)

    Stoller-Navarro Joint Venture

    2004-06-24

    This document is the collaborative effort of the members of an ad hoc subcommittee of the Underground Test Area (UGTA) Technical Working Group (TWG). The UGTA Project relies on data from a variety of sources; therefore, a process is needed to identify relevant factors for determining whether material-property data collected from other areas can be used to support groundwater flow, radionuclide transport, and other models within a Corrective Action Unit (CAU), and for documenting the data transfer decision and process. This document describes the overall data transfer process. Separate Parameter Descriptions will be prepared that provide information for selected specific parameters as determined by the U.S. Department of Energy (DOE) UGTA Project Manager. This document and its accompanying appendices do not provide the specific criteria to be used for transfer of data for specific uses. Rather, the criteria will be established by separate parameter-specific and model-specific Data Transfer Protocols. The CAU Data Documentation Packages and data analysis reports will apply the protocols and provide or reference a document with the data transfer evaluations and decisions.

  11. Offsite environmental monitoring report: Radiation monitoring around United States nuclear test areas, calendar year 1991

    SciTech Connect (OSTI)

    Chaloud, D.J.; Dicey, B.B.; Mullen, A.A.; Neale, A.C.; Sparks, A.R.; Fontana, C.A.; Carroll, L.D.; Phillips, W.G.; Smith, D.D.; Thome, D.J.

    1992-01-01

    This report describes the Offsite Radiation Safety Program conducted during 1991 by the Environmental Protection Agency`s (EPA`s) Environmental Monitoring Systems Laboratory-Las Vegas. This laboratory operates an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling milk, water, and air; by deploying thermoluminescent dosimeters (TLDs) and using pressurized ion chambers (PICs); and by biological monitoring of animals, food crops, and humans. Personnel with mobile monitoring equipment are placed in areas downwind from the test site prior to each nuclear weapons test to implement protective actions, provide immediate radiation monitoring, and obtain environmental samples rapidly after any occurrence of radioactivity release. Comparison of the measurements and sample analysis results with background levels and with appropriate standards and regulations indicated that there was no radioactivity detected offsite by the various EPA monitoring networks and no exposure above natural background to the population living in the vicinity of the NTS that could be attributed to current NTS activities. Annual and long-term trends were evaluated in the Noble Gas, Tritium, Milk Surveillance, Biomonitoring, TLD, PIC networks, and the Long-Term Hydrological Monitoring Program.

  12. Report on expedited site characterization of the Central Nevada Test Area, Nye County, Nevada

    SciTech Connect (OSTI)

    Yuhr, L. [Technos Inc., Miami, FL (United States)] [Technos Inc., Miami, FL (United States); Wonder, J.D.; Bevolo, A.J. [Ames Lab., IA (United States)] [Ames Lab., IA (United States)

    1997-09-01

    This report documents data collection, results, and interpretation of the expedited site characterization (ESC) pilot project conducted from September 1996 to June 1997 at the Central Nevada Test Area (CNTA), Nye County, Nevada. Characterization activities were limited to surface sites associated with deep well drilling and ancillary operations at or near three emplacement well areas. Environmental issues related to the underground nuclear detonation (Project Faultless) and hydrologic monitoring wells were not addressed as a part of this project. The CNTA was divided into four functional areas for the purpose of this investigation and report. These areas include the vicinity of three emplacement wells (UC-1, UC-3, and UC-4) and one mud waste drilling mud collection location (Central Mud Pit; CMP). Each of these areas contain multiple, potentially contaminated features, identified either from historic information, on-site inspections, or existing data. These individual features are referred to hereafter as ``sites.`` The project scope of work involved site reconnaissance, establishment of local grid systems, site mapping and surveying, geophysical measurements, and collection and chemical analysis of soil and drilling mud samples. Section 2.0 through Section 4.0 of this report provide essential background information about the site, project, and details of how the ESC method was applied at CNTA. Detailed discussion of the scope of work is provided in Section 5.0, including procedures used and locations and quantities of measurements obtained. Results and interpretations for each of the four functional areas are discussed separately in Sections 6.0, 7.0, 8.0, and 9.0. These sections provide a chronological presentation of data collected and results obtained, followed by interpretation on a site-by-site basis. Key data is presented in the individual sections. The comprehensive set of data is contained in appendices.

  13. Closure Plan for the Area 5 Radioactive Waste Management Site at the Nevada Test Site

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2008-09-01

    The Area 5 Radioactive Waste Management Site (RMWS) at the Nevada Test Site (NTS) is managed and operated by National Security Technologies, LLC (NSTec), for the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO). This document is the first update of the preliminary closure plan for the Area 5 RWMS at the NTS that was presented in the Integrated Closure and Monitoring Plan (DOE, 2005a). The major updates to the plan include a new closure schedule, updated closure inventory, updated site and facility characterization data, the Title II engineering cover design, and the closure process for the 92-Acre Area of the RWMS. The format and content of this site-specific plan follows the Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans (DOE, 1999a). This interim closure plan meets closure and post-closure monitoring requirements of the order DOE O 435.1, manual DOE M 435.1-1, Title 40 Code of Federal Regulations (CFR) Part 191, 40 CFR 265, Nevada Administrative Code (NAC) 444.743, and Resource Conservation and Recovery Act (RCRA) requirements as incorporated into NAC 444.8632. The Area 5 RWMS accepts primarily packaged low-level waste (LLW), low-level mixed waste (LLMW), and asbestiform low-level waste (ALLW) for disposal in excavated disposal cells.

  14. Performance of AGR-1 High-Temperature Reactor Fuel During Post-Irradiation Heating Tests

    SciTech Connect (OSTI)

    Morris, Robert Noel [ORNL; Baldwin, Charles A [ORNL; Hunn, John D [ORNL; Demkowicz, Paul [Idaho National Laboratory (INL); Reber, Edward [Idaho National Laboratory (INL)

    2014-01-01

    The fission product retention of irradiated low-enriched uranium oxide/uranium carbide TRISO fuel compacts from the AGR-1 experiment has been evaluated at temperatures of 1600 1800 C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4 to 19.1% FIMA have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium, and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 10-6 after 300 h at 1600 C or 100 h at 1800 C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 C, and 85Kr release was very low during the tests (particles with breached SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 C in one compact. Post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.

  15. Measurement of Nuclear Interaction Rates in Crystal Using the CERN-SPS North Area Test Beams

    E-Print Network [OSTI]

    Losito, R; Taratin, A

    2010-01-01

    A number of tests were performed in the North area of the SPS in view of investigating crystal-particles interactions for future application in hadron colliders. The rate of nuclear interactions was measured with 400 GeV proton beams directed into a silicon bent crystal. In this way the background induced by the crystal either in amorphous or in channeling orientation was revealed. The results provide fundamental information to put in perspective the use of silicon crystals to assist halo collimation in hadron colliders, whilst minimizing the induced loss.

  16. Nevada Test Site 2007 Data Report: Groundwater Monitoring Program Area 5 Radioactive Waste Management Site

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2008-01-01

    This report is a compilation of the groundwater sampling results from three monitoring wells located near the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site (NTS), Nye County, Nevada, for calendar year 2007. The NTS is an approximately 3,561 square kilometer (1,375 square mile) restricted-access federal installation located approximately 105 kilometers (65 miles) northwest of Las Vegas, Nevada (Figure 1). Pilot wells UE5PW-1, UE5PW-2, and UE5PW-3 are used to monitor the groundwater at the Area 5 RWMS (Figure 2). In addition to groundwater monitoring results, this report includes information regarding site hydrogeology, well construction, sample collection, and meteorological data measured at the Area 5 RWMS. The disposal of low-level radioactive waste and mixed low-level radioactive waste at the Area 5 RWMS is regulated by U.S. Department of Energy (DOE) Order 435.1, 'Radioactive Waste Management'. The disposal of mixed low-level radioactive waste is also regulated by the state of Nevada under the Resource Conservation and Recovery Act (RCRA) regulation Title 40 Code of Federal Regulations (CFR) Part 265, 'Interim Status Standards for Owners and Operators of Hazardous Waste Treatment, Storage, and Disposal Facilities' (CFR, 1999). The format of this report was requested by the Nevada Division of Environmental Protection (NDEP) in a letter dated August 12, 1997. The appearance and arrangement of this document have been modified slightly since that date to provide additional information and to facilitate the readability of the document. The objective of this report is to satisfy any Area 5 RWMS reporting agreements between DOE and NDEP.

  17. TREATABILITY TEST REPORT FOR THE REMOVAL OF CHROMIUM FROM GROUNDWATER AT 100-D AREA USING ELECTROCOAGULATION

    SciTech Connect (OSTI)

    PETERSEN SW

    2009-09-24

    The U.S. Department of Energy (DOE) has committed to accelerate cleanup of contaminated groundwater along the Columbia River. The current treatment approach was driven by a series of Interim Action Records of Decision (IAROD) issued in the mid-1990s. Part of the approach for acceleration involves increasing the rate of groundwater extraction for the chromium plume north of the 100-D Reactor and injecting the treated water in strategic locations to hydraulically direct contaminated groundwater toward the extraction wells. The current treatment system uses ion exchange for Cr(VI) removal, with off-site regeneration of the ion exchange resins. Higher flow rates will increase the cost and frequency of ion exchange resin regeneration; therefore, alternative technologies are being considered for treatment at high flow rates. One of these technologies, electrocoagulation (EC), was evaluated through a pilot-scale treatability test. The primary purpose of the treatability study was to determine the effectiveness of Cr(VI) removal and the robustness/implementability of an EC system. Secondary purposes of the study were to gather information about derivative wastes and to obtain data applicable to scaling the process from the treatability scale to full-scale. The treatability study work plan identified a performance objective and four operational objectives. The performance objective for the treatability study was to determine the efficiency (effectiveness) of hexavalent chromium removal from the groundwater, with a desired concentration of {le} 20 micrograms per liter ({micro}g/L) Cr(VI) in the effluent prior to re-injection. Influent and effluent total chromium and hexavalent chromium data were collected using a field test kit for multiple samples per week, and from off-site laboratory analysis of samples collected approximately monthly. These data met all data quality requirements. Two of three effluent chromium samples analyzed in the off-site (that is, fixed) laboratory met the performance objective during the continuous operational testing. Effluent hexavalent chromium analyzed by the field laboratory met the performance goal in over 90 percent of the samples. All effluent hexavalent chromium samples during the batch testing with high influent hexavalent chromium concentrations ({approx}2000 {micro}g/L) met the performance objective. Although the EC system was able to meet the performance goal, it must be noted that it was not uncommon for the system to be operated in recycle mode to achieve the performance goal. The EC unit was sometimes, but not always, capable of a single pass treatment efficiency high enough to meet the performance goal, and recycling water for multiple treatment passes was effective. An operational objective was to determine the volume and composition of the waste streams to enable proper waste designation. The toxicity characteristic leaching procedure (TCLP) concentrations, pH, and free liquids were determined for solid material from the EC electrodes (mechanically removed scale), the filter press, and the tank bottoms for the effluent and waste collector tanks. These data met all data quality requirements. All solid-phase secondary waste streams were found to be below the TCLP limits for the toxicity characteristic, and a pH value within the limits for the corrosivity characteristic. Out of three samples, two (one of scale from the EC unit and one from filter press solids) failed the free liquid (paint filter) test, which is one of the acceptability criteria for Hanford's Environmental Restoration Disposal Facility (ERDF). The solid-phase waste generation rate was about 0.65-gallon of solid waste per 100 gallons of water treated. It is concluded that the solid-phase secondary waste generated from this technology under the conditions at the test site will meet the toxicity and corrosivity criteria for disposal. It is also concluded that with engineering and/or operational improvements, a solid-phase secondary waste could be produced that would meet the free liquid disposal requirements. The second oper

  18. Closure Report for Corrective Action Unit 135: Areas 25 Underground Storage Tanks, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    D. H. Cox

    2001-06-01

    Corrective Action Unit (CAU) 135, Area 25 Underground Storage Tanks, was closed in accordance with the approved Corrective Action Plan (DOE/NV, 2000). CAU 135 consists of three Corrective Action Sites (CAS). Two of these CAS's were identified in the Corrective Action Investigation Data Quality Objective meeting as being improperly identified as underground storage tanks. CAS 25-02-03 identified as the Deluge Valve Pit was actually an underground electrical vault and CAS 25-02-10 identified as an Underground Storage Tank was actually a former above ground storage tank filled with demineralized water. Both of these CAS's are recommended for a no further action closure. CAS 25-02-01 the Underground Storage Tanks commonly referred to as the Engine Maintenance Assembly and Disassembly Waste Holdup Tanks and Vault was closed by decontaminating the vault structure and conducting a radiological verification survey to document compliance with the Nevada Test Site unrestricted use release criteria. The Area 25 Underground Storage Tanks, (CAS 25-02-01), referred to as the Engine Maintenance, Assembly, and Disassembly (E-MAD) Waste Holdup Tanks and Vault, were used to receive liquid waste from all of the radioactive and cell service area drains at the E-MAD Facility. Based on the results of the Corrective Action Investigation conducted in June 1999, discussed in ''The Corrective Action Investigation Plan for Corrective Action Unit 135: Area 25 Underground Storage Tanks, Nevada Test Site, Nevada'' (DOE/NV, 199a), one sample from the radiological survey of the concrete vault interior exceeded radionuclide preliminary action levels. The analytes from the sediment samples exceeded the preliminary action levels for polychlorinated biphenyls, Resource Conservation and Recovery Act metals, total petroleum hydrocarbons as diesel-range organics, and radionuclides. The CAU 135 closure activities consisted of scabbling radiological ''hot spots'' from the concrete vault, and the drilling removal of the cement-lined vault sump. Field activities began on November 28, 2000, and ended on December 4, 2000. After verification samples were collected, the vault was repaired with cement. The concrete vault sump, soil excavated beneath the sump, and compactable hot line trash were disposed at the Area 23 Sanitary Landfill. The vault interior was field surveyed following the removal of waste to verify that unrestricted release criteria had been achieved. Since the site is closed by unrestricted release decontamination and verification, post-closure care is not required.

  19. Offsite environmental monitoring report: Radiation monitoring around United States nuclear test areas, calendar year 1997

    SciTech Connect (OSTI)

    Davis, M.G.; Flotard, R.D.; Fontana, C.A.; Hennessey, P.A.; Maunu, H.K.; Mouck, T.L.; Mullen, A.A.; Sells, M.D.

    1999-01-01

    This report describes the Offsite Radiological Environmental Monitoring Program (OREMP) conducted during 1997 by the US Environmental Protection Agency`s (EPAs), Radiation and Indoor Environments National Laboratory, Las Vegas, Nevada. This laboratory operated an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling and analyzing milk, water, and air; by deploying and reading thermoluminescent dosimeters (TLDs); and using pressurized ionization chambers (PICs) to measure ambient gamma exposure rates with a sensitivity capable of detecting low level exposures not detected by other monitoring methods.

  20. Offsite environmental monitoring report: Radiation monitoring around United States nuclear test areas, calendar year 1993

    SciTech Connect (OSTI)

    Chaloud, D.J; Daigler, D.M.; Davis, M.G.

    1996-06-01

    This report describes the Offsite Radiation Safety Program conducted during 1993 by the Environmental Protection Agency`s (EPA`s) Environmental Monitoring Systems Laboratory - Las Vegas (EMSL-LV). This laboratory operates an environmental radiation monitoring program in the region surrounding the Nevada Test Site (NTS) and at former test sites in Alaska, Colorado, Mississippi, Nevada, and New Mexico. The surveillance program is designed to measure levels and trends of radioactivity, if present, in the environment surrounding testing areas to ascertain whether current radiation levels and associated doses to the general public are in compliance with existing radiation protection standards. The surveillance program additionally has the responsibility to take action to protect the health and well being of the public in the event of any accidental release of radioactive contaminants. Offsite levels of radiation and radioactivity are assessed by sampling milk, water, and air; by deploying thermoluminescent dosimeters (TLDs) and using pressurized ionization chambers (PICs); by biological monitoring of foodstuffs including animal tissues and food crops; and by measurement of radioactive material deposited in humans.

  1. 100 Area soil washing: Bench scale tests on 116-F-4 pluto crib soil

    SciTech Connect (OSTI)

    Field, J.G.

    1994-06-10

    The Pacific Northwest Laboratory conducted a bench-scale treatability study on a pluto crib soil sample from 100 Area of the Hanford Site. The objective of this study was to evaluate the use of physical separation (wet sieving), treatment processes (attrition scrubbing, and autogenous surface grinding), and chemical extraction methods as a means of separating radioactively-contaminated soil fractions from uncontaminated soil fractions. The soil washing treatability study was conducted on a soil sample from the 116-F-4 Pluto Crib that had been dug up as part of an excavation treatability study. Trace element analyses of this soil showed no elevated concentrations above typically uncontaminated soil background levels. Data on the distribution of radionuclide in various size fractions indicated that the soil-washing tests should be focused on the gravel and sand fractions of the 116-F-4 soil. The radionuclide data also showed that {sup 137}Cs was the only contaminant in this soil that exceeded the test performance goal (TPG). Therefore, the effectiveness of subsequent soil-washing tests for 116-F-4 soil was evaluated on the basis of activity attenuation of {sup 137}Cs in the gravel- and sand-size fractions.

  2. Hydrologic Resources Management Program and Underground Test Area Project FY 2006 Progress Report

    SciTech Connect (OSTI)

    Culham, H W; Eaton, G F; Genetti, V; Hu, Q; Kersting, A B; Lindvall, R E; Moran, J E; Blasiyh Nuno, G A; Powell, B A; Rose, T P; Singleton, M J; Williams, R W; Zavarin, M; Zhao, P

    2008-04-08

    This report describes FY 2006 technical studies conducted by the Chemical Biology and Nuclear Science Division (CBND) at Lawrence Livermore National Laboratory (LLNL) in support of the Hydrologic Resources Management Program (HRMP) and the Underground Test Area Project (UGTA). These programs are administered by the U.S. Department of Energy, National Nuclear Security Administration, Nevada Site Office (NNSA/NSO) through the Defense Programs and Environmental Restoration Divisions, respectively. HRMP-sponsored work is directed toward the responsible management of the natural resources at the Nevada Test Site (NTS), enabling its continued use as a staging area for strategic operations in support of national security. UGTA-funded work emphasizes the development of an integrated set of groundwater flow and contaminant transport models to predict the extent of radionuclide migration from underground nuclear testing areas at the NTS. The report is organized on a topical basis and contains four chapters that highlight technical work products produced by CBND. However, it is important to recognize that most of this work involves collaborative partnerships with the other HRMP and UGTA contract organizations. These groups include the Energy and Environment Directorate at LLNL (LLNL-E&E), Los Alamos National Laboratory (LANL), the Desert Research Institute (DRI), the U.S. Geological Survey (USGS), Stoller-Navarro Joint Venture (SNJV), and National Security Technologies (NSTec). Chapter 1 is a summary of FY 2006 sampling efforts at near-field 'hot' wells at the NTS, and presents new chemical and isotopic data for groundwater samples from four near-field wells. These include PM-2 and U-20n PS 1DDh (CHESHIRE), UE-7ns (BOURBON), and U-19v PS No.1ds (ALMENDRO). Chapter 2 is a summary of the results of chemical and isotopic measurements of groundwater samples from three UGTA environmental monitoring wells. These wells are: ER-12-4 and U12S located in Area 12 on Rainier Mesa and USGS HGH No.2 WW2 located in Yucca Flat. In addition, three springs were sampled White Rock Spring and Captain Jack Spring in Area 12 on Rainier Mesa and Topopah Spring in Area 29. Chapter 3 is a compilation of existing noble gas data that has been reviewed and edited to remove inconsistencies in presentation of total vs. single isotope noble gas values reported in the previous HRMP and UGTA progress reports. Chapter 4 is a summary of the results of batch sorption and desorption experiments performed to determine the distribution coefficients (Kd) of Pu(IV), Np(V), U(VI), Cs and Sr to zeolitized tuff (tuff confining unit, TCU) and carbonate (lower carbonate aquifer, LCA) rocks in synthetic NTS groundwater Chapter 5 is a summary of the results of a series of flow-cell experiments performed to examine Np(V) and Pu(V) sorption to and desorption from goethite. Np and Pu desorption occur at a faster rate and to a greater extent than previously reported. In addition, oxidation changes occurred with the Pu whereby the surface-sorbed Pu(IV) was reoxidized to aqueous Pu(V) during desorption.

  3. Status of the Norwegian thorium light water reactor (LWR) fuel development and irradiation test program

    SciTech Connect (OSTI)

    Drera, S.S.; Bjork, K.I.; Kelly, J.F.; Asphjell, O. [Thor Energy AS: Sommerrogaten 13-15, Oslo, NO255 (Norway)

    2013-07-01

    Thorium based fuels offer several benefits compared to uranium based fuels and should thus be an attractive alternative to conventional fuel types. In order for thorium based fuel to be licensed for use in current LWRs, material properties must be well known for fresh as well as irradiated fuel, and accurate prediction of fuel behavior must be possible to make for both normal operation and transient scenarios. Important parameters are known for fresh material but the behaviour of the fuel under irradiation is unknown particularly for low Th content. The irradiation campaign aims to widen the experience base to irradiated (Th,Pu)O{sub 2} fuel and (Th,U)O{sub 2} with low Th content and to confirm existing data for fresh fuel. The assumptions with respect to improved in-core fuel performance are confirmed by our preliminary irradiation test results, and our fuel manufacture trials so far indicate that both (Th,U)O{sub 2} and (Th,Pu)O{sub 2} fuels can be fabricated with existing technologies, which are possible to upscale to commercial volumes.

  4. Process Knowledge Summary Report for Advanced Test Reactor Complex Contact-Handled Transuranic Waste Drum TRA010029

    SciTech Connect (OSTI)

    B. R. Adams; R. P. Grant; P. R. Smith; J. L. Weisgerber

    2013-09-01

    This Process Knowledge Summary Report summarizes information collected to satisfy the transportation and waste acceptance requirements for the transfer of one drum containing contact-handled transuranic (TRU) actinide standards generated by the Idaho National Laboratory at the Advanced Test Reactor (ATR) Complex to the Advanced Mixed Waste Treatment Project (AMWTP) for storage and subsequent shipment to the Waste Isolation Pilot Plant for final disposal. The drum (i.e., Integrated Waste Tracking System Bar Code Number TRA010029) is currently stored at the Materials and Fuels Complex. The information collected includes documentation that addresses the requirements for AMWTP and applicable sections of their Resource Conservation and Recovery Act permits for receipt and disposal of this TRU waste generated from ATR. This Process Knowledge Summary Report includes information regarding, but not limited to, the generation process, the physical form, radiological characteristics, and chemical contaminants of the TRU waste, prohibited items, and packaging configuration. This report, along with the referenced supporting documents, will create a defensible and auditable record for this TRU waste originating from ATR.

  5. 2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2012-02-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  6. Nevada Test Site Radionuclide Inventory and Distribution Program: Report No. 2. Areas 2 and 4

    SciTech Connect (OSTI)

    McArthur, R.D.; Kordas, J.F.

    1985-09-01

    Radionuclide activity was measured by in situ spectrometry at 349 locations in Areas 2 and 4 of the Nevada Test Site. The data were analyzed by kriging and other methods to estimate the total inventory and distribution of six man-made radionuclides that were present in measurable amounts. Isotope ratios in soil samples were then used to infer the inventories of three other radionuclides. The estimated inventories were: /sup 241/Am, 8 curies; /sup 238/Pu, 18 curies; /sup 239,240/Pu, 51 curies; /sup 60/Co, 7 curies; /sup 137/Cs, 34 curies; /sup 90/Sr, 71 curies; /sup 152/Eu, 35 curies; /sup 154/Eu, 6 curies; and /sup 155/Eu, 3 curies.

  7. Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 118: Area 27 Super Kukla Facility, Nevada Test Site, Nevada, Rev. No.: 1

    SciTech Connect (OSTI)

    David Strand

    2006-09-01

    This Streamlined Approach for Environmental Restoration (SAFER) plan addresses closure for Corrective Action Unit (CAU) 118, Area 27 Super Kukla Facility, identified in the ''Federal Facility Agreement and Consent Order''. Corrective Action Unit 118 consists of one Corrective Action Site (CAS), 27-41-01, located in Area 27 of the Nevada Test Site. Corrective Action Site 27-41-01 consists of the following four structures: (1) Building 5400A, Reactor High Bay; (2) Building 5400, Reactor Building and access tunnel; (3) Building 5410, Mechanical Building; and (4) Wooden Shed, a.k.a. ''Brock House''. This plan provides the methodology for field activities needed to gather the necessary information for closing the CAS. There is sufficient information and process knowledge from historical documentation and site confirmation data collected in 2005 and 2006 to recommend closure of CAU 118 using the SAFER process. The Data Quality Objective process developed for this CAU identified the following expected closure option: closure in place with use restrictions. This expected closure option was selected based on available information including contaminants of potential concern, future land use, and assumed risks. There are two decisions that need to be answered for closure. Decision I is to determine the nature of contaminants of concern in environmental media or potential source material that could impact human health or the environment. Decision II is to determine whether or not sufficient information has been obtained to confirm that closure objectives were met. This decision includes determining whether the extent of any contamination remaining on site has been defined, and whether actions have been taken to eliminate exposure pathways.

  8. THE WIDE-AREA ENERGY STORAGE AND MANAGEMENT SYSTEM PHASE II Final Report - Flywheel Field Tests

    SciTech Connect (OSTI)

    Lu, Ning; Makarov, Yuri V.; Weimar, Mark R.; Rudolph, Frank; Murthy, Shashikala; Arseneaux, Jim; Loutan, Clyde; Chowdhury, S.

    2010-08-31

    This research was conducted by Pacific Northwest National Laboratory (PNNL) operated for the U.S. department of Energy (DOE) by Battelle Memorial Institute for Bonneville Power Administration (BPA), California Institute for Energy and Environment (CIEE) and California Energy Commission (CEC). A wide-area energy management system (WAEMS) is a centralized control system that operates energy storage devices (ESDs) located in different places to provide energy and ancillary services that can be shared among balancing authorities (BAs). The goal of this research is to conduct flywheel field tests, investigate the technical characteristics and economics of combined hydro-flywheel regulation services that can be shared between Bonneville Power Administration (BPA) and California Independent System Operator (CAISO) controlled areas. This report is the second interim technical report for Phase II of the WAEMS project. This report presents: 1) the methodology of sharing regulation service between balancing authorities, 2) the algorithm to allocate the regulation signal between the flywheel and hydro power plant to minimize the wear-and-tear of the hydro power plants, 3) field results of the hydro-flywheel regulation service (conducted by the Beacon Power), and 4) the performance metrics and economic analysis of the combined hydro-flywheel regulation service.

  9. Closure Plan for the Area 3 Radioactive Waste Management Site at the Nevada Test Site

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2007-09-01

    The Area 3 Radioactive Waste Management Site (RMWS) at the Nevada Test Site (NTS) is managed and operated by National Security Technologies, LLC (NSTec) for the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office (NNSA/NSO). This document is the first update of the interim closure plan for the Area 3 RWMS, which was presented in the Integrated Closure and Monitoring Plan (ICMP) (DOE, 2005). The format and content of this plan follows the Format and Content Guide for U.S. Department of Energy Low-Level Waste Disposal Facility Closure Plans (DOE, 1999a). The major updates to the plan include a new closure date, updated closure inventory, the new institutional control policy, and the Title II engineering cover design. The plan identifies the assumptions and regulatory requirements, describes the disposal sites and the physical environment in which they are located, presents the design of the closure cover, and defines the approach and schedule for both closing and monitoring the site. The Area 3 RWMS accepts low-level waste (LLW) from across the DOE Complex in compliance with the NTS Waste Acceptance Criteria (NNSA/NSO, 2006). The Area 3 RWMS accepts both packaged and unpackaged unclassified bulk LLW for disposal in subsidence craters that resulted from deep underground tests of nuclear devices in the early 1960s. The Area 3 RWMS covers 48 hectares (119 acres) and comprises seven subsidence craters--U-3ax, U-3bl, U-3ah, U-3at, U-3bh, U-3az, and U-3bg. The area between craters U-3ax and U-3bl was excavated to form one large disposal unit (U-3ax/bl); the area between craters U-3ah and U-3at was also excavated to form another large disposal unit (U-3ah/at). Waste unit U-3ax/bl is closed; waste units U-3ah/at and U-3bh are active; and the remaining craters, although currently undeveloped, are available for disposal of waste if required. This plan specifically addresses the closure of the U-3ah/at and the U-3bh LLW units. A final closure cover has been placed on unit U-3ax/bl (Corrective Action Unit 110) at the Area 3 RWMS. Monolayer-evapotranspirative closure cover designs for the U-3ah/at and U-3bh units are provided in this plan. The current-design closure cover thickness is 3 meters (10 feet). The final design cover will have an optimized cover thickness, which is expected to be less than 3 m (10 ft). Although waste operations at the Area 3 RWMS have ceased at the end of June 2006, disposal capacity is available for future disposals at the U-3ah/at and U-3bh units. The Area 3 RWMS is expected to start closure activities in fiscal year 2025, which include the development of final performance assessment and composite analysis documents, closure plan, closure cover design for construction, cover construction, and initiation of the post-closure care and monitoring activities. Current monitoring at the Area 3 RWMS includes monitoring the cover of the closed mixed waste unit U-3ax/bl as required by the Nevada Department of Environmental Protection, and others required under federal regulations and DOE orders. Monitoring data, collected via sensors and analysis of samples, are needed to evaluate radiation doses to the general public, for performance assessment maintenance, to demonstrate regulatory compliance, and to evaluate the actual performance of the RWMSs. Monitoring provides data to ensure the integrity and performance of waste disposal units. The monitoring program is designed to forewarn management and regulators of any failure and need for mitigating actions. The plan describes the program for monitoring direct radiation, air, vadose zone, biota, groundwater, meteorology, and subsidence. The requirements of post-closure cover maintenance and monitoring will be determined in the final closure plan.

  10. Corrective Action Investigation Plan for Corrective Action Unit 262: Area 25 Septic Systems and Underground Discharge Point, Nevada Test Site, Nevada, Revision No. 1 (9/2001)

    SciTech Connect (OSTI)

    NNSA /NV

    2000-07-20

    This corrective action investigation plan contains the U.S. Department of Energy, Nevada Operations Office's approach to collect data necessary to evaluate corrective action alternatives appropriate for the closure of Corrective Action Unit (CAU) 262 under the Federal Facility Agreement and Consent Order. Corrective Action Unit 262 consists of nine Corrective Action Sites (CASs): Underground Storage Tank (25-02-06), Septic Systems A and B (25-04-06), Septic System (25-04-07), Leachfield (25-05-03), Leachfield (25-05-05), Leachfield (25-05-06), Radioactive Leachfield (25-05-08), Leachfield (25-05-12), and Dry Well (25-51-01). Situated in Area 25 at the Nevada Test Site (NTS), sites addressed by CAU 262 are located at the Reactor-Maintenance, Assembly, and Disassembly (R-MAD); Test Cell C; and Engine-Maintenance, Assembly, and Disassembly (E-MAD) facilities. The R-MAD, Test Cell C, and E-MAD facilities supported nuclear rocket reactor and engine testing as part of the Nuclear Rocket Development Station. The activities associated with the testing program were conducted between 1958 and 1973. Based on site history collected to support the Data Quality Objectives process, contaminants of potential concern (COPCs) for the site include oil/diesel-range total petroleum hydrocarbons, volatile organic compounds, semivolatile organic compounds, polychlorinated biphenyls, Resource Conservation and Recovery Act metals, and gamma-emitting radionuclides, isotopic uranium, isotopic plutonium, strontium-90, and tritium. The scope of the corrective action field investigation at the CAU will include the inspection of portions of the collection systems, sampling the contents of collection system features in situ of leachfield logging materials, surface soil sampling, collection of samples of soil underlying the base of inlet and outfall ends of septic tanks and outfall ends of diversion structures and distribution boxes, collection of soil samples from biased or a combination of biased and random locations within the boundaries of the leachfields, collection of soil samples at stepout locations (where needed) to further define lateral and vertical extent of contamination, conduction of discrete field screening, and logging of soil borings and collection of geotechnical samples to assess soil characteristics. Historical information indicates that significant quantities of radioactive material were produced during the rocket engine testing program, some of which was disposed of in radioactive waste disposal systems (posted leachfields) at each of these locations. Process and sanitary effluents were generated and disposed of in other leachfields. The results of this field investigation will be used to develop and evaluate corrective action alternatives for these CASs.

  11. Comprehensive baseline environmental audit of former underground test areas in Colorado, Nevada, and New Mexico

    SciTech Connect (OSTI)

    Not Available

    1994-05-01

    This report documents the results of the Comprehensive Baseline Environmental Audit of Former Underground Test Areas (FUTAS) in the States of Colorado, Nevada, and New Mexico. DOE and contractor systems for management of environmental protection activities on the Nevada Test Site (NTS) were not within the scope of the audit. The audit was conducted May 16-May 26, 1994, by the Office of Environmental Audit (EH-24). DOE 5482.1 B, {open_quotes}Environment, Safety, and Health Appraisal Program{close_quotes}, establishes the mission of EH-24, which is to provide comprehensive, independent oversight of Department-wide environmental programs on behalf of the Secretary of Energy. The ultimate goal of EH-24 is to enhance environmental protection and minimize risk to public health and the environment. EH-24 accomplishes its mission using systematic and periodic evaluations of DOE`s environmental programs within line organizations and supplemental activities that strengthen self-assessment and oversight functions within program, field, and contractor organizations. These evaluations function as a vehicle through which the Secretary and program managers are apprised of the status and vulnerabilities of Departmental environmental activities and environmental management systems. Several types of evaluations are conducted, including: (1) comprehensive baseline environmental audits; (2) routine environmental audits; (3) environmental management assessments; and (4) special issue reviews.

  12. Characterization ReportOperational Closure Covers for the Area 5 Radioactive Waste Management Site at the Nevada Test Site

    SciTech Connect (OSTI)

    Bechtel Nevada Geotechnical Sciences

    2005-06-01

    Bechtel Nevada (BN) manages two low-level Radioactive Waste Management Sites (RWMSs) at the Nevada Test Site (NTS) for the U.S. Department of Energy (DOE) National Nuclear Security Administration Nevada Site Office (NNSA/NSO). The Area 3 RWMS is located in south-central Yucca Flat and the Area 5 RWMS is located about 15 miles south, in north-central Frenchman Flat. Though located in two separate topographically closed basins, they are similar in climate and hydrogeologic setting. The Area 5 RWMS uses engineered shallow-land burial cells to dispose of packaged waste, while the Area 3 RWMS uses subsidence craters formed from underground testing of nuclear weapons for the disposal of packaged and unpackaged bulk waste. Over the next several decades, most waste disposal units at both the Area 3 and Area 5 RWMSs are anticipated to be closed. Closure of the Area 3 and Area 5 RWMSs will proceed through three phases: operational closure, final closure, and institutional control. Many waste disposal units at the Area 5RWMS are operationally closed and final closure has been placed on one unit at the Area 3 RWMS (U-3ax/bl). Because of the similarities between the two sites (e.g., type of wastes, environmental factors, operational closure cover designs, etc.), many characterization studies and data collected at the Area 3 RWMS are relevant and applicable to the Area 5 RWMS. For this reason, data and closure strategies from the Area 3 RWMS are referred to as applicable. This document is an interim Characterization Report – Operational Closure Covers, for the Area 5 RWMS. The report briefly describes the Area 5 RWMS and the physical environment where it is located, identifies the regulatory requirements, reviews the approach and schedule for closing, summarizes the monitoring programs, summarizes characterization studies and results, and then presents conclusions and recommendations.

  13. 2012 Groundwater Monitoring Report Central Nevada Test Area, Subsurface Corrective Action Unit 443

    SciTech Connect (OSTI)

    None

    2013-04-01

    The Central Nevada Test Area was the site of a 0.2- to 1-megaton underground nuclear test in 1968. The surface of the site has been closed, but the subsurface is still in the corrective action process. The corrective action alternative selected for the site was monitoring with institutional controls. Annual sampling and hydraulic head monitoring are conducted as part of the subsurface corrective action strategy. The site is currently in the fourth year of the 5-year proof-of-concept period that is intended to validate the compliance boundary. Analytical results from the 2012 monitoring are consistent with those of previous years. Tritium remains at levels below the laboratory minimum detectable concentration in all wells in the monitoring network. Samples collected from reentry well UC-1-P-2SR, which is not in the monitoring network but was sampled as part of supplemental activities conducted during the 2012 monitoring, indicate concentrations of tritium that are consistent with previous sampling results. This well was drilled into the chimney shortly after the detonation, and water levels continue to rise, demonstrating the very low permeability of the volcanic rocks. Water level data from new wells MV-4 and MV-5 and recompleted well HTH-1RC indicate that hydraulic heads are still recovering from installation and testing. Data from wells MV-4 and MV-5 also indicate that head levels have not yet recovered from the 2011 sampling event during which several thousand gallons of water were purged. It has been recommended that a low-flow sampling method be adopted for these wells to allow head levels to recover to steady-state conditions. Despite the lack of steady-state groundwater conditions, hydraulic head data collected from alluvial wells installed in 2009 continue to support the conceptual model that the southeast-bounding graben fault acts as a barrier to groundwater flow at the site.

  14. CLOSURE REPORT FOR CORRECTIVE ACTION UNIT165: AREA 25 AND 26 DRY WELL AND WASH DOWN AREAS, NEVADA TEST SITE, NEVADA

    SciTech Connect (OSTI)

    BECHTEL NEVADA

    2005-12-01

    This Closure Report (CR) documents the closure activities for Corrective Action Unit (CAU) 165, Area 25 and 26 Dry Well and Washdown Areas, according to the Federal Facility Agreement and Consent Order (FFACO) of 1996. CAU 165 consists of 8 Corrective Action Sites (CASs) located in Areas 25 and 26 of the Nevada Test Site (NTS). The NTS is located approximately 105 kilometers (65 miles) northwest of Las Vegas, nevada. Site closure activities were performed according to the Nevada Division of Environmental Protection (NDEP)-approved Corrective Action Plan (CAP) for CAU 165. CAU 165 consists of the following CASs: (1) CAS 25-07-06, Train Decontamination Area; (2) CAS 25-07-07, Vehicle Washdown; (3) CAS 25-20-01, Lab Drain Dry Well; (4) CAS 25-47-01, Reservoir and French Drain; (5) CAS 25-51-02, Drywell; (6) CAS 25-59-01, Septic System; (7) CAS 26-07-01, Vehicle Washdown Station; and (8) CAS 26-59-01, Septic System. CAU 165, Area 25 and 26 Dry Well and Washdown Areas, consists of eight CASs located in Areas 25 and 26 of the NTS. The approved closure alternatives included No Further Action, Clean Closure, and Closure in Place with Administrative Controls.

  15. Sealed head access area enclosure

    DOE Patents [OSTI]

    Golden, Martin P. (Trafford, PA); Govi, Aldo R. (Greensburg, PA)

    1978-01-01

    A liquid-metal-cooled fast breeder power reactor is provided with a sealed head access area enclosure disposed above the reactor vessel head consisting of a plurality of prefabricated structural panels including a center panel removably sealed into position with inflatable seals, and outer panels sealed into position with semipermanent sealant joints. The sealant joints are located in the joint between the edge of the panels and the reactor containment structure and include from bottom to top an inverted U-shaped strip, a lower layer of a room temperature vulcanizing material, a separator strip defining a test space therewithin, and an upper layer of a room temperature vulcanizing material. The test space is tapped by a normally plugged passage extending to the top of the enclosure for testing the seal or introducing a buffer gas thereinto.

  16. Materials testing and development of functionally graded composite fuel cladding and piping for the Lead-Bismuth cooled nuclear reactor

    E-Print Network [OSTI]

    Fray, Elliott Shepard

    2013-01-01

    This study has extended the development of an exciting technology which promises to enable the Pb-Bi eutectic cooled reactors to operate at temperatures up to 650-700°C. This new technology is a functionally graded composite ...

  17. A Small Area In-Situ MEMS Test Structure to Accurately Measure Fracture Strength by Electrostatic Probing

    SciTech Connect (OSTI)

    Bitsie, Fernando; Jensen, Brian D.; de Boer, Maarten

    1999-07-15

    We have designed, fabricated, tested and modeled a first generation small area test structure for MEMS fracture studies by electrostatic rather than mechanical probing. Because of its small area, this device has potential applications as a lot monitor of strength or fatigue of the MEMS structural material. By matching deflection versus applied voltage data to a 3-D model of the test structure, we develop high confidence that the local stresses achieved in the gage section are greater than 1 GPa. Brittle failure of the polycrystalline silicon was observed.

  18. F and H Area Effluent Treatment Facility (F/H ETF): ultrafiltration and hyperfiltration systems testing at Carre, Inc. with simulated F and H area effluents

    SciTech Connect (OSTI)

    Ryan, J.P.

    1984-05-23

    The F and H Area Effluent Treatment Facility is essentially a four-stage process that will decontaminate the waste water that is currently being discharged to seepage basins in the Separations Areas. The stages include pretreatment, reverse osmosis, ion exchange, and evaporation. A series of tests were performed at Carre, Inc. (Seneca, SC) from March 5 through March 13, to determine the usefulness of ultrafiltration (UF) in the pretreatment stage of the ETF. The results of that testing program indicate that UF would be an excellent means of removing entrained activity from the 200 Area process effluents. Hyperfiltration (HF) was also tested as a means of providing an improved concentration factor from the reverse osmosis stage. The results show that the membranes that were tested would not reject salt well enough at high salt concentrations to be useful in the final reverse osmosis stage. However, there are several membranes which are commercially available that would provide the needed rejection if they could be applied (dynamically) on the Carre support structure. This avenue is still being explored, as theoretically, it could eliminate the need for the F/H ETF evaporator.

  19. 03/01/2006 09:51 AMLoading "People's Daily Online --Chinese experimental thermonuclear reactor on discharge test in July" Page 1 of 1http://english.people.com.cn/200603/01/print20060301_247035.html

    E-Print Network [OSTI]

    03/01/2006 09:51 AMLoading "People's Daily Online -- Chinese experimental thermonuclear reactor experimental thermonuclear reactor on discharge test in July China's new generation experimental Tokamak fusion and the former Soviet Union launched a 10 billion- euro ambitious plan, the International Thermonuclear

  20. Corrective Action Investigation Plan for Corrective Action Unit 165: Areas 25 and 26 Dry Well and Washdown Areas, Nevada Test Site, Nevada (including Record of Technical Change Nos. 1, 2, and 3) (January 2002, Rev. 0)

    SciTech Connect (OSTI)

    U.S. Department of Energy, National Nuclear Security Administration Nevada Operations Office (NNSA/NV)

    2002-01-09

    This Corrective Action Investigation Plan contains the U.S. Department of Energy, National Nuclear Security Administration Nevada Operations Office's approach to collect the data necessary to evaluate corrective action alternatives appropriate for the closure of Corrective Action Unit (CAU) 165 under the Federal Facility Agreement and Consent Order. Corrective Action Unit 165 consists of eight Corrective Action Sites (CASs): CAS 25-20-01, Lab Drain Dry Well; CAS 25-51-02, Dry Well; CAS 25-59-01, Septic System; CAS 26-59-01, Septic System; CAS 25-07-06, Train Decontamination Area; CAS 25-07-07, Vehicle Washdown; CAS 26-07-01, Vehicle Washdown Station; and CAS 25-47-01, Reservoir and French Drain. All eight CASs are located in the Nevada Test Site, Nevada. Six of these CASs are located in Area 25 facilities and two CASs are located in Area 26 facilities. The eight CASs at CAU 165 consist of dry wells, septic systems, decontamination pads, and a reservoir. The six CASs in Area 25 are associated with the Nuclear Rocket Development Station that operated from 1958 to 1973. The two CASs in Area 26 are associated with facilities constructed for Project Pluto, a series of nuclear reactor tests conducted between 1961 to 1964 to develop a nuclear-powered ramjet engine. Based on site history, the scope of this plan will be a two-phased approach to investigate the possible presence of hazardous and/or radioactive constituents at concentrations that could potentially pose a threat to human health and the environment. The Phase I analytical program for most CASs will include volatile organic compounds, semivolatile organic compounds, Resource Conservation and Recovery Act metals, total petroleum hydrocarbons, polychlorinated biphenyls, and radionuclides. If laboratory data obtained from the Phase I investigation indicates the presence of contaminants of concern, the process will continue with a Phase II investigation to define the extent of contamination. Based on the results of Phase I sampling, the analytical program for Phase II investigation may be reduced. The results of this field investigation will support a defensible evaluation of corrective action alternatives in the corrective action decision document.

  1. Closure Report for Corrective Action Unit 536: Area 3 Release Site, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2007-06-01

    Corrective Action Unit (CAU) 536 is located in Area 3 of the Nevada Test Site. CAU 536 is listed in the Federal Facility Agreement and Consent Order of 1996 as Area 3 Release Site, and comprises a single Corrective Action Site (CAS): {sm_bullet} CAS 03-44-02, Steam Jenny Discharge The Nevada Division of Environmental Protection (NDEP)-approved corrective action alternative for CAS 03-44-02 is clean closure. Closure activities included removing and disposing of total petroleum hydrocarbon (TPH)- and polyaromatic hydrocarbon (PAH)-impacted soil, soil impacted with plutonium (Pu)-239, and concrete pad debris. CAU 536 was closed in accordance with the NDEP-approved CAU 536 Corrective Action Plan (CAP), with minor deviations as approved by NDEP. The closure activities specified in the CAP were based on the recommendations presented in the CAU 536 Corrective Action Decision Document (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, 2004). This Closure Report documents CAU 536 closure activities. During closure activities, approximately 1,000 cubic yards (yd3) of hydrocarbon waste in the form of TPH- and PAH-impacted soil and debris, approximately 8 yd3 of Pu-239-impacted soil, and approximately 100 yd3 of concrete debris were generated, managed, and disposed of appropriately. Additionally, a previously uncharacterized, buried drum was excavated, removed, and disposed of as hydrocarbon waste as a best management practice. Waste minimization techniques, such as the utilization of laboratory analysis to characterize and classify waste streams, were employed during the performance of closure

  2. Closure Report for Corrective Action Unit 151: Septic Systems and Discharge Area, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2008-04-01

    Corrective Action Unit (CAU) 151 is identified in the Federal Facility Agreement and Consent Order (FFACO) as Septic Systems and Discharge Area. CAU 151 consists of the following eight Corrective Action Sites (CASs), located in Areas 2, 12, and 18 of the Nevada Test Site, approximately 65 miles northwest of Las Vegas, Nevada: (1) CAS 02-05-01, UE-2ce Pond; (2) CAS 12-03-01, Sewage Lagoons (6); (3) CAS 12-04-01, Septic Tanks; (4) CAS 12-04-02, Septic Tanks; (5) CAS 12-04-03, Septic Tank; (6) CAS 12-47-01, Wastewater Pond; (7) CAS 18-03-01, Sewage Lagoon; and (8) CAS 18-99-09, Sewer Line (Exposed). CAU 151 closure activities were conducted according to the FFACO (FFACO, 1996; as amended February 2008) and the Corrective Action Plan for CAU 151 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, 2007) from October 2007 to January 2008. The corrective action alternatives included no further action, clean closure, and closure in place with administrative controls. CAU 151 closure activities are summarized in Table 1. Closure activities generated liquid remediation waste, sanitary waste, hydrocarbon waste, and mixed waste. Waste generated was appropriately managed and disposed. Waste that is currently staged onsite is being appropriately managed and will be disposed under approved waste profiles in permitted landfills. Waste minimization activities included waste characterization sampling and segregation of waste streams. Some waste exceeded land disposal restriction limits and required offsite treatment prior to disposal. Other waste meeting land disposal restrictions was disposed of in appropriate onsite or offsite landfills. Waste disposition documentation is included as Appendix C.

  3. Corrective Action Plan for Corrective Action Unit 417: Central Nevada Test Area Surface, Nevada

    SciTech Connect (OSTI)

    K. Campbell

    2000-04-01

    This Corrective Action Plan provides methods for implementing the approved corrective action alternative as provided in the Corrective Action Decision Document for the Central Nevada Test Area (CNTA), Corrective Action Unit (CAU) 417 (DOE/NV, 1999). The CNTA is located in the Hot Creek Valley in Nye County, Nevada, approximately 137 kilometers (85 miles) northeast of Tonopah, Nevada. The CNTA consists of three separate land withdrawal areas commonly referred to as UC-1, UC-3, and UC-4, all of which are accessible to the public. CAU 417 consists of 34 Corrective Action Sites (CASs). Results of the investigation activities completed in 1998 are presented in Appendix D of the Corrective Action Decision Document (DOE/NV, 1999). According to the results, the only Constituent of Concern at the CNTA is total petroleum hydrocarbons (TPH). Of the 34 CASs, corrective action was proposed for 16 sites in 13 CASs. In fiscal year 1999, a Phase I Work Plan was prepared for the construction of a cover on the UC-4 Mud Pit C to gather information on cover constructibility and to perform site management activities. With Nevada Division of Environmental Protection concurrence, the Phase I field activities began in August 1999. A multi-layered cover using a Geosynthetic Clay Liner as an infiltration barrier was constructed over the UC-4 Mud Pit. Some TPH impacted material was relocated, concrete monuments were installed at nine sites, signs warning of site conditions were posted at seven sites, and subsidence markers were installed on the UC-4 Mud Pit C cover. Results from the field activities indicated that the UC-4 Mud Pit C cover design was constructable and could be used at the UC-1 Central Mud Pit (CMP). However, because of the size of the UC-1 CMP this design would be extremely costly. An alternative cover design, a vegetated cover, is proposed for the UC-1 CMP.

  4. MANHATTAN PROJECT B REACTOR HANFORD WASHINGTON [HANFORD'S HISTORIC B REACTOR (12-PAGE BOOKLET)

    SciTech Connect (OSTI)

    GERBER MS

    2009-04-28

    The Hanford Site began as part of the United States Manhattan Project to research, test and build atomic weapons during World War II. The original 670-square mile Hanford Site, then known as the Hanford Engineer Works, was the last of three top-secret sites constructed in order to produce enriched uranium and plutonium for the world's first nuclear weapons. B Reactor, located about 45 miles northwest of Richland, Washington, is the world's first full-scale nuclear reactor. Not only was B Reactor a first-of-a-kind engineering structure, it was built and fully functional in just 11 months. Eventually, the shoreline of the Columbia River in southeastern Washington State held nine nuclear reactors at the height of Hanford's nuclear defense production during the Cold War era. The B Reactor was shut down in 1968. During the 1980's, the U.S. Department of Energy began removing B Reactor's support facilities. The reactor building, the river pumphouse and the reactor stack are the only facilities that remain. Today, the U.S. Department of Energy (DOE) Richland Operations Office offers escorted public access to B Reactor along a designated tour route. The National Park Service (NPS) is studying preservation and interpretation options for sites associated with the Manhattan Project. A draft is expected in summer 2009. A final report will recommend whether the B Reactor, along with other Manhattan Project facilities, should be preserved, and if so, what roles the DOE, the NPS and community partners will play in preservation and public education. In August 2008, the DOE announced plans to open B Reactor for additional public tours. Potential hazards still exist within the building. However, the approved tour route is safe for visitors and workers. DOE may open additional areas once it can assure public safety by mitigating hazards.

  5. 2009 Groundwater Monitoring Report Central Nevada Test Area, Corrective Action Unit 443

    SciTech Connect (OSTI)

    2010-09-01

    This report presents the 2009 groundwater monitoring results collected by the U.S. Department of Energy (DOE) Office of Legacy Management (LM) for the Central Nevada Test Area (CNTA) Subsurface Corrective Action Unit (CAU) 443. Responsibility for the environmental site restoration of CNTA was transferred from the DOE Office of Environmental Management to LM on October 1, 2006. The environmental restoration process and corrective action strategy for CAU 443 are conducted in accordance with the Federal Facility Agreement and Consent Order entered into by DOE, the U.S. Department of Defense, and the State of Nevada. The corrective action strategy for the site includes proof-of-concept monitoring in support of site closure. This report summarizes investigation activities associated with CAU 443 that were conducted at the site from October 2008 through December 2009. It also represents the first year of the enhanced monitoring network and begins the new 5-year proof-of-concept monitoring period that is intended to validate the compliance boundary

  6. Underground Test Area Activity Quality Assurance Plan Nevada National Security Site, Nevada. Revision 1

    SciTech Connect (OSTI)

    Farnham, Irene; Krenzien, Susan

    2012-10-01

    This Quality Assurance Plan (QAP) provides the overall quality assurance (QA) requirements and general quality practices to be applied to the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO) Underground Test Area (UGTA) activities. The requirements in this QAP are consistent with DOE Order 414.1C, Quality Assurance (DOE, 2005); U.S. Environmental Protection Agency (EPA) Guidance for Quality Assurance Project Plans for Modeling (EPA, 2002); and EPA Guidance on the Development, Evaluation, and Application of Environmental Models (EPA, 2009). NNSA/NSO, or designee, must review this QAP every two years. Changes that do not affect the overall scope or requirements will not require an immediate QAP revision but will be incorporated into the next revision cycle after identification. Section 1.0 describes UGTA objectives, participant responsibilities, and administrative and management quality requirements (i.e., training, records, procurement). Section 1.0 also details data management and computer software requirements. Section 2.0 establishes the requirements to ensure newly collected data are valid, existing data uses are appropriate, and environmental-modeling methods are reliable. Section 3.0 provides feedback loops through assessments and reports to management. Section 4.0 provides the framework for corrective actions. Section 5.0 provides references for this document.

  7. Underground Test Area Activity Quality Assurance Plan Nevada National Security Site, Nevada. Revision 2

    SciTech Connect (OSTI)

    Krenzien, Susan; Farnham, Irene

    2015-06-01

    This Quality Assurance Plan (QAP) provides the overall quality assurance (QA) requirements and general quality practices to be applied to the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO) Underground Test Area (UGTA) activities. The requirements in this QAP are consistent with DOE Order 414.1D, Change 1, Quality Assurance (DOE, 2013a); U.S. Environmental Protection Agency (EPA) Guidance for Quality Assurance Project Plans for Modeling (EPA, 2002); and EPA Guidance on the Development, Evaluation, and Application of Environmental Models (EPA, 2009). If a participant’s requirement document differs from this QAP, the stricter requirement will take precedence. NNSA/NFO, or designee, must review this QAP every two years. Changes that do not affect the overall scope or requirements will not require an immediate QAP revision but will be incorporated into the next revision cycle after identification. Section 1.0 describes UGTA objectives, participant responsibilities, and administrative and management quality requirements (i.e., training, records, procurement). Section 1.0 also details data management and computer software requirements. Section 2.0 establishes the requirements to ensure newly collected data are valid, existing data uses are appropriate, and environmental-modeling methods are reliable. Section 3.0 provides feedback loops through assessments and reports to management. Section 4.0 provides the framework for corrective actions. Section 5.0 provides references for this document.

  8. Well Completion Report for Corrective Action Unit 443 Central Nevada Test Area Nye County, Nevada

    SciTech Connect (OSTI)

    2009-12-01

    The drilling program described in this report is part of a new corrective action strategy for Corrective Action Unit (CAU) 443 at the Central Nevada Test Area (CNTA). The drilling program included drilling two boreholes, geophysical well logging, construction of two monitoring/validation (MV) wells with piezometers (MV-4 and MV-5), development of monitor wells and piezometers, recompletion of two existing wells (HTH-1 and UC-1-P-1S), removal of pumps from existing wells (MV-1, MV-2, and MV-3), redevelopment of piezometers associated with existing wells (MV-1, MV-2, and MV-3), and installation of submersible pumps. The new corrective action strategy includes initiating a new 5-year proof-of-concept monitoring period to validate the compliance boundary at CNTA (DOE 2007). The new 5-year proof-of-concept monitoring period begins upon completion of the new monitor wells and collection of samples for laboratory analysis. The new strategy is described in the Corrective Action Decision Document/Corrective Action Plan addendum (DOE 2008a) that the Nevada Division of Environmental Protection approved (NDEP 2008).

  9. External Peer Review Team Report Underground Testing Area Subproject for Frenchman Flat, Revision 1

    SciTech Connect (OSTI)

    Sam Marutzky

    2010-09-01

    An external peer review was conducted to review the groundwater models used in the corrective action investigation stage of the Underground Test Area (UGTA) subproject to forecast zones of potential contamination in 1,000 years for the Frenchman Flat area. The goal of the external peer review was to provide technical evaluation of the studies and to assist in assessing the readiness of the UGTA subproject to progress to monitoring activities for further model evaluation. The external peer review team consisted of six independent technical experts with expertise in geology, hydrogeology,'''groundwater modeling, and radiochemistry. The peer review team was tasked with addressing the following questions: 1. Are the modeling approaches, assumptions, and model results for Frenchman Flat consistent with the use of modeling studies as a decision tool for resolution of environmental and regulatory requirements? 2. Do the modeling results adequately account for uncertainty in models of flow and transport in the Frenchman Flat hydrological setting? a. Are the models of sufficient scale/resolution to adequately predict contaminant transport in the Frenchman Flat setting? b. Have all key processes been included in the model? c. Are the methods used to forecast contaminant boundaries from the transport modeling studies reasonable and appropriate? d. Are the assessments of uncertainty technically sound and consistent with state-of-the-art approaches currently used in the hydrological sciences? 3. Are the datasets and modeling results adequate for a transition to Corrective Action Unit monitoring studies—the next stage in the UGTA strategy for Frenchman Flat? The peer review team is of the opinion that, with some limitations, the modeling approaches, assumptions, and model results are consistent with the use of modeling studies for resolution of environmental and regulatory requirements. The peer review team further finds that the modeling studies have accounted for uncertainty in models of flow and transport in the Frenchman Flat except for a few deficiencies described in the report. Finally, the peer review team concludes that the UGTA subproject has explored a wide range of variations in assumptions, methods, and data, and should proceed to the next stage with an emphasis on monitoring studies. The corrective action strategy, as described in the Federal Facility Agreement and Consent Order, states that the groundwater flow and transport models for each corrective action unit will consider, at a minimum, the following: • Alternative hydrostratigraphic framework models of the modeling domain. • Uncertainty in the radiological and hydrological source terms. • Alternative models of recharge. • Alternative boundary conditions and groundwater flows. • Multiple permissive sets of calibrated flow models. • Probabilistic simulations of transport using plausible sets of alternative framework and recharge models, and boundary and groundwater flows from calibrated flow models. • Ensembles of forecasts of contaminant boundaries. • Sensitivity and uncertainty analyses of model outputs. The peer review team finds that these minimum requirements have been met. While the groundwater modeling and uncertainty analyses have been quite detailed, the peer review team has identified several modeling-related issues that should be addressed in the next phase of the corrective action activities: • Evaluating and using water-level gradients from the pilot wells at the Area 5 Radioactive Waste Management Site in model calibration. • Re-evaluating the use of geochemical age-dating data to constrain model calibrations. • Developing water budgets for the alluvial and upper volcanic aquifer systems in Frenchman Flat. • Considering modeling approaches in which calculated groundwater flow directions near the water table are not predetermined by model boundary conditions and areas of recharge, all of which are very uncertain. • Evaluating local-scale variations in hydraulic conductivity on the calculated contaminant boundaries. • Evaluat

  10. Emission Testing of Washington Metropolitan Area Transit Authority (WMATA) Natural Gas and Diesel Transit Buses

    SciTech Connect (OSTI)

    Melendez, M.; Taylor, J.; Wayne, W. S.; Smith, D.; Zuboy, J.

    2005-12-01

    An evaluation of emissions of natural gas and diesel buses operated by the Washington Metro Area Transit Authority.

  11. Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

    SciTech Connect (OSTI)

    Richard P. Wells

    2007-03-23

    This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year.

  12. Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

    SciTech Connect (OSTI)

    Wang, R. C.; Xu, Y.; Downar, T.; Hudson, N.

    2012-07-01

    The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

  13. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M. (Plum Borough, PA)

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  14. Underground Test Area Subproject Phase I Data Analysis Task. Volume III - Groundwater Recharge and Discharge Data Documentation Package

    SciTech Connect (OSTI)

    1996-10-01

    Volume III of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the data covering groundwater recharge and discharge. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  15. Underground Test Area Subproject Phase I Data Analysis Task. Volume V - Transport Parameter and Source Term Data Documentation Package

    SciTech Connect (OSTI)

    1996-12-01

    Volume V of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the transport parameter and source term data. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  16. Safety evaluation report related to the Department of Energy`s proposal for the irradiation of lead test assemblies containing tritium-producing burnable absorber rods in commercial light-water reactors. Project Number 697

    SciTech Connect (OSTI)

    1997-05-01

    The NRC staff has reviewed a report, submitted by DOE to determine whether the use of a commercial light-water reactor (CLWR) to irradiate a limited number of tritium-producing burnable absorber rods (TPBARs) in lead test assemblies (LTAs) raises generic issues involving an unreviewed safety question. The staff has prepared this safety evaluation to address the acceptability of these LTAs in accordance with the provision of 10 CFR 50.59 without NRC licensing action. As summarized in Section 10 of this safety evaluation, the staff has identified issues that require NRC review. The staff has also identified a number of areas in which an individual licensee undertaking irradiation of TPBAR LTAs will have to supplement the information in the DOE report before the staff can determine whether the proposed irradiation is acceptable at a particular facility. The staff concludes that a licensee undertaking irradiation of TPBAR LTAs in a CLWR will have to submit an application for amendment to its facility operating license before inserting the LTAs into the reactor.

  17. The use of U/sub 3/Si/sub 2/ dispersed in aluminum in plate-type fuel elements for research and test reactors

    SciTech Connect (OSTI)

    Snelgrove, J.L.; Domagala, R.F.; Hofman, G.L.; Wiencek, T.C.; Copeland, G.L.; Hobbs, R.W.; Senn, R.L.

    1987-10-01

    A high-density fuel based on U/sub 3/Si/sub 2/ dispersed in aluminum has been developed and tested for use in converting plate-type research and test reactors from the use of highly enriched uranium to the use of low-enriched uranium. Results of preirradiation testing and the irradiation and postirradiation examination of miniature fuel plates and full-sized fuel elements are summarized. Swelling of the U/sub 3/Si/sub 2/ fuel particles is a linear function of the fission density in the particle to well beyond the fission density achievable in low-enriched fuels. U/sub 3/Si/sub 2/ particle swelling rate is approximately the same as that of the commonly used UAl/sub x/ fuel particle. The presence of minor amounts of U/sub 3/Si or uranium solid solution in the fuel result in greater, but still acceptable, fuel swelling. Blister threshold temperatures are at least as high as those of currently used fuels. An exothermic reaction occurs near the aluminum melting temperature, but the measured energy releases were low enough not to substantially worsen the consequences of an accident. U/sub 3/Si/sub 2/-aluminum dispersion fuel with uranium densities up to at least 4.8 Mg/m/sup 3/ is a suitable LEU fuel for typical plate-type research and test reactors. 42 refs., 28 figs., 7 tabs.

  18. Multiple missions: The 300 Area in Hanford Site history

    SciTech Connect (OSTI)

    Gerber, M.S.

    1993-09-01

    This report provides an historical overview of the role of the 300 Area buildings at the Hanford Reservation. Topics covered are: Early fuel fabrication at the Hanford site (313 and 314 Buildings); N reactor fuel fabrication in the 300 Area; 305 test pile was Hanford`s first operating reactor; Early process improvement chemical research (321 and 3706 Buildings); Major 1952 and 1953 expansions in the 300 area (325 and 329 Buildings); Early 300 area facilities constructed to support reactor development (326 and 327 Buildings); Hanford site ventures with the peaceful atom (309, 308 and 318 Buildings); Modern 300 Area Buildings; Significant miscellaneous buildings in the 300 area; 300 Area process waste handling and disposal.

  19. Seismic Monitoring of EGS Tests at the Coso Geothermal Area, California, Using Accurate MEQ Locations and Full Moment Tensors

    E-Print Network [OSTI]

    Foulger, G. R.

    of microearthquakes occurring before, during and following Enhanced Geothermal Systems (EGS) experiments in threeSeismic Monitoring of EGS Tests at the Coso Geothermal Area, California, Using Accurate MEQ, julian@usgs.gov Francis C. Monastero Geothermal Program Office, U. S. Navy, China Lake, CA 93555

  20. Final Report - Hydraulic Conductivity with Depth for Underground Test Area (UGTA) Wells

    SciTech Connect (OSTI)

    P. Oberlander; D. McGraw; C. Russell

    2007-10-31

    Hydraulic conductivity with depth has been calculated for Underground Test Area (UGTA) wells in volcanic tuff and carbonate rock. The following wells in volcanic tuff are evaluated: ER-EC-1, ER-EC-2a, ER-EC-4, ER-EC-5, ER-5-4#2, ER-EC-6, ER-EC-7, and ER-EC-8. The following wells in carbonate rock are evaluated: ER-7-1, ER-6-1, ER-6-1#2, and ER-12-3. There are a sufficient number of wells in volcanic tuff and carbonate rock to associate the conductivity values with the specific hydrogeologic characteristics such as the stratigraphic unit, hydrostratigraphic unit, hydrogeologic unit, lithologic modifier, and alteration modifier used to describe the hydrogeologic setting. Associating hydraulic conductivity with hydrogeologic characteristics allows an evaluation of the data range and the statistical distribution of values. These results are relevant to how these units are considered in conceptual models and represented in groundwater models. The wells in volcanic tuff illustrate a wide range of data values and data distributions when associated with specific hydrogeologic characteristics. Hydraulic conductivity data within a hydrogeologic characteristic can display normal distributions, lognormal distributions, semi-uniform distribution, or no identifiable distribution. There can be multiple types of distributions within a hydrogeologic characteristic such as a single stratigraphic unit. This finding has implications for assigning summary hydrogeologic characteristics to hydrostratigraphic and hydrogeologic units. The results presented herein are specific to the hydrogeologic characteristic and to the wells used to describe hydraulic conductivity. The wells in carbonate rock are associated with a fewer number of hydrogeologic characteristics. That is, UGTA wells constructed in carbonate rock have tended to be in similar hydrogeologic materials, and show a wide range in hydraulic conductivity values and data distributions. Associations of hydraulic conductivity and hydrogeologic characteristics are graphically presented even when there are only a few data. This approach benchmarks what is currently known about the association of depth-specific hydraulic conductivity and hydrogeologic characteristics.

  1. Design and Nuclear-Safety Related Simulations of Bare-Pellet Test Irradiations for the Production of Pu-238 in the High Flux Isotope Reactor using COMSOL

    SciTech Connect (OSTI)

    Freels, James D; Jain, Prashant K; Hobbs, Randy W

    2012-01-01

    The Oak Ridge National Laboratory (ORNL)is developing technology to produce plutonium-238 for the National Aeronautics and Space Administration (NASA) as a power source material for powering vehicles while in deep-space[1]. The High Flux Isotope Reactor (HFIR) of ORNL has been utilized to perform test irradiations of incapsulated neptunium oxide (NpO2) and aluminum powder bare pellets for purposes of understanding the performance of the pellets during irradiation[2]. Post irradiation examinations (PIE) are currently underway to assess the effect of temperature, thermal expansion, swelling due to gas production, fission products, and other phenomena

  2. Closure Report for Corrective Action Unit 104: Area 7 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada

    SciTech Connect (OSTI)

    2013-06-27

    This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 104, Area 7 Yucca Flat Atmospheric Test Sites, and provides documentation supporting the completed corrective actions and confirmation that closure objectives for CAU 104 were met. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management. CAU 104 consists of the following 15 Corrective Action Sites (CASs), located in Area 7 of the Nevada National Security Site: · CAS 07-23-03, Atmospheric Test Site T-7C · CAS 07-23-04, Atmospheric Test Site T7-1 · CAS 07-23-05, Atmospheric Test Site · CAS 07-23-06, Atmospheric Test Site T7-5a · CAS 07-23-07, Atmospheric Test Site - Dog (T-S) · CAS 07-23-08, Atmospheric Test Site - Baker (T-S) · CAS 07-23-09, Atmospheric Test Site - Charlie (T-S) · CAS 07-23-10, Atmospheric Test Site - Dixie · CAS 07-23-11, Atmospheric Test Site - Dixie · CAS 07-23-12, Atmospheric Test Site - Charlie (Bus) · CAS 07-23-13, Atmospheric Test Site - Baker (Buster) · CAS 07-23-14, Atmospheric Test Site - Ruth · CAS 07-23-15, Atmospheric Test Site T7-4 · CAS 07-23-16, Atmospheric Test Site B7-b · CAS 07-23-17, Atmospheric Test Site - Climax Closure activities began in October 2012 and were completed in April 2013. Activities were conducted according to the Corrective Action Decision Document/Corrective Action Plan for CAU 104. The corrective actions included No Further Action and Clean Closure. Closure activities generated sanitary waste, mixed waste, and recyclable material. Some wastes exceeded land disposal limits and required treatment prior to disposal. Other wastes met land disposal restrictions and were disposed in appropriate onsite landfills. The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) requests the following: · A Notice of Completion from the Nevada Division of Environmental Protection to NNSA/NFO for closure of CAU 104 · The transfer of CAU 104 from Appendix III to Appendix IV, Closed Corrective Action Units, of the FFACO

  3. Hydrologic Resources Management Program and Underground Test Area Project FY 2001-2002 Progress Report

    SciTech Connect (OSTI)

    Rose, T P; Kersting, A B; Harris, L J; Hudson, G B; Smith, D K; Williams, R W; Loewen, D R; Nelson, E J; Allen, P G; Ryerson, F J; Pawloski, G A; Laue, C A; Moran, J E

    2003-08-15

    This report contains highlights of FY 2001 and 2002 technical studies conducted by the Analytical and Nuclear Chemistry Division (ANCD) at Lawrence Livermore National Laboratory (LLNL) in support of the Hydrologic Resources Management Program (HRMP) and the Underground Test Area (UGTA) Project. These programs are administered by the U.S. Department of Energy, National Nuclear Security Administration, Nevada Site Office (NNSA/NSO) through the Defense Programs and Environmental Restoration Divisions, respectively. HRMP-sponsored work emphasizes the Defense Programs goal of responsible management of natural resources at the NTS, while UGTA-funded work focuses on defining the extent of radionuclide contamination in NTS groundwater resulting from underground nuclear testing. The report is organized on a topical basis, and contains eight chapters that reflect the range of technical work performed by LLNL-ANCD in support of HRMP and UGTA. Chapter 1 describes recent hot well sampling efforts at the NTS, and presents the results of chemical and isotopic analyses of groundwater samples from six near-field wells. These include the Cambric (UE-5n), Bilby (U-3cn PS No.2), Bourbon (UE-7nS), Nash (UE-2ce), Tybo/Benham (ER-20-5 No.3), and Almendro (U-19v PS No.1ds) sites. The data generated by the hot well program is vital to the development and validation of contaminant transport models at the NTS. Chapter 2 discusses the results of xenon isotope measurements of groundwater samples from the six near-field wells described in Chapter 1. This work demonstrates that fission xenon is present in the water at levels that are readily measurable and highlights the significant differences in xenon concentrations and isotopic abundances at different sites. These differences provide insight into the early cooling history of nuclear test cavities, and may assist in predicting the distribution of the source term in the near-field environment. Chapter 3 is an investigation of the distribution and abundance of actinides in a nuclear test cavity and chimney. This work demonstrates that early-time processes can widely disperse actinides at low concentrations outside the melt glass, implying that melt glass dissolution may not be the sole mechanism for the release of actinides to groundwater. The study also provides evidence for the isotopic fractionation of plutonium under the extreme conditions accompanying nuclear explosions. In Chapter 4, X-ray absorption spectroscopy measurements were used to determine the redox state of Fe and U in nuclear melt glass samples from the NTS. Both elements were found to occur in mixed valence states (Fe{sup 2+}/Fe{sup 3+} and U{sup 5+}/U{sup 6+}) in all samples. Comparison of the Fe and U redox states with published redox studies of synthetic glasses suggests that plutonium is predominantly in the Pu{sup 4+} oxidation state in the melt glasses. In Chapter 5, alpha autoradiography is used in a NTS field study to investigate the spatial distribution and transport of actinides in soils, and to help identify the size distribution and morphology of the actinide particles. It was found that {alpha}-emitting radionuclides have moved to at least 39 cm depth in the soil profile, far deeper than expected. The methodology that was developed could easily be applied to other field locations where actinides are dispersed in the soil zone. Chapter 6 summarizes the development of a method for measuring environmental levels of {sup 241}Am on the multi-collector inductively coupled plasma mass spectrometer. The method detection limit of 0.017 pCi/L is about two times lower than the best analyses possible by alpha spectrometry. Chapter 7 describes a chlorine-36 study of vertical groundwater transport processes in Frenchman Flat. Mass balance calculations developed from a {sup 36}Cl mixing model at well ER-5-3 No.2 are used to estimate vertical transport fluxes and average vertical flow velocities through the thick volcanic section underlying the basin. The study also documents the variations in {sup 36}Cl/Cl ratios within the three princ

  4. Idaho National Engineering Laboratory, Test Area North, Hangar 629 -- Photographs, written historical and descriptive data

    SciTech Connect (OSTI)

    1994-12-31

    The report describes the history of the Idaho National Engineering Laboratory`s Hangar 629. The hangar was built to test the possibility of linking jet engine technology with nuclear power. The history of the project is described along with the development and eventual abandonment of the Flight Engine Test hangar. The report contains historical photographs and architectural drawings.

  5. Prototype Tests for the Recovery and Conversion of UF6Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    SciTech Connect (OSTI)

    Del Cul, G.D.

    2000-06-07

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of {approx}11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide (U{sub 3}O{sub 8})], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

  6. Prototype Tests for the Recovery and Conversion of UF6 Chemisorbed in NaF Traps for the Molten Salt Reactor Remediation Project

    SciTech Connect (OSTI)

    Del Cul, G.D.; Icenhour, A.S.; Simmons, D.W.

    2000-04-01

    The remediation of the Molten Salt Reactor Experiment (MSRE) site includes the removal of about 37 kg of uranium. Of that inventory, about 23 kg have already been removed from the piping system and chemisorbed in 25 NaF traps. This material is being stored in Building 3019. The planned recovery of -11 kg of uranium from the fuel salt will generate another 15 to 19 NaF traps. The remaining 2 to 3 kg of uranium are present in activated charcoal beds, which are also scheduled to be removed from the reactor site. Since all of these materials (NaF traps and the uranium-laden charcoal) are not suitable for long-term storage, they will be converted to a chemical form [uranium oxide], which is suitable for long-term storage. This document describes the process that will be used to recover and convert the uranium in the NaF traps into a stable oxide for long-term storage. Included are a description of the process, equipment, test results, and lessons learned. The process was developed for remote operation in a hot cell. Lessons learned from the prototype testing were incorporated into the process design.

  7. Design of an Online Fission Gas Monitoring System for Post-irradiation Examination Heating Tests of Coated Fuel Particles for High-Temperature Gas-Cooled Reactors

    SciTech Connect (OSTI)

    Dawn Scates

    2010-10-01

    A new Fission Gas Monitoring System (FGMS) has been designed at the Idaho National Laboratory (INL) for use of monitoring online fission gas-released during fuel heating tests. The FGMS will be used with the Fuel Accident Condition Simulator (FACS) at the Hot Fuels Examination Facility (HFEF) located at the Materials and Fuels Complex (MFC) within the INL campus. Preselected Advanced Gas Reactor (AGR) TRISO (Tri-isotropic) fuel compacts will undergo testing to assess the fission product retention characteristics under high temperature accident conditions. The FACS furnace will heat the fuel to temperatures up to 2,000şC in a helium atmosphere. Released fission products such as Kr and Xe isotopes will be transported downstream to the FGMS where they will accumulate in cryogenically cooledcollection traps and monitored with High Purity Germanium (HPGe) detectors during the heating process. Special INL developed software will be used to monitor the accumulated fission products and will report data in near real-time. These data will then be reported in a form that can be readily available to the INL reporting database. This paper describes the details of the FGMS design, the control and acqusition software, system calibration, and the expected performance of the FGMS. Preliminary online data may be available for presentation at the High Temperature Reactor (HTR) conference.

  8. The Underground Test Area Project of the Nevada Test Site: Building Confidence in Groundwater Flow and Transport Models at Pahute Mesa Through Focused Characterization Studies

    SciTech Connect (OSTI)

    Pawloski, G A; Wurtz, J; Drellack, S L

    2009-12-29

    Pahute Mesa at the Nevada Test Site contains about 8.0E+07 curies of radioactivity caused by underground nuclear testing. The Underground Test Area Subproject has entered Phase II of data acquisition, analysis, and modeling to determine the risk to receptors from radioactivity in the groundwater, establish a groundwater monitoring network, and provide regulatory closure. Evaluation of radionuclide contamination at Pahute Mesa is particularly difficult due to the complex stratigraphy and structure caused by multiple calderas in the Southwestern Nevada Volcanic Field and overprinting of Basin and Range faulting. Included in overall Phase II goals is the need to reduce the uncertainty and improve confidence in modeling results. New characterization efforts are underway, and results from the first year of a three-year well drilling plan are presented.

  9. Reactor operation environmental information document

    SciTech Connect (OSTI)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  10. Images of Plasma Disruption Effects in the Tokamak Fusion Test Reactor Ricardo J. Maqueda and Glen A. Wurden

    E-Print Network [OSTI]

    's test cell basement. The imager itself was mounted on one of TFTR's periscopes beneath the machine

  11. In-service Inspection Ultrasonic Testing of Reactor Pressure Vessel Welds for Assessing Flaw Density and Size Distribution per 10 CFR 50.61a, Alternate Fracture Toughness Requirements

    SciTech Connect (OSTI)

    Sullivan, Edmund J.; Anderson, Michael T.; Norris, Wallace

    2012-09-17

    Pressurized thermal shock (PTS) events are system transients in a pressurized water reactor (PWR) in which there is a rapid operating temperature cool-down that results in cold vessel temperatures with or without repressurization of the vessel. The rapid cooling of the inside surface of the reactor pressure vessel (RPV) causes thermal stresses that can combine with stresses caused by high pressure. The aggregate effect of these stresses is an increase in the potential for fracture if a pre-existing flaw is present in a material susceptible to brittle failure. The ferritic, low alloy steel of the reactor vessel beltline adjacent to the core, where neutron radiation gradually embrittles the material over the lifetime of the plant, can be susceptible to brittle fracture. The PTS rule, described in the Code of Federal Regulations, Title 10, Section 50.61 (§50.61), “Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events,” adopted on July 23, 1985, establishes screening criteria to ensure that the potential for a reactor vessel to fail due to a PTS event is deemed to be acceptably low. The U.S. Nuclear Regulatory Commission (NRC) completed a research program that concluded that the risk of through-wall cracking due to a PTS event is much lower than previously estimated. The NRC subsequently developed a rule, §50.61a, published on January 4, 2010, entitled “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (75 FR 13). Use of the new rule by licensees is optional. The §50.61a rule differs from §50.61 in that it requires licensees who choose to follow this alternate method to analyze the results from periodic volumetric examinations required by the ASME Code, Section XI, Rules for Inservice Inspection (ISI) of Nuclear Power Plants. These analyses are intended to determine if the actual flaw density and size distribution in the licensee’s reactor vessel beltline welds are bounded by the flaw density and size distribution values used in the PTS technical basis. Under a contract with the NRC, Pacific Northwest National Laboratory (PNNL) has been working on a program to assess the ability of current inservice inspection (ISI)-ultrasonic testing (UT) techniques, as qualified through ASME Code, Appendix VIII, Supplements 4 and 6, to detect small fabrication or inservice-induced flaws located in RPV welds and adjacent base materials. As part of this effort, the investigators have pursued an evaluation, based on the available information, of the capability of UT to provide flaw density/distribution inputs for making RPV weld assessments in accordance with §50.61a. This paper presents the results of an evaluation of data from the 1993 Browns Ferry Nuclear Plant, Unit 3, Spirit of Appendix VIII reactor vessel examination, a comparison of the flaw density/distribution from this data with the distribution in §50.61a, possible reasons for differences, and plans and recommendations for further work in this area.

  12. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    SciTech Connect (OSTI)

    Bess, John D.; Fujimoto, Nozomu

    2014-10-09

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in the experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.

  13. Benchmark Evaluation of Start-Up and Zero-Power Measurements at the High-Temperature Engineering Test Reactor

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Bess, John D.; Fujimoto, Nozomu

    2014-10-09

    Benchmark models were developed to evaluate six cold-critical and two warm-critical, zero-power measurements of the HTTR. Additional measurements of a fully-loaded subcritical configuration, core excess reactivity, shutdown margins, six isothermal temperature coefficients, and axial reaction-rate distributions were also evaluated as acceptable benchmark experiments. Insufficient information is publicly available to develop finely-detailed models of the HTTR as much of the design information is still proprietary. However, the uncertainties in the benchmark models are judged to be of sufficient magnitude to encompass any biases and bias uncertainties incurred through the simplification process used to develop the benchmark models. Dominant uncertainties in themore »experimental keff for all core configurations come from uncertainties in the impurity content of the various graphite blocks that comprise the HTTR. Monte Carlo calculations of keff are between approximately 0.9 % and 2.7 % greater than the benchmark values. Reevaluation of the HTTR models as additional information becomes available could improve the quality of this benchmark and possibly reduce the computational biases. High-quality characterization of graphite impurities would significantly improve the quality of the HTTR benchmark assessment. Simulation of the other reactor physics measurements are in good agreement with the benchmark experiment values. The complete benchmark evaluation details are available in the 2014 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments.« less

  14. Corrective Action Investigation Plan for Corrective Action Unit 321: Area 22 Weather Station Fuel Storage, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    DOE /NV

    1999-01-28

    This Corrective Action Investigation Plan (CAIP) has been developed in accordance with the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the US Department of Energy, Nevada Operations Office (DOE/NV); the State of Nevada Division of Environmental Protection (NDEP); and the US Department of Defense (FFACO, 1996). The CAIP is a document that provides or references all of the specific information for investigation activities associated with Corrective Action Units (CAUs) or Corrective Action Sites (CASs). According to the FFACO (1996), CASs are sites potentially requiring corrective action(s) and may include solid waste management units or individual disposal or release sites. A CAU consists of one or more CASs grouped together based on geography, technical similarity, or agency responsibility for the purpose of determining corrective actions. This CAIP contains the environmental sample collection objectives and the criteria for conducting site investigation activities at the CAU 321 Area 22 Weather Station Fuel Storage, CAS 22-99-05 Fuel Storage Area. For purposes of this discussion, this site will be referred to as either CAU 321 or the Fuel Storage Area. The Fuel Storage Area is located in Area 22 of the Nevada Test Site (NTS). The NTS is approximately 105 kilometers (km) (65 miles [mi]) northwest of Las Vegas, Nevada (Figure 1-1) (DOE/NV, 1996a). The Fuel Storage Area (Figure 1-2) was used to store fuel and other petroleum products necessary for motorized operations at the historic Camp Desert Rock facility which was operational from 1951 to 1958 at the Nevada Test Site, Nevada. The site was dismantled after 1958 (DOE/NV, 1996a).

  15. Offsite environmental monitoring report. Radiation monitoring around United States nuclear test areas, calendar year 1981

    SciTech Connect (OSTI)

    Black, S.C.; Grossman, R.F.; Mullen, A.A.; Potter, G.D.; Smith, D.D.; Hopper, J.L. (comps.)

    1982-08-01

    This report, prepared in accordance with the guidelines in DOE/E-0023 (DOE 1981), covers the program activities conducted around Nevada Test Site (NTS) for calendar year 1981. It contains descriptions of pertinent features of the NTS and its environs, summaries of the dosimetry and sampling methods, analytical procedures, and the analytical results from environmental measurements. Where applicable, dosimetry and sampling data are compared to appropriate guides for external and internal exposures of humans to ionizing radiation. The monitoring networks detected no radioactivity in the various media which could be attributed to US nuclear testing. Small amounts of fission products were detected in air samples as a result of the People's Republic of China nuclear test and atmospheric krypton-85 increased, following the trend beginning in 1960, due to increased use of nuclear technology. Strontium-90 in milk and cesium-137 in meat samples continued the slow decline as observed for the last several years.

  16. Addendum to the Closure Report for Corrective Action Unit 165: Area 25 and 26 Dry Well and Washdown Areas, Nevada Test Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Krauss, Mark J

    2013-10-01

    This document constitutes an addendum to the Closure Report for Corrective Action Unit 165: Area 25 and 26 Dry Well and Washdown Areas, Nevada Test Site, Nevada as described in the document Recommendations and Justifications To Remove Use Restrictions Established under the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office Federal Facility Agreement and Consent Order dated September 2013. The Use Restriction Removal document was approved by the Nevada Division of Environmental Protection on October 16, 2013. The approval of the UR Removal document constituted approval of each of the recommended UR removals. In conformance with the UR Removal document, this addendum consists of: This page that refers the reader to the UR Removal document for additional information The cover, title, and signature pages of the UR Removal document The NDEP approval letter The corresponding section of the UR Removal document This addendum provides the documentation justifying the cancellation of the UR for CAS 25-20-01, Lab Drain Dry Well. This UR was established as part of FFACO corrective actions and was based on the presence of tetrachloroethene contamination at concentrations greater than the action level established at the time of the initial investigation. Although total petroleum hydrocarbon diesel-range organics contamination at concentrations greater than the NDEP action level was present at the site, no hazardous constituents of TPH-DRO exceeded the U.S. Environmental Protection Agency (EPA) Region 9 preliminary remediation goals established at the time of the initial investigation.

  17. Integrated Closure and Monitoring Plan for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site

    SciTech Connect (OSTI)

    S. E. Rawlinson

    2001-09-01

    Bechtel Nevada (BN) manages two low-level Radioactive Waste Management Sites (RWMSs) (one site is in Area 3 and the other is in Area 5) at the Nevada Test Site (NTS) for the U.S. Department of Energy's (DOE's) National Nuclear Security Administration Nevada Operations Office (NNSA/NV). The current DOE Order governing management of radioactive waste is 435.1. Associated with DOE Order 435.1 is a Manual (DOE M 435.1-1) and Guidance (DOE G 435.1-1). The Manual and Guidance specify that preliminary closure and monitoring plans for a low-level waste (LLW) management facility be developed and initially submitted with the Performance Assessment (PA) and Composite Analysis (CA) for that facility. The Manual and Guidance, and the Disposal Authorization Statement (DAS) issued for the Area 3 RWMS further specify that the preliminary closure and monitoring plans be updated within one year following issuance of a DAS. This Integrated Closure and Monitoring Plan (ICMP) fulfills both requirements. Additional updates will be conducted every third year hereafter. This document is an integrated plan for closing and monitoring both RWMSs, and is based on guidance issued in 1999 by the DOE for developing closure plans. The plan does not follow the format suggested by the DOE guidance in order to better accommodate differences between the two RWMSs, especially in terms of operations and site characteristics. The modification reduces redundancy and provides a smoother progression of the discussion. The closure and monitoring plans were integrated because much of the information that would be included in individual plans is the same, and integration provides efficient presentation and program management. The ICMP identifies the regulatory requirements, describes the disposal sites and the physical environment where they are located, and defines the approach and schedule for both closing and monitoring the sites.

  18. Nevada Test Site 2009 Data Report: Groundwater Monitoring Program, Area 5 Radioactive Waste Management Site

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2010-01-19

    This report is a compilation of the groundwater sampling results from the Area 5 Radioactive Waste Management Site (RWMS). The data have been collected since 1993 and include calendar year 2009 results. During 2009, groundwater at each of the three pilot wells was sampled on March 10, 2009, and August 18, 2009, and water levels at each of the three pilot wells were measured on February 17, May 6, August 17, and November 10, 2009. Groundwater samples were analyzed for the following indicators of contamination: pH, specific conductance, total organic carbon, total organic halides, and tritium. Indicators of general water chemistry (cations and anions) were also measured. Results from all samples collected in 2009 were within the limits established by agreement with the Nevada Division of Environmental Protection for each analyte. These data indicate that there has been no measurable impact to the uppermost aquifer from the Area 5 RWMS. There were no significant changes in measured groundwater parameters compared to previous years. The report contains an updated cumulative chronology for the Area 5 RWMS Groundwater Monitoring Program and a brief description of the site hydrogeology.

  19. Composite Analysis for the Area 5 Radioactive Waste Management Site at the Nevada Test Site, Nye County, Nevada

    SciTech Connect (OSTI)

    V. Yucel

    2001-09-01

    This report summarizes the results of a Composite Analysis (CA) for the Area 5 Radioactive Waste Management Site (RWMS). The Area 5 RWMS is a US Department of Energy (DOE)-operated low-level radioactive waste (LLW) management site located in northern Frenchman Flat on the Nevada Test Site (NTS). The Area 5 RWMS has disposed of low-level radioactive waste in shallow unlined pits and trenches since 1960. Transuranic waste (TRU) and high-specific activity waste was disposed in Greater Confinement Disposal (GCD) boreholes from 1983 to 1989. The purpose of this CA is to determine if continuing operation of the Area 5 RWMS poses an acceptable or unacceptable risk to the public considering the total waste inventory and all other interacting sources of radioactive material in the vicinity. Continuing operation of the Area 5 RWMS will be considered acceptable if the total effective dose equivalent (TEDE) is less than 100 mrem in a year. If the TEDE exceeds 30 mrem in a year, a cost-benefit options analysis must be performed to determine if cost-effective management options exist to reduce the dose further. If the TEDE is found to be less than 30 mrem in a year, an analysis may be performed if warranted to determine if doses are as low as reasonably achievable (ALARA).

  20. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    in AP1000 reactor core Test run signals emergence of the next generation in nuclear power reactor analysis tools OAK RIDGE, Tenn., Feb. 18, 2014 - Scientists and engineers...

  1. A versatile elevated-pressure reactor combined with an ultrahigh vacuum surface setup for efficient testing of model and powder catalysts under clean gas-phase conditions

    SciTech Connect (OSTI)

    Morfin, Franck; Piccolo, Laurent [Institut de recherches sur la catalyse et l'environnement de Lyon (IRCELYON), UMR 5256 CNRS and Université Lyon 1, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)] [Institut de recherches sur la catalyse et l'environnement de Lyon (IRCELYON), UMR 5256 CNRS and Université Lyon 1, 2 avenue Albert Einstein, F-69626 Villeurbanne (France)

    2013-09-15

    A small-volume reaction cell for catalytic or photocatalytic testing of solid materials at pressures up to 1000 Torr has been coupled to a surface-science setup used for standard sample preparation and characterization under ultrahigh vacuum (UHV). The reactor and sample holder designs allow easy sample transfer from/to the UHV chamber, and investigation of both planar and small amounts of powder catalysts under the same conditions. The sample is heated with an infrared laser beam and its temperature is measured with a compact pyrometer. Combined in a regulation loop, this system ensures fast and accurate temperature control as well as clean heating. The reaction products are automatically sampled and analyzed by mass spectrometry and/or gas chromatography (GC). Unlike previous systems, our GC apparatus does not use a recirculation loop and allows working in clean conditions at pressures as low as 1 Torr while detecting partial pressures smaller than 10{sup ?4} Torr. The efficiency and versatility of the reactor are demonstrated in the study of two catalytic systems: butadiene hydrogenation on Pd(100) and CO oxidation over an AuRh/TiO{sub 2} powder catalyst.

  2. Geologic and hydrologic records of observation wells, test holes, test wells, supply wells, springs, and surface water stations in the Los Alamos area

    SciTech Connect (OSTI)

    Purtymun, W.D.

    1995-01-01

    Hundreds of holes have been drilled into the Pajarito Plateau and surrounding test areas of the Los Alamos National Laboratory since the end of World War II. They range in depth from a few feet to more than 14,000 ft. The holes were drilled to provide geologic, hydrologic, and engineering information related to development of a water supply, to provide data on the likelihood or presence of subsurface contamination from hazardous and nuclear materials, and for engineering design for construction. The data contained in this report provide a basis for further investigations into the consequences of our past, present, and future interactions with the environment.

  3. Flow Test At Flint Geothermal Area (DOE GTP) | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX ECoopButtePowerEdistoWhiskey flatsInformationFlint Geothermal Area (DOE GTP) Jump

  4. Flow Test At Jemez Pueblo Area (DOE GTP) | Open Energy Information

    Open Energy Info (EERE)

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  5. Flow Test At Lake City Hot Springs Area (Benoit Et Al., 2005) | Open Energy

    Open Energy Info (EERE)

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  6. Flow Test At Lake City Hot Springs Area (Warpinski, Et Al., 2004) | Open

    Open Energy Info (EERE)

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  7. Flow Test At Lassen Volcanic National Park Area (Janik & Mclaren, 2010) |

    Open Energy Info (EERE)

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  8. Flow Test At Lightning Dock Area (Cunniff & Bowers, 2005) | Open Energy

    Open Energy Info (EERE)

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  9. Flow Test At Long Valley Caldera Geothermal Area (Farrar, Et Al., 2003) |

    Open Energy Info (EERE)

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  10. Flow Test At Mcgee Mountain Area (DOE GTP) | Open Energy Information

    Open Energy Info (EERE)

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  11. Flow Test At Neal Hot Springs Geothermal Area (U.S. Geothermal Inc., 2008)

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX ECoopButtePowerEdistoWhiskey flatsInformationFlint Geothermal AreaOpen| Open Energy

  12. Flow Test At New River Area (DOE GTP) | Open Energy Information

    Open Energy Info (EERE)

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  13. Flow Test At Newberry Caldera Area (DOE GTP) | Open Energy Information

    Open Energy Info (EERE)

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  14. Flow Test At Raft River Geothermal Area (1979) | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX ECoopButtePowerEdistoWhiskey flatsInformationFlint Geothermal AreaOpen|

  15. Flow Test At Raft River Geothermal Area (2004) | Open Energy Information

    Open Energy Info (EERE)

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  16. Flow Test At Raft River Geothermal Area (2006) | Open Energy Information

    Open Energy Info (EERE)

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  17. Flow Test At Raft River Geothermal Area (2008) | Open Energy Information

    Open Energy Info (EERE)

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  18. Flow Test At Roosevelt Hot Springs Geothermal Area (Faulder, 1991) | Open

    Open Energy Info (EERE)

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  19. Flow Test At Roosevelt Hot Springs Geothermal Area (Faulder, 1994) | Open

    Open Energy Info (EERE)

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  20. Flow Test At Rye Patch Area (DOE GTP, 2011) | Open Energy Information

    Open Energy Info (EERE)

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  1. Flow Test At Silver Peak Area (DOE GTP) | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX ECoopButtePowerEdistoWhiskey flatsInformationFlint GeothermalSilver Peak Area (DOE

  2. Flow Test At Soda Lake Area (DOE GTP) | Open Energy Information

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIX ECoopButtePowerEdistoWhiskey flatsInformationFlint GeothermalSilver Peak Area

  3. Injectivity Test At Reese River Area (Henkle & Ronne, 2008) | Open Energy

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIXsource History View NewGuam:on Openei |sourceAndInformation Reese River Area (Henkle

  4. Injectivity Test At Vale Hot Springs Area (Combs, Et Al., 1999) | Open

    Open Energy Info (EERE)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on QA:QA J-E-1 SECTION J APPENDIXsource History View NewGuam:on Openei |sourceAndInformation Reese River Area

  5. Joint Minimization of Power and Area in Scan Testing by Scan Cell Reordering Technical Report: UT-CERC-TR-NAT02-1

    E-Print Network [OSTI]

    Touba, Nur A.

    Joint Minimization of Power and Area in Scan Testing by Scan Cell Reordering Technical Report: UT@cs.utexas.edu Abstract This paper describes a technique for minimizing power dissipation that is also capable of reducing the area overhead of the circuit, compared to a random ordering of the scan cells. For a given set of test

  6. Closure Report for Corrective Action Unit 116: Area 25 Test Cell C Facility, Nevada National Security Site, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2011-09-29

    This Closure Report (CR) presents information supporting closure of Corrective Action Unit (CAU) 116, Area 25 Test Cell C Facility. This CR complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; the U.S. Department of Energy (DOE), Environmental Management; the U.S. Department of Defense; and DOE, Legacy Management (FFACO, 1996 [as amended March 2010]). CAU 116 consists of the following two Corrective Action Sites (CASs), located in Area 25 of the Nevada National Security Site: (1) CAS 25-23-20, Nuclear Furnace Piping and (2) CAS 25-41-05, Test Cell C Facility. CAS 25-41-05 consisted of Building 3210 and the attached concrete shield wall. CAS 25-23-20 consisted of the nuclear furnace piping and tanks. Closure activities began in January 2007 and were completed in August 2011. Activities were conducted according to Revision 1 of the Streamlined Approach for Environmental Restoration Plan for CAU 116 (U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office [NNSA/NSO], 2008). This CR provides documentation supporting the completed corrective actions and provides data confirming that closure objectives for CAU 116 were met. Site characterization data and process knowledge indicated that surface areas were radiologically contaminated above release limits and that regulated and/or hazardous wastes were present in the facility.

  7. Design and Initial Tests of the Tracker-Converter ofthe Gamma-ray Large Area Space Telescope

    SciTech Connect (OSTI)

    Atwood, W.B.; Bagagli, R.; Baldini, L.; Bellazzini, R.; Barbiellini, G.; Belli, F.; Borden, T.; Brez, A.; Brigida, M.; Caliandro, G.A.; Cecchi, C.; Cohen-Tanugi, J.; De Angelis, A.; Drell, P.; Favuzzi, C.; Fukazawa, Y.; Fusco, P.; Gargano, F.; Germani, S.; Giannitrapani, R.; Giglietto, N.; /UC, Santa Cruz /INFN, Pisa /Pisa U. /INFN, Trieste /INFN, Rome /Rome U.,Tor Vergata /SLAC /INFN, Bari /Bari U. /INFN, Perugia /Perugia U. /Udine U. /Hiroshima U. /NASA, Goddard /Maryland U. /Tokyo Inst. Tech. /INFN, Padua /Padua U. /Pisa, Scuola Normale Superiore

    2007-04-16

    The Tracker subsystem of the Large Area Telescope (LAT) science instrument of the Gamma-ray Large Area Space Telescope (GLAST) mission has been completed and tested. It is the central detector subsystem of the LAT and serves both to convert an incident gamma-ray into an electron-positron pair and to track the pair in order to measure the gamma-ray direction. It also provides the principal trigger for the LAT. The Tracker uses silicon strip detectors, read out by custom electronics, to detect charged particles. The detectors and electronics are packaged, along with tungsten converter foils, in 16 modular, high-precision carbon-composite structures. It is the largest silicon-strip detector system ever built for launch into space, and its aggressive design emphasizes very low power consumption, passive cooling, low noise, high efficiency, minimal dead area, and a structure that is highly transparent to charged particles. The test program has demonstrated that the system meets or surpasses all of its performance specifications as well as environmental requirements. It is now installed in the completed LAT, which is being prepared for launch in early 2008.

  8. Corrective Action Investigation Plan for Corrective Action Unit 232: Area 25 Sewage Lagoons Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    DOE/NV Operations Office

    1999-05-01

    This Corrective Action Investigation Plan (CAIP) has been developed in accordance with the Federal Facility Agreement and Consent Order (FFACO) (1996) that was agreed to by the US Department of Energy, Nevada Operations Office (DOE/NV); the State of Nevada Division of Environmental Protection (NDEP); and the US Department of Defense. The CAIP is a document that provides or references all of the specific information for investigation activities associated with Corrective Action Units (CAUs) or Corrective Action Sites (CASs). According to the FFACO, CASs are sites potentially requiring corrective action(s) and may include solid waste management units or individual disposal or release sites. A CAU consists of one or more CASs grouped together based on geography, technical similarity, or agency responsibility for the purpose of determining corrective actions. This CAIP contains the environmental sample collection objectives and criteria for conducting site investigation activities at CAU 232, Area 25 Sewage Lagoons. Corrective Action Unit 232 consists of CAS 25-03-01, Sewage Lagoon, located in Area 25 of the Nevada Test Site (NTS). The NTS is approximately 65 miles (mi) northwest of Las Vegas, Nevada (Figure 1-1) (DOE/NV, 1996a). The Area 25 Sewage Lagoons (Figure 1-2) (IT, 1999b) are located approximately 0.3 mi south of the Test Cell 'C' (TCC) Facility and were used for the discharge of sanitary effluent from the TCC facility. For purposes of this discussion, this site will be referred to as either CAU 232 or the sewage lagoons.

  9. Development of a fuel-rod simulator and small-diameter thermocouples for high-temperature, high-heat-flux tests in the Gas-Cooled Fast Reactor Core Flow Test Loop

    SciTech Connect (OSTI)

    McCulloch, R.W.; MacPherson, R.E.

    1983-03-01

    The Core Flow Test Loop was constructed to perform many of the safety, core design, and mechanical interaction tests in support of the Gas-Cooled Fast Reactor (GCFR) using electrically heated fuel rod simulators (FRSs). Operation includes many off-normal or postulated accident sequences including transient, high-power, and high-temperature operation. The FRS was developed to survive: (1) hundreds of hours of operation at 200 W/cm/sup 2/, 1000/sup 0/C cladding temperature, and (2) 40 h at 40 W/cm/sup 2/, 1200/sup 0/C cladding temperature. Six 0.5-mm type K sheathed thermocouples were placed inside the FRS cladding to measure steady-state and transient temperatures through clad melting at 1370/sup 0/C.

  10. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 398: Area 25 Spill Sites, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    K. B. Campbell

    2001-11-01

    This Streamlined Approach for Environmental Restoration (SAFER) plan addresses the activities necessary to close Corrective Action Unit (CAU) 398: Area 25 Spill Sites. CAU 398, located in Area 25 of the Nevada Test Site, is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) (FFACO, 1996), and consists of the following 13 Corrective Action Sites (CASs) (Figure 1): (1) CAS 25-44-01 , a fuel spill on soil that covers a concrete pad. The origins and use of the spill material are unknown, but the spill is suspected to be railroad bedding material. (2) CAS 25-44-02, a spill of liquid to the soil from leaking drums. (3) CAS 25-44-03, a spill of oil from two leaking drums onto a concrete pad and surrounding soil. (4) CAS 25-44-04, a spill from two tanks containing sulfuric acid and sodium hydroxide used for a water demineralization process. (5) CAS 25-25-02, a fuel or oil spill from leaking drums that were removed in 1992. (6) CAS 25-25-03, an oil spill adjacent to a tipped-over drum. The source of the drum is not listed, although it is noted that the drum was removed in 1991. (7) CAS 25-25-04, an area on the north side of the Engine-Maintenance, Assembly, and Disassembly (E-MAD) facility, where oils and cooling fluids from metal machining operations were poured directly onto the ground. (8) CAS 25-25-05, an area of oil and/or hydraulic fluid spills beneath the heavy equipment once stored there. (9) CAS 25-25-06, an area of diesel fuel staining beneath two generators that have since been removed. (10) CAS 25-25-07, an area of hydraulic oil spills associated with a tunnel-boring machine abandoned inside X-Tunnel. (11) CAS 25-25-08, an area of hydraulic fluid spills associated with a tunnel-boring machine abandoned inside Y-Tunnel. (12) CAS 25-25-16, a diesel fuel spill from an above-ground storage tank located near Building 3320 at Engine Test Stand-1 (ETS-1) that was removed in 1998. (13) CAS 25-25-17, a hydraulic oil spill associated with the historical operations of a vacuum pump oil recovery system at the E-MAD facility.

  11. CLOSURE REPORT FOR CORRECTIVE ACTION UNIT 390: AREAS 9, 10, AND 12 SPILL SITES, NEVADA TEST SITE, NEVADA

    SciTech Connect (OSTI)

    2005-10-01

    Corrective Action Unit (CAU) 390 consists four Corrective Action Sites (CASs) located in Areas 9, 10, and 12 of the Nevada Test Site. The closure activities performed at the CASs include: (1) CAS 09-99-03, Wax, Paraffin: 2 cubic yards of drilling polymer was removed on June 20,2005, and transported to the Area 9 Landfill for disposal. (2) CAS 10-99-01, Epoxy Tar Spill: 2 cubic feet of asphalt waste was removed on June 20,2005, and transported to the Area 9 Landfill for disposal. (3) CAS 10-99-03, Tar Spills: 3 cubic yards of deteriorated asphalt waste was removed on June 20,2005, and transported to the Area 9 Landfill for disposal. (4) CAS 12-25-03, Oil Stains (2); Container: Approximately 16 ounces of used oil were removed from ventilation equipment on June 28,2005, and recycled. One CAS 10-22-19, Drums, Stains, was originally part of CAU 390 but was transferred out of CAU 390 and into CAU 550, Drums, Batteries, and Lead Materials. The transfer was approved by the Nevada Division of Environmental Protection on August 19,2005, and a copy of the approval letter is included in Appendix D of this report.

  12. Closure Report for Corrective Action Unit 481: Area 12 T-Tunnel Conditional Release Storage Yard, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2008-11-01

    Corrective Action Unit (CAU) 481 is identified in the Federal Facility Agreement and Consent Order (FFACO) as Area 12 T-Tunnel Conditional Release Storage Yard. CAU 481 is located in Area 12 of the Nevada Test Site, which is approximately 65 miles northwest of Las Vegas, Nevada. This CAU consists of one Corrective Action Site (CAS), CAS 12-42-05, Housekeeping Waste. CAU 481 closure activities were conducted by the Defense Threat Reduction Agency from August 2007 through July 2008 according to the FFACO and Revision 3 of the Sectored Clean-up Work Plan for Housekeeping Category Waste Sites. Closure activities included removal and disposal of construction debris and low-level waste. Drained fluids, steel, and lead was recycled as appropriate. Waste generated during closure activities was appropriately managed and disposed.

  13. Addendum to the Closure Report for Corrective Action Unit 339: Area 12 Fleet Operations Steam Cleaning Discharge Area, Nevada Test Site, Revision 0

    SciTech Connect (OSTI)

    Grant Evenson

    2009-05-01

    This document constitutes an addendum to the Closure Report for CAU 339: Area 12 Fleet Operations Steam Cleaning Discharge Area Nevada Test Site, December 1997 as described in the document Supplemental Investigation Report for FFACO Use Restrictions, Nevada Test Site, Nevada (SIR) dated November 2008. The SIR document was approved by NDEP on December 5, 2008. The approval of the SIR document constituted approval of each of the recommended UR removals. In conformance with the SIR document, this addendum consists of: • This page that refers the reader to the SIR document for additional information • The cover, title, and signature pages of the SIR document • The NDEP approval letter • The corresponding section of the SIR document This addendum provides the documentation justifying the cancellation of the UR for CAS 12-19-01, A12 Fleet Ops Steam Cleaning Efflu. This UR was established as part of a Federal Facility Agreement and Consent Order (FFACO) corrective action and is based on the presence of contaminants at concentrations greater than the action levels established at the time of the initial investigation (FFACO, 1996). Since this UR was established, practices and procedures relating to the implementation of risk-based corrective actions (RBCA) have changed. Therefore, this UR was reevaluated against the current RBCA criteria as defined in the Industrial Sites Project Establishment of Final Action Levels (NNSA/NSO, 2006). This re-evaluation consisted of comparing the original data (used to define the need for the UR) to risk-based final action levels (FALs) developed using the current Industrial Sites RBCA process. The re-evaluation resulted in a recommendation to remove the UR because contamination is not present at the site above the risk-based FALs. Requirements for inspecting and maintaining this UR will be canceled, and the postings and signage at this site will be removed. Fencing and posting may be present at this site that are unrelated to the FFACO UR such as for radiological control purposes as required by the NV/YMP Radiological Control Manual (NNSA/NSO, 2004). This modification will not affect or modify any non-FFACO requirements for fencing, posting, or monitoring at this site.

  14. Identification and Characterization of Hydrogeologic Units at the Nevada Test Site Using Geophysical Logs: Examples from the Underground Test Area Project

    SciTech Connect (OSTI)

    Lance Prothro, Sigmund Drellack, Margaret Townsend

    2009-03-25

    The diverse and complex geology of the Nevada Test Site region makes for a challenging environment for identifying and characterizing hydrogeologic units penetrated by wells drilled for the U.S. Department of Energy, National Nuclear Security Administration, Underground Test Area (UGTA) Environmental Restoration Sub-Project. Fortunately, UGTA geoscientists have access to large and robust sets of subsurface geologic data, as well as a large historical knowledge base of subsurface geological analyses acquired mainly during the underground nuclear weapons testing program. Of particular importance to the accurate identification and characterization of hydrogeologic units in UGTA boreholes are the data and interpretation principles associated with geophysical well logs. Although most UGTA participants and stakeholders are probably familiar with drill hole data such as drill core and cuttings, they may be less familiar with the use of geophysical logs; this document is meant to serve as a primer on the use of geophysical logs in the UGTA project. Standard geophysical logging tools used in the UGTA project to identify and characterize hydrogeologic units are described, and basic interpretation principles and techniques are explained. Numerous examples of geophysical log data from a variety of hydrogeologic units encountered in UGTA wells are presented to highlight the use and value of geophysical logs in the accurate hydrogeologic characterization of UGTA wells.

  15. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    SciTech Connect (OSTI)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is inadequate to permit steady-state operation at reasonable conditions. 4. To enable the HTTF to operate at a more representative steady-state conditions, DOE recently allocated funding via a DOE subcontract to HTTF to permit an OSU infrastructure upgrade such that 2.2 MW will become available for HTTF experiments. 5. Analyses have been performed to study the relationship between HTTF and MHTGR via the hierarchical two-tiered scaling methodology which has been used successfully in the past, e.g., APEX facility scaling to the Westinghouse AP600 plant. These analyses have focused on the relationship between key variables that will be measured in the HTTF to the counterpart variables in the MHTGR with a focus on natural circulation, using nitrogen as a working fluid, and core heat transfer. 6. Both RELAP5-3D and computational fluid dynamics (CD-Adapco’s STAR-CCM+) numerical models of the MHTGR and the HTTF have been constructed and analyses are underway to study the relationship between the reference reactor and the HTTF. The HTTF is presently being designed. It has Ľ-scaling relationship to the MHTGR in both the height and the diameter. Decisions have been made to design the reactor cavity cooling system (RCCS) simulation as a boundary condition for the HTTF to ensure that (a) the boundary condition is well defined and (b) the boundary condition can be modified easily to achieve the desired heat transfer sink for HTTF experimental operations.

  16. Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 539: Area 25 and Area 26 Railroad Tracks, Nevada Test Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Mark Krauss

    2010-06-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the actions needed to achieve closure for Corrective Action Unit (CAU) 539, Areas 25 and 26 Railroad Tracks, as identified in the Federal Facility Agreement and Consent Order (FFACO). A modification to the FFACOwas approved in May 2010 to transfer the two Railroad Tracks corrective action sites (CASs) from CAU 114 into CAU539. The two CASs are located in Areas 25 and 26 of the Nevada Test Site: • 25-99-21, Area 25 Railroad Tracks • 26-99-05, Area 26 Railroad Tracks This plan provides the methodology for field activities needed to gather the necessary information for closing the two CASs. There is sufficient information and process knowledge from historical documentation and investigations of similar sites regarding the expected nature and extent of potential contaminants to recommend closure of the CAU 539 Railroad Tracks CASs using the SAFER process. Additional information will be obtained by conducting a field investigation before selecting the appropriate corrective action for each CAS. The results of the field investigation should support a defensible recommendation that no further corrective action is necessary. If it is determined that complete clean closure cannot be accomplished during the SAFER, then a hold point will have been reached and the Nevada Division of Environmental Protection (NDEP) will be consulted to determine whether the remaining contamination will be closed under the alternative corrective action of closure in place with use restrictions. This will be presented in a closure report that will be prepared and submitted to the NDEP for review and approval. The sites will be investigated based on the data quality objectives (DQOs) developed on December 14, 2009, by representatives of U.S.Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office; Navarro Nevada Environmental Services, LLC (NNES); and National Security Technologies, LLC. The DQO process has been used to identify and define the type, amount, and quality of data needed to determine and implement appropriate corrective actions for each Railroad Tracks CAS in CAU 539. The following text summarizes the SAFER activities that will support the closure of CAU 539: • Perform site preparation activities (e.g., utilities clearances, radiological surveys). • Collect in situ dose measurements. • Collect environmental samples from designated target populations (e.g., lead bricks) to confirm or disprove the presence of contaminants of concern (COCs) as necessary to supplement existing information. • If no COCs are present at a CAS, establish no further action as the corrective action. • If COCs exist, collect environmental samples from designated target populations (e.g., clean soil adjacent to contaminated soil) and submit for laboratory analyses to define the extent of COC contamination. If a COC is present at a CAS, NNES will consult NDEP to determine the path forward, then either: • Establish clean closure as the corrective action. The material to be remediated will be removed, disposed of as waste, and verification samples will be collected from remaining soil, or • Establish closure in place as the corrective action and implement the appropriate use restrictions.

  17. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    SciTech Connect (OSTI)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  18. Corrective Action Investigation Plan for Corrective Action Unit 552: Area 12 Muckpile and Ponds, Nevada Test Site, Nevada: Revision 0

    SciTech Connect (OSTI)

    U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office

    2004-04-06

    This Corrective Action Investigation Plan contains the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office's approach for collecting the data necessary to evaluate corrective action alternatives appropriate for the closure of Corrective Action Unit (CAU) 552: Area 12 Muckpile and Ponds, Nevada Test Site (NTS), Nevada, under the Federal Facility Agreement and Consent Order. Located in Area 12 on the NTS, CAU 552 consists of two Corrective Action Sites (CASs): 12-06-04, Muckpile; 12-23-05, Ponds. Corrective Action Site 12-06-04 in Area 12 consists of the G-Tunnel muckpile, which is the result of tunneling activities. Corrective Action Site 12-23-05 consists of three dry ponds adjacent to the muckpile. The toe of the muckpile extends into one of the ponds creating an overlap of two CASs. The purpose of the investigation is to ensure that adequate data are collected to provide sufficient and reliable information to identify, evaluate, and select technic ally viable corrective actions. The results of the field investigation will support a defensible evaluation of corrective action alternatives in the corrective action decision document.

  19. Native Plant Uptake Model for Radioactive Waste Disposal Areas at the Nevada Test Site

    SciTech Connect (OSTI)

    BROWN,THERESA J.; WIRTH,SHARON

    1999-09-01

    This report defines and defends the basic framework, methodology, and associated input parameters for modeling plant uptake of radionuclides for use in Performance Assessment (PA) activities of Radioactive Waste Management Sites (RWMS) at the Nevada Test Site (NTS). PAs are used to help determine whether waste disposal configurations meet applicable regulatory standards for the protection of human health, the environment, or both. Plants adapted to the arid climate of the NTS are able to rapidly capture infiltrating moisture. In addition to capturing soil moisture, plant roots absorb nutrients, minerals, and heavy metals, transporting them within the plant to the above-ground biomass. In this fashion, plant uptake affects the movement of radionuclides. The plant uptake model presented reflects rooting characteristics important to plant uptake, biomass turnover rates, and the ability of plants to uptake radionuclides from the soil. Parameters are provided for modeling plant uptake and estimating surface contaminant flux due to plant uptake under both current and potential future climate conditions with increased effective soil moisture. The term ''effective moisture'' is used throughout this report to indicate the soil moisture that is available to plants and is intended to be inclusive of all the variables that control soil moisture at a site (e.g., precipitation, temperature, soil texture, and soil chemistry). Effective moisture is a concept used to simplify a number of complex, interrelated soil processes for which there are too little data to model actual plant available moisture. The PA simulates both the flux of radionuclides across the land surface and the potential dose to humans from that flux. Surface flux is modeled here as the amount of soil contamination that is transferred from the soil by roots and incorporated into aboveground biomass. Movement of contaminants to the surface is the only transport mechanism evaluated with the model presented here. Parameters necessary for estimating surface contaminant flux due to native plants expected to inhabit the NTS RWMSS are developed in this report. The model is specific to the plant communities found at the NTS and is designed for both short-term (<1,000 years) and long-term (>1,000 years) modeling efforts. While the model has been crafted for general applicability to any NTS PA, the key radionuclides considered are limited to the transuranic (TRU) wastes disposed of at the NTS.

  20. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 116: Area 25 Test Cell C Facility, Nevada Test Site, Nevada, Revision 1

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2008-12-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan identifies the activities required for the closure of Corrective Action Unit (CAU) 116, Area 25 Test Cell C Facility. The Test Cell C (TCC) Facility is located in Area 25 of the Nevada Test Site (NTS) approximately 25 miles northwest of Mercury, Nevada (Figure 1). CAU 116 is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO) of 1996 (as amended February 2008) and consists of two Corrective Action Sites (CASs): (1) CAS 25-23-20, Nuclear Furnace Piping; and (2) CAS 25-41-05, Test Cell C Facility. CAS 25-41-05 is described in the FFACO as the TCC Facility but actually includes Building 3210 and attached concrete shield wall only. CAU 116 will be closed by demolishing Building 3210, the attached concrete shield wall, and the nuclear furnace piping. In addition, as a best management practice (BMP), Building 3211 (moveable shed) will be demolished due to its close proximity to Building 3210. This will aid in demolition and disposal operations. Radiological surveys will be performed on the demolition debris to determine the proper disposal pathway. As much of the demolition debris as space allows will be placed into the Building 3210 basement structure. After filling to capacity with demolition debris, the basement structure will be mounded or capped and closed with administrative controls. Prior to beginning demolition activities and according to an approved Sampling and Analysis Plan (SAP), representative sampling of surface areas that are known, suspected, or have the potential to contain hazardous constituents such as lead or polychlorinated biphenyls (PCBs) will be performed throughout all buildings and structures. Sections 2.3.2, 4.2.2.2, 4.2.2.3, 4.3, and 6.2.6.1 address the methodologies employed that assure the solid debris placed in the basement structure will not contain contaminants of concern (COCs) above hazardous waste levels. The anticipated post-closure-posting requirements for the mounded/capped basement structure, as well as for the entire CAU, are addressed in Section 4.2.10. The site contains radiologically impacted surfaces and hazardous materials. Based on review of the historical information for CAU 116 and recent site inspections, there is sufficient process knowledge to close CAU 116 using the SAFER process. CAUs that may be closed using the SAFER process have conceptual corrective actions that are clearly identified. Consequently, corrective action alternatives can be chosen prior to completing a corrective action investigation, given anticipated investigation results. The SAFER process combines elements of the data quality objective (DQO) process and the observational approach to plan and conduct closure activities. The DQOs are used to identify the problem and define the type and quality of data needed to complete the investigation phase of the SAFER process. The purpose of the investigation phase is to verify the adequacy of existing information used to determine the chosen corrective action. The observational approach provides a framework for managing uncertainty during the planning and decision-making phases of the project. The SAFER process allows for technical decisions to be made based on information gathered during site visits, interviews, meetings, research, and a consensus of opinion by the decontamination and decommissioning (D&D) team members. Any uncertainties are addressed by documented assumptions that are verified by sampling and analysis, data evaluation, onsite observations, and contingency plans, as necessary. Closure activities may proceed simultaneously with site characterization as sufficient data are gathered to confirm or disprove the assumptions made during selection of the corrective action. If, at any time during the closure process, new information is discovered that indicates that closure activities should be revised, closure activities will be reevaluated as appropriate. Based on a detailed review of historical documentation, there is sufficient process know

  1. Closure Report for Corrective Action Unit 408: Bomblet Target Area Tonopah Test Range (TTR), Nevada, Revision 0

    SciTech Connect (OSTI)

    Mark Krauss

    2010-09-01

    This Closure Report (CR) presents information supporting the closure of Corrective Action Unit (CAU) 408: Bomblet Target Area (TTR), Tonopah Test Range, Nevada. This CR complies with the requirements of the Federal Facility Agreement and Consent Order that was agreed to by the State of Nevada; U.S. Department of Energy (DOE), Environmental Management; U.S. Department of Defense; and DOE, Legacy Management. Corrective Action Unit 408 is located at the Tonopah Test Range, Nevada, and consists of Corrective Action Site (CAS) TA-55-002-TAB2, Bomblet Target Areas. This CAS includes the following seven target areas: • Mid Target • Flightline Bomblet Location • Strategic Air Command (SAC) Target Location 1 • SAC Target Location 2 • South Antelope Lake • Tomahawk Location 1 • Tomahawk Location 2 The purpose of this CR is to provide documentation supporting the completed corrective actions and data confirming that the closure objectives for the CAS within CAU 408 were met. To achieve this, the following actions were performed: • Review the current site conditions, including the concentration and extent of contamination. • Implement any corrective actions necessary to protect human health and the environment. • Properly dispose of corrective action and investigation wastes. • Document Notice of Completion and closure of CAU 408 issued by the Nevada Division of Environmental Protection. From July 2009 through August 2010, closure activities were performed as set forth in the Streamlined Approach for Environmental Restoration Plan for CAU 408: Bomblet Target Area, Tonopah Test Range (TTR), Nevada. The purposes of the activities as defined during the data quality objectives process were as follows: • Identify and remove munitions of explosive concern (MEC) associated with DOE activities. • Investigate potential disposal pit locations. • Remove depleted uranium-contaminated fragments and soil. • Determine whether contaminants of concern (COCs) are present. • If COCs are present, determine their nature and extent, implement appropriate corrective actions, and properly dispose of wastes. Analytes detected during the closure activities were evaluated against final action levels to determine COCs for CAU 408. Assessment of the data indicated COCs are not present at CAS TA-55-002-TAB2; therefore, no corrective action is necessary. No use restrictions are required to be placed on this CAU because the investigation showed no evidence of remaining soil contamination or remaining debris/waste upon completion of all investigation activities. The MEC was successfully removed and dispositioned as planned using current best available technologies. As MEC guidance and general MEC standards acknowledge that MEC response actions cannot determine with 100 percent certainty that all MEC and unexploded ordnance (UXO) are removed, the clean closure of CAU 408 will implement a best management practice of posting UXO hazard warning signs near the seven target areas. The signs will warn future land users of the potential for encountering residual UXO hazards. The DOE, National Nuclear Security Administration Nevada Site Office, provides the following recommendations: • A Notice of Completion to the DOE, National Nuclear Security Administration Nevada Site Office, is requested from the Nevada Division of Environmental Protection for closure of CAU 408. • Corrective Action Unit 408 should be moved from Appendix III to Appendix IV of the Federal Facility Agreement and Consent Order.

  2. Underground Test Area Fiscal Year 2012 Annual Quality Assurance Report Nevada National Security Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Farnham, Irene; Marutzky, Sam

    2013-01-01

    This report is mandated by the Underground Test Area (UGTA) Quality Assurance Project Plan (QAPP) and identifies the UGTA quality assurance (QA) activities for fiscal year (FY) 2012. All UGTA organizations—U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO); Desert Research Institute (DRI); Lawrence Livermore National Laboratory (LLNL); Los Alamos National Laboratory (LANL); Navarro-Intera, LLC (N-I); National Security Technologies, LLC (NSTec); and the U.S. Geological Survey (USGS)—conducted QA activities in FY 2012. The activities included conducting assessments, identifying findings and completing corrective actions, evaluating laboratory performance, revising the QAPP, and publishing documents. In addition, processes and procedures were developed to address deficiencies identified in the FY 2011 QAPP gap analysis.

  3. Underground Test Area Fiscal Year 2013 Annual Quality Assurance Report Nevada National Security Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Krenzien, Susan; Marutzky, Sam

    2014-01-01

    This report is required by the Underground Test Area (UGTA) Quality Assurance Plan (QAP) and identifies the UGTA quality assurance (QA) activities for fiscal year (FY) 2013. All UGTA organizations—U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO); Desert Research Institute (DRI); Lawrence Livermore National Laboratory (LLNL); Los Alamos National Laboratory (LANL); Navarro-Intera, LLC (N-I); National Security Technologies, LLC (NSTec); and the U.S. Geological Survey (USGS)—conducted QA activities in FY 2013. The activities included conducting assessments, identifying findings and completing corrective actions, evaluating laboratory performance, and publishing documents. In addition, integrated UGTA required reading and corrective action tracking was instituted.

  4. Corrective Action Investigation Plan for Corrective Action Unit 374: Area 20 Schooner Unit Crater Nevada Test Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Patrick Matthews

    2010-02-01

    Corrective Action Unit 374 is located in Areas 18 and 20 of the Nevada Test Site, which is approximately 65 miles northwest of Las Vegas, Nevada. Corrective Action Unit 374 comprises the five corrective action sites (CASs) listed below: • 18-22-05, Drum • 18-22-06, Drums (20) • 18-22-08, Drum • 18-23-01, Danny Boy Contamination Area • 20-45-03, U-20u Crater (Schooner) These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives (CAAs). Additional information will be obtained by conducting a corrective action investigation before evaluating CAAs and selecting the appropriate corrective action for each CAS. The results of the field investigation will support a defensible evaluation of viable CAAs that will be presented in the Corrective Action Decision Document. The sites will be investigated based on the data quality objectives (DQOs) developed on October 20, 2009, by representatives of the Nevada Division of Environmental Protection and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office. The DQO process was used to identify and define the type, amount, and quality of data needed to develop and evaluate appropriate corrective actions for CAU 374.

  5. Corrective Action Decision Document for Corrective Action Unit 321: Area 22 Weather Station Fuel Storage, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    U.S. Department of Energy, Nevada Operations Office

    1999-07-22

    This Corrective Action Decision Document identifies and rationalizes the U.S. Department of Energy, Nevada Operations Office's selection of a recommended corrective action alternative appropriate to facilitate the closure of Corrective Action Unit (CAU) 321, Weather Station Fuel Storage, under the Federal Facility Agreement and Consent Order. Corrective Action Unit 321 is located at the Nevada Test Site (NTS) in Area 22, and consists of a single Corrective Action Site (CAS) 22-99-05, Fuel Storage Area. This CAS contains a fuel storage area approximately 325 by 540 feet, which was used to store fuel and other petroleum products necessary for motorized operations at the historical Camp Desert Rock facility, which was operational from 1951 to 1958. The corrective action investigation conducted in February 1999 found the only contaminant of concern above preliminary action levels to be total petroleum hydrocarbons as diesel-range organics at two sample locations. During this investigation, the two corrective action objectives identified were (1) to prevent or mitigate exposure to near-surface soil containing contaminants of concern, and (2) to prevent spread of contaminants of concern beyond the corrective action unit. Based on the corrective action objectives, the two corrective action alternatives developed for consideration were: Alternative 1 - No Further Action; and Alternative 2 - Clean Closure by Excavation and Disposal. The two alternatives were evaluated based on four general corrective action standards and five remedy selection decision factors, and the preferred corrective action alternative chosen on technical merit, focusing on performance, reliability, feasibility, and safety was Alternative 2. This alternative meets all applicable state and federal regulations for closure of the site and will eliminate potential future exposure pathways to the contaminated soils at the Weather Station Fuel Storage site.

  6. Design and Testing of a Labview- Controlled Catalytic Packed- Bed Reactor System For Production of Hydrocarbon Fuels

    SciTech Connect (OSTI)

    Street, J.; Yu, F.; Warnock, J.; Wooten, J.; Columbus, E.; White, M. G.

    2012-05-01

    Gasified woody biomass (producer gas) was converted over a Mo/H+ZSM-5 catalyst to produce gasolinerange hydrocarbons. The effect of contaminants in the producer gas showed that key retardants in the system included ammonia and oxygen. The production of gasoline-range hydrocarbons derived from producer gas was studied and compared with gasoline-range hydrocarbon production from two control syngas mixes. Certain mole ratios of syngas mixes were introduced into the system to evaluate whether or not the heat created from the exothermic reaction could be properly controlled. Contaminant-free syngas was used to determine hydrocarbon production with similar mole values of the producer gas from the gasifier. Contaminant-free syngas was also used to test an ideal contaminant-free synthesis gas situation to mimic our particular downdraft gasifier. Producer gas was used in this study to determine the feasibility of using producer gas to create gasoline-range hydrocarbons on an industrial scale using a specific Mo/H+ZSM-5 catalyst. It was determined that after removing the ammonia, other contaminants poisoned the catalyst and retarded the hydrocarbon production process as well.

  7. 2013 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2014-02-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (#LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Pond from November 1, 2012–October 31, 2013. The report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Groundwater monitoring data • Status of compliance activities • Noncompliance issues • Discussion of the facility’s environmental impacts. During the 2013 permit year, approximately 238 million gallons of wastewater was discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters are below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  8. The role of the neutral beam fueling profile in the performance of the Tokamak Fusion Test Reactor and other tokamak plasmas

    SciTech Connect (OSTI)

    Park, H.K.; Batha, S.; Sabbagh, S.A. |

    1997-02-01

    Scalings for the stored energy and neutron yield, determined from experimental data are applied to both deuterium-only and deuterium-tritium plasmas in different neutral beam heated operational domains in Tokamak Fusion Test Reactor. The domain of the data considered includes the Supershot, High poloidal beta, Low-mode, and limiter High-mode operational regimes, as well as discharges with a reversed magnetic shear configuration. The new important parameter in the present scaling is the peakedness of the heating beam fueling profile shape. Ion energy confinement and neutron production are relatively insensitive to other plasma parameters compared to the beam fueling peakedness parameter and the heating beam power when considering plasmas that are stable to magnetohydrodynamic modes. However, the stored energy of the electrons is independent of the beam fueling peakedness. The implication of the scalings based on this parameter is related to theoretical transport models such as radial electric field shear and Ion Temperature Gradient marginality models. Similar physics interpretation is provided for beam heated discharges on other major tokamaks.

  9. The development and operational testing of an experimental reactor for gas-liquid-solid reaction systems at high temperatures and pressures 

    E-Print Network [OSTI]

    Hess, Richard Kenneth

    1985-01-01

    , but REACTANTS IN P L'MP CATALYST BED PROD L'CT5 OUT Figure 2. A schematic drawing of an external recycle reactor. ARROWS SHOW FLUID FLOW PATTERN TC PORT CATALYST BED TC PORT INLET IMPELLER OUTLET IMPELLER SHAFT Figure 3. A ISerty reactor. 10...

  10. Integrated Closure and Monitoring Plan for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site

    SciTech Connect (OSTI)

    Bechtel Nevada

    2005-06-01

    This document is an integrated plan for closing and monitoring two low-level radioactive waste disposal sites at the Nevada Test Site.

  11. Imaging Fukushima Daiichi reactors with muons

    SciTech Connect (OSTI)

    Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Milner, Edward C.; Morris, Christopher L. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Lukic, Zarija [Computational Cosmology Center, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Masuda, Koji [University of New Mexico, Albuquerque, NM 87131 (United States); Perry, John O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); University of New Mexico, Albuquerque, NM 87131 (United States)

    2013-05-15

    A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  12. RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

    SciTech Connect (OSTI)

    Lee, Sukho; Kim, Manwoong; Kim, Hho-Jung; Lee, John C.

    2005-09-15

    Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test.

  13. Upgrading scientific capabilities at the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    West, C.D.; Farrar, M.B.

    1997-07-14

    Following termination of the Advanced Neutron Source (ANS) Project, a program of upgrades to the Department of Energy`s High Flux Isotope Reactor (HFIR) was devised by a team of researchers and reactor operators and has been proposed to the department. HFIR is a multipurpose research reactor, commissioned in 1965, with missions in four nationally important areas: isotope production, especially transuranic isotopes; neutron scattering; neutron activation analysis; and irradiation testing of materials. For neutron scattering, there are two major enhancements and several smaller ones. The first is the installation of a small, hydrogen cold neutron source in one of the four existing beam tubes: because of the high reactor power, and the use of new design concepts developed for ANS, the cold source will be as bright as, or brighter than, the Institute Laue Langevin liquid deuterium vertical cold source, although space limitations mean that there will be far fewer cold beams and instruments at HFIR. This project is underway, and the cold source is expected to come on line following an extended shutdown in 1999 to replace the reactor`s beryllium reflector. The second major change proposed would put five thermal neutron guides at an existing beam port and construct a new guide hall to accommodate instruments on these very intense beams.

  14. NETL - Chemical Looping Reactor

    SciTech Connect (OSTI)

    2013-07-24

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  15. NETL - Chemical Looping Reactor

    ScienceCinema (OSTI)

    None

    2014-06-26

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  16. A Cultural Resources Inventory and Historical Evaluation of the Smoky Atmospheric Nuclear Test, Areas 8, 9, and 10, Nevada National Security Site, Nye County, Nevada

    SciTech Connect (OSTI)

    Jones, Robert C.; King, Maureen L.; Beck, Colleen M.; Falvey, Lauren W.; Menocal, Tatianna M.

    2014-09-01

    This report presents the results of a National Historic Preservation Act Section 106 cultural resources inventory and historical evaluation of the 1957 Smoky atmospheric test location on the Nevada National Security Site (NNSS). The Desert Research Institute (DRI) was tasked to conduct a cultural resources study of the Smoky test area as a result of a proposed undertaking by the Department of Energy Environmental Management. This undertaking involves investigating Corrective Action Unit (CAU) 550 for potential contaminants of concern as delineated in a Corrective Action Investigation Plan. CAU 550 is an area that spatially overlaps portions of the Smoky test location. Smoky, T-2c, was a 44 kt atmospheric nuclear test detonated at 5:30 am on August 31, 1957, on top of a 213.4 m (700 ft) 200 ton tower (T-2c) in Area 8 of the NNSS. Smoky was a weapons related test of the Plumbbob series (number 19) and part of the Department of Defense Exercise Desert Rock VII and VIII. The cultural resources effort involved the development of a historic context based on archival documents and engineering records, the inventory of the cultural resources in the Smoky test area and an associated military trench location in Areas 9 and 10, and an evaluation of the National Register eligibility of the cultural resources. The inventory of the Smoky test area resulted in the identification of structures, features, and artifacts related to the physical development of the test location and the post-test remains. The Smoky test area was designated historic district D104 and coincides with a historic archaeological site recorded as 26NY14794 and the military trenches designed for troop observation, site 26NY14795. Sites 26NY14794 and 26NY14795 are spatially discrete with the trenches located 4.3 km (2.7 mi) southeast of the Smoky ground zero. As a result, historic district D104 is discontiguous and in total it covers 151.4 hectares (374 acres). The Smoky test location, recorded as historic district D104 and historic sites 26NY14794 and 26NY14795, is the best preserved post-shot atmospheric nuclear tower test at the NNSS and possibly in the world. It is of local, national, and international importance due to nuclear testing’s pivotal role in the Cold War between the United States and the former Soviet Union. The district and sites are linked to the historic theme of atmospheric nuclear testing. D104 retains aspects of the engineering plan and design for the Smoky tower, instrument stations used to measure test effects, German and French personnel shelters, and military trenches. A total of 33 structures contribute to the significance of D104. Artifacts and features provide significant post-test information. Historic district D104 (discontiguous) and historic site 26NY14794 (the Smoky test area) are eligible for listing on the NRHP under Criteria A, B, C, and D. The historic site 26NY14795 (the Smoky military trenches) is eligible for listing under Criteria A, C, and D. Several items have been identified for removal by the CAU 550 investigation. However, none of them is associated with the Smoky atmospheric test, but with later activities in the area. The military trenches are not part of CAU 550 and no actions are planned there. A proposed closure of the Smoky test area with restrictions will limit access and contribute to the preservation of the cultural resources. It is recommended that the Smoky historic district and sites be included in the NNSS cultural resources monitoring program.

  17. Closure Report for Corrective Action Unit 117: Area 26 Pluto Disassembly Facility, Nevada Test Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Mark Burmeister

    2009-06-01

    This Closure Report (CR) presents information supporting the closure of Corrective Action Unit (CAU) 117: Area 26 Pluto Disassembly Facility, Nevada Test Site, Nevada. This CR complies with the requirements of the Federal Facility Agreement and Consent Order that was agreed to by the State of Nevada; U.S. Department of Energy (DOE), Environmental Management; U.S. Department of Defense; and DOE, Legacy Management. Corrective Action Unit 117 comprises Corrective Action Site (CAS) 26-41-01, Pluto Disassembly Facility, located in Area 26 of the Nevada Test Site. The purpose of this CR is to provide documentation supporting the completed corrective actions and provide data confirming that the closure objectives for CAU 117 were met. To achieve this, the following actions were performed: • Review the current site conditions, including the concentration and extent of contamination. • Implement any corrective actions necessary to protect human health and the environment. • Properly dispose of corrective action and investigation wastes. • Document Notice of Completion and closure of CAU 117 issued by the Nevada Division of Environmental Protection. From May 2008 through February 2009, closure activities were performed as set forth in the Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 117, Area 26 Pluto Disassembly Facility, Nevada Test Site, Nevada. The purpose of the activities as defined during the data quality objectives process were: • Determine whether contaminants of concern (COCs) are present. • If COCs are present, determine their nature and extent, implement appropriate corrective actions, and properly dispose of wastes. Analytes detected during the closure activities were evaluated against final action levels to determine COCs for CAU 117. Assessment of the data generated from closure activities indicated that the final action levels were exceeded for polychlorinated biphenyls (PCBs) reported as total Aroclor and radium-226. A corrective action was implemented to remove approximately 50 cubic yards of PCB-contaminated soil, approximately 1 cubic foot of radium-226 contaminated soil (and scabbled asphalt), and a high-efficiency particulate air filter that was determined to meet the criteria of a potential source material (PSM). Electrical and lighting components (i.e., PCB-containing ballasts and capacitors) and other materials (e.g., mercury-containing thermostats and switches, lead plugs and bricks) assumed to be PSM were also removed from Building 2201, as practical, without the need for sampling. Because the COC contamination and PSMs have been removed, clean closure of CAS 26-41-01 is recommended, and no use restrictions are required to be placed on this CAU. No further action is necessary because no other contaminants of potential concern were found above preliminary action levels. The physical end state for Building 2201 is expected to be eventual demolition to slab. The DOE, National Nuclear Security Administration Nevada Site Office provides the following recommendations: • Clean closure is the recommended corrective action for CAS 26-41-01 in CAU 117. • A Notice of Completion to the DOE, National Nuclear Security Administration Nevada Site Office is requested from the Nevada Division of Environmental Protection for closure of CAU 117. • Corrective Action Unit 117 should be moved from Appendix III to Appendix IV of the Federal Facility Agreement and Consent Order.

  18. CORRECTIVE ACTION DECISION DOCUMENT FOR THE AREA 3 LANDFILL COMPLEX, TONOPAH TEST RANGE, CAU 424, REVISION 0, MARCH 1998

    SciTech Connect (OSTI)

    DOE /NV

    1998-03-03

    This Corrective Action Decision Document (CADD) has been prepared for the Area 3 Landfill Complex (Corrective Action Unit [CAU] 424) in accordance with the Federal Facility Agreement and Consent Order (FFACO) of 1996. Corrective Action Unit 424 is located at the Tonopah Test Range (TTR) and is comprised of the following Corrective Action Sites (CASs), each an individual landfill located around and within the perimeter of the Area 3 Compound (DOE/NV, 1996a): (1) Landfill A3-1 is CAS No. 03-08-001-A301. (2) Landfill A3-2 is CAS No. 03-08-002-A302. (3) Landfill A3-3 is CAS No. 03-08-002-A303. (4) Landfill A3-4 is CAS No. 03-08-002-A304. (5) Landfill A3-5 is CAS No. 03-08-002-A305. (6) Landfill A3-6 is CAS No. 03-08-002-A306. (7) Landfill A3-7 is CAS No. 03-08-002-A307. (8) Landfill A3-8 is CAS No. 03-08-002-A308. The purpose of this CADD is to identify and provide a rationale for the selection of a recommended corrective action alternative for each CAS. The scope of this CADD consists of the following: (1) Develop corrective action objectives. (2) Identify corrective action alternative screening criteria. (3) Develop corrective action alternatives. (4) Perform detailed and comparative evaluations of the corrective action alternatives in relation to the corrective action objectives and screening criteria. (6) Recommend and justify a preferred corrective action alternative for each CAS. In June and July 1997, a corrective action investigation was performed as set forth in the Corrective Action Investigation Plan (CAIP) for CAU No. 424: Area 3 Landfill Complex, Tonopah Test Range, Nevada (DOE/NV, 1997). Details can be found in Appendix A of this document. The results indicated four groupings of site characteristics as shown in Table ES-1. Based on the potential exposure pathways, the following corrective action objectives have been identified for CAU No. 424: (1) Prevent or mitigate human exposure to subsurface soils containing waste. (2) Remediate the site per applicable state and federal regulations (NAC, 1996c). (3) Prevent adverse impacts to groundwater quality. Based on the review of existing data, future land use, and current operations at the TTR, the following alternatives were developed for consideration at the Area 3 Landfill Complex CAU: Alternative 1 - No Action; Alternative 2 - Administrative Closure; Alternative 3 - Partial Excavation, Backfill, and Recontouring The corrective action alternatives were evaluated based on four general corrective action standards and five remedy-selection decision factors. Based on the results of this evaluation, preferred alternatives were selected for each CAS as indicated in Table ES-2. The preferred corrective action alternatives were evaluated on their technical merits, focusing on performance, reliability, feasibility, and safety. The alternatives were judged to meet all requirements for the technical components evaluated. These alternatives meet all applicable state and federal regulations for closure of the site and will reduce potential future exposure pathways to the contents of the landfills. During corrective action implementation, these alternatives will present minimal potential threat to site workers who come in contact with the waste. However, procedures will be developed and implemented to ensure worker health and safety.

  19. Packed bed reactor for photochemical sup 196 Hg isotope separation

    SciTech Connect (OSTI)

    Grossman, M.W.; Speer, R.

    1992-03-03

    This patent describes a photochemical reactor useful for the isotopic enrichment of a predetermined isotope of mercury comprising a reactor cell and a monoisotopic light source It comprises: a plurality of transparent, straight reactor cell tubes disposed axially within the internal volume of the reactor cell to increase the surface area thereof for production deposition.

  20. Compact Reactor

    SciTech Connect (OSTI)

    Williams, Pharis E. [Williams Research, P.O. Box 554, Los Alamos, NM87544 (United States)

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  1. Reactor Technology | Nuclear Science | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Reactor Technology Advanced Reactor Concepts Advanced Instrumentation & Controls Light Water Reactor Sustainability Safety and Regulatory Technology Small Modular Reactors Nuclear...

  2. Consortium for Advanced Simulation of Light Water Reactors (CASL...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    to achieve challenge problem solutions A strong VERA infrastructure supporting software development, testing, and releases. Requirements Drivers Modeling of reactors...

  3. Experimental Validation of Passive Safety System Models: Application to Design and Optimization of Fluoride-Salt-Cooled, High-Temperature Reactors

    E-Print Network [OSTI]

    Zweibaum, Nicolas

    2015-01-01

    test SFR – Sodium-cooled fast reactor SWU – Separative workvalues than sodium-cooled fast reactors ( SFRs) and HTGRs.

  4. Massive Hanford Test Reactor Removed - Plutonium Recycle Test Reactor

    Broader source: Energy.gov (indexed) [DOE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirleyEnergyTher i nAand DOE Safetyof Energy ThisSites |andofMassachusetts --As the

  5. Annex D-200 Area Interim Storage Area Final Safety Analysis Report [FSAR] [Section 1 & 2

    SciTech Connect (OSTI)

    CARRELL, R D

    2002-07-16

    The 200 Area Interim Storage Area (200 Area ISA) at the Hanford Site provides for the interim storage of non-defense reactor spent nuclear fuel (SNF) housed in aboveground dry cask storage systems. The 200 Area ISA is a relatively simple facility consisting of a boundary fence with gates, perimeter lighting, and concrete and gravel pads on which to place the dry storage casks. The fence supports safeguards and security and establishes a radiation protection buffer zone. The 200 Area ISA is nominally 200,000 ft{sup 2} and is located west of the Canister Storage Building (CSB). Interim storage at the 200 Area ISA is intended for a period of up to 40 years until the materials are shipped off-site to a disposal facility. This Final Safety Analysis Report (FSAR) does not address removal from storage or shipment from the 200 Area ISA. Three different SNF types contained in three different dry cask storage systems are to be stored at the 200 Area ISA, as follows: (1) Fast Flux Test Facility Fuel--Fifty-three interim storage casks (ISC), each holding a core component container (CCC), will be used to store the Fast Flux Test Facility (FFTF) SNF currently in the 400 Area. (2) Neutron Radiography Facility (NRF) TRIGA'--One Rad-Vault' container will store two DOT-6M3 containers and six NRF TRIGA casks currently stored in the 400 Area. (3) Commercial Light Water Reactor Fuel--Six International Standards Organization (ISO) containers, each holding a NAC-I cask4 with an inner commercial light water reactor (LWR) canister, will be used for commercial LWR SNF from the 300 Area. An aboveground dry cask storage location is necessary for the spent fuel because the current storage facilities are being shut down and deactivated. The spent fuel is being transferred to interim storage because there is no permanent repository storage currently available.

  6. Final Report on In-Reactor Creep-Fatigue Deformation Behaviour of a

    E-Print Network [OSTI]

    , in-reactor creep- fatigue tests have been performed at strain amplitudes of 0.25 and 0 temperatures of 326K and 323K. For comparison purposes corresponding out-of-reactor creep-fatigue tests were.2 Test module and irradiation rig 6 2.3 In-reactor creep-fatigue tests 7 2.4 Out-of-reactor creep

  7. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul (Pittsburgh, PA)

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  8. Corrective Action Plan for Corrective Action Unit 271: Areas 25, 26, and 27 Septic Systems, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    R. B. Jackson

    2003-05-01

    The Areas 25, 26 and 27 Septic Systems are in the Federal Facility Agreement and Consent Order (FFACO) of 1996 as Corrective Action Unit (CAU) 271. This Corrective Action Plan (CAP) provides selected corrective action alternatives and proposes the closure methodology for CAU 271. CAU 271 is located on the Nevada Test Site (NTS) approximately 105 kilometers (65 miles) northwest of Las Vegas, Nevada, and consists of the following 15 Corrective Action Sites (CAS): CAS 25-04-1, Septic System; CAS 25-04-03, Septic System; CAS25-04-04, Septic System; CAS 25-04-08, Septic System; CAS 25-04-09, Septic System; CAS 25-04-10, Septic System; CAS 25-04-11, Septic System; CAS 26-03-01, Contaminated Water Reservoir; CAS 26-04-1, Septic System; CAS 26-04-02, Septic System; CAS 26-05-01, Radioactive Leachfield; CAS-26-05-03, Septic System; CAS 26-05-04, Septic System; CAS 26-05-05, Septic System; and CAS 27-05-02, Leachfield.

  9. ANNUAL REPORT FOR THE FINAL GROUNDWATER REMEDIATION, TEST AREA NORTH, OPERABLE UNIT 1-07B, FISCAL YEAR 2009

    SciTech Connect (OSTI)

    FORSYTHE, HOWARD S

    2010-04-14

    This Annual Report presents the data and evaluates the progress of the three-component remedy implemented for remediation of groundwater contamination at Test Area North, Operable Unit 1-07B, at the Idaho National Laboratory Site. Overall, each component is achieving progress toward the goal of total plume remediation. In situ bioremediation operations in the hot spot continue to operate as planned. Progress toward the remedy objectives is being made, as evidenced by continued reduction in the amount of accessible residual source and decreases in downgradient contaminant flux, with the exception of TAN-28. The injection strategy is maintaining effective anaerobic reductive dechlorination conditions, as evidenced by complete degradation of trichloroethene and ethene production in the biologically active wells. In the medial zone, the New Pump and Treat Facility operated in standby mode. Trichloroethene concentrations in the medial zone wells are significantly lower than the historically defined concentration range of 1,000 to 20,000 ?g/L. The trichloroethene concentrations in TAN-33, TAN-36, and TAN-44 continue to be below 200 ?g/L. Monitoring in the distal zone wells outside and downgradient of the plume boundary demonstrate that some plume expansion has occurred, but less than the amount allowed in the Record of Decision Amendment. Additional data need to be collected for wells in the monitored natural attenuation part of the plume to confirm that the monitored natural attenuation part of the remedy is proceeding as predicted in the modeling.

  10. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    SciTech Connect (OSTI)

    Shott, Gregory [NSTec

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  11. Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 114: Area 25 EMAD Facility Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    Mark Krauss

    2010-06-01

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the actions needed to achieve closure for Corrective Action Unit (CAU) 114, Area 25 EMAD Facility, identified in the Federal Facility Agreement and Consent Order (FFACO). Corrective Action Unit 114 comprises the following corrective action site (CAS) located in Area 25 of the Nevada Test Site: • 25-41-03, EMAD Facility This plan provides the methodology for field activities needed to gather the necessary information for closing CAS 25-41-03. There is sufficient information and process knowledge from historical documentation and investigations of similar sites regarding the expected nature and extent of potential contaminants to recommend closure of CAU 114 using the SAFER process. Additional information will be obtained by conducting a field investigation before selecting the appropriate corrective action for CAS 25-41-03. It is anticipated that the results of the field investigation and implementation of corrective actions will support a defensible recommendation that no further corrective action is necessary. If it is determined that complete clean closure cannot be accomplished during the SAFER, then a hold point will have been reached and the Nevada Division of Environmental Protection (NDEP) will be consulted to determine whether the remaining contamination will be closed under the alternative corrective action of closure in place. This will be presented in a closure report that will be prepared and submitted to NDEP for review and approval. The CAS will be investigated based on the data quality objectives (DQOs) developed on April 30, 2009, by representatives of NDEP and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office. The DQO process was used to identify and define the type, amount, and quality of data needed to determine and implement appropriate corrective actions for CAS 25-41-03. The following text summarizes the SAFER activities that will support the closure of CAU 114: • Perform site preparation activities (e.g., utilities clearances, radiological surveys). • Collect samples of materials to determine whether potential source material (PSM) is present that may cause the future release of a contaminant of concern to environmental media. • If no PSMs are present at the CAS, establish no further action as the corrective action. • If a PSM is present at the CAS, either: - Establish clean closure as the corrective action. The material to be remediated will be removed and disposed of as waste, or - Establish closure in place as the corrective action and implement the appropriate use restrictions. • Confirm the selected closure option is sufficient to protect human health and the environment.

  12. Underground Test Area Fiscal Year 2014 Annual Quality Assurance Report Nevada National Security Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Krenzien, Susan

    2015-01-01

    This report is required by the Underground Test Area (UGTA) Quality Assurance Plan (QAP) and identifies the UGTA quality assurance (QA) activities from October 1, 2013, through September 30, 2014 (fiscal year [FY] 2014). All UGTA organizations—U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO); Desert Research Institute (DRI); Lawrence Livermore National Laboratory (LLNL); Los Alamos National Laboratory (LANL); National Security Technologies, LLC (NSTec); Navarro-Intera, LLC (N-I); and the U.S. Geological Survey (USGS)—conducted QA activities in FY 2014. The activities included conducting oversight assessments for QAP compliance, identifying findings and completing corrective actions, evaluating laboratory performance, and publishing documents. UGTA Activity participants conducted 25 assessments on topics including safe operations, QAP compliance, activity planning, and sampling. These assessments are summarized in Section 2.0. Corrective actions tracked in FY 2014 are presented in Appendix A. Laboratory performance was evaluated based on three approaches: (1) established performance evaluation programs (PEPs), (2) interlaboratory comparisons, or (3) data review. The results of the laboratory performance evaluations, and interlaboratory comparison results are summarized in Section 4.0. The UGTA Activity published three public documents and a variety of other publications in FY 2014. The titles, dates, and main authors are identified in Section 5.0. The Contract Managers, Corrective Action Unit (CAU) Leads, Preemptive Review (PER) Committee members, and Topical Committee members are listed by name and organization in Section 6.0. Other activities that affected UGTA quality are discussed in Section 7.0. Section 8.0 provides the FY 2014 UGTA QA program conclusions, and Section 9.0 lists the references not identified in Section 5.0.

  13. Alternate-fuel reactor studies

    SciTech Connect (OSTI)

    Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

    1983-02-01

    A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

  14. Type B Accident Investigation of the Serious Personal Injury while Doble Testing at the Western Area Power Administration Hayden Substation, May 19, 1999

    Broader source: Energy.gov [DOE]

    On May 19, 1999, at 10:31 a.m., four Western Area Power Administration (Western) employees were performing Doble testing on a circuit breaker at Hayden Substation in Routt County, Colorado. Three electricians were injured when the high-voltage lead (HVL) of the Doble test set encroached on the minimum approach distance to an energized part outside clearance boundaries, drawing arcing faults.

  15. Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 408: Bomblet Target Area Tonopah Test Range (TTR), Nevada, Revision 1

    SciTech Connect (OSTI)

    Mark Krauss

    2010-03-01

    This Streamlined Approach for Environmental Restoration Plan addresses the actions needed to achieve closure of Corrective Action Unit (CAU) 408, Bomblet Target Area (TTR). Corrective Action Unit 408 is located at the Tonopah Test Range and is currently listed in Appendix III of the Federal Facility Agreement and Consent Order. Corrective Action Unit 408 comprises Corrective Action Site TA-55-002-TAB2, Bomblet Target Areas. Clean closure of CAU 408 will be accomplished by removal of munitions and explosives of concern within seven target areas and potential disposal pits. The target areas were used to perform submunitions related tests for the U.S. Department of Energy (DOE). The scope of CAU 408 is limited to submunitions released from DOE activities. However, it is recognized that the presence of other types of unexploded ordnance and munitions may be present within the target areas due to the activities of other government organizations. The CAU 408 closure activities consist of: • Clearing bomblet target areas within the study area. • Identifying and remediating disposal pits. • Collecting verification samples. • Performing radiological screening of soil. • Removing soil containing contaminants at concentrations above the action levels. Based on existing information, contaminants of potential concern at CAU 408 include unexploded submunitions, explosives, Resource Conservation Recovery Act metals, and depleted uranium. Contaminants are not expected to be present in the soil at concentrations above the action levels; however, this will be determined by radiological surveys and verification sample results.

  16. Phoenix: A Reactor Burnup Code With Uncertainty Quantification 

    E-Print Network [OSTI]

    Spence, Grant R

    2014-12-15

    , and will continue to be, an area of great interest in nuclear research. Several methods have been developed to simulate reactor burnup; however, quantifying the uncertainty in reactor burnup simulations is in its relative infancy. This research developed a...

  17. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

    2009-03-10

    A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

  18. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

    1995-01-01

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  19. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  20. HEDL FACILITIES CATALOG 400 AREA

    SciTech Connect (OSTI)

    MAYANCSIK BA

    1987-03-01

    The purpose of this project is to provide a sodium-cooled fast flux test reactor designed specifically for irradiation testing of fuels and materials and for long-term testing and evaluation of plant components and systems for the Liquid Metal Reactor (LMR) Program. The FFTF includes the reactor, heat removal equipment and structures, containment, core component handling and examination, instrumentation and control, and utilities and other essential services. The complex array of buildings and equipment are arranged around the Reactor Containment Building.

  1. Corrective Action Decision Document/Closure Report for Corrective Action Unit 105: Area 2 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Matthews, Patrick

    2013-09-01

    This Corrective Action Decision Document/Closure Report presents information supporting the closure of Corrective Action Unit (CAU) 105: Area 2 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada. CAU 105 comprises the following five corrective action sites (CASs): -02-23-04 Atmospheric Test Site - Whitney Closure In Place -02-23-05 Atmospheric Test Site T-2A Closure In Place -02-23-06 Atmospheric Test Site T-2B Clean Closure -02-23-08 Atmospheric Test Site T-2 Closure In Place -02-23-09 Atmospheric Test Site - Turk Closure In Place The purpose of this Corrective Action Decision Document/Closure Report is to provide justification and documentation supporting the recommendation that no further corrective action is needed for CAU 105 based on the implementation of the corrective actions. Corrective action investigation (CAI) activities were performed from October 22, 2012, through May 23, 2013, as set forth in the Corrective Action Investigation Plan for Corrective Action Unit 105: Area 2 Yucca Flat Atmospheric Test Sites; and in accordance with the Soils Activity Quality Assurance Plan, which establishes requirements, technical planning, and general quality practices.

  2. Field studies of the potential for wind transport of plutonium- contaminated soils at sites in Areas 6 and 11, Nevada Test Site

    SciTech Connect (OSTI)

    Lancaster, N.; Bamford, R.; Metzger, S.

    1995-07-01

    This report describes and documents a series of field experiments carried out in Areas 6 and 11 of the Nevada Test Site in June and July 1994 to determine parameters of boundary layer winds, surface characteristics, and vegetation cover that can be used to predict dust emissions from the affected sites. Aerodynamic roughness of natural sites is determined largely by the lateral cover of the larger and more permanent roughness elements (shrubs). These provide a complete protection of the surface from wind erosion. Studies using a field-portable wind tunnel demonstrated that natural surfaces in the investigated areas of the Nevada Test Site are stable except at very high wind speeds (probably higher than normally occur, except perhaps in dust devils). However, disturbance of silty-clay surfaces by excavation devices and vehicles reduces the entrainment threshold by approximately 50% and makes these areas potentially very susceptible to wind erosion and transport of sediments.

  3. Radiological survey and evaluation of the fallout area from the Trinity test: Chupadera Mesa and White Sands Missile Range, New Mexico

    SciTech Connect (OSTI)

    Hansen, W.R.; Rodgers, J.C.

    1985-06-01

    Current radiological conditions were evaluated for the site of the first nuclear weapons test, the Trinity test, and the associated fallout zone. The test, located on White Sands Missile Range, was conducted as part of the research with nuclear materials for the World War II Manhattan Engineer District atomic bomb project. Some residual radioactivity attributable to the test was found in the soils of Ground Zero on White Sands Missile Range and the areas that received fallout from the test. The study considered relevant information including historical records, environmental data extending back to the 1940s, and new data acquired by field sampling and measurements. Potential exposures to radiation were evaluated for current land uses. Maximum estimated doses on Chupadera Mesa and other uncontrolled areas are less than 3% of the DOE Radiation Protection Standards (RPSs). Radiation exposures during visits to the US Army-controlled Ground Zero area are less than 1 mrem per annual visit or less than 0.2% of the RPS for a member of the public. Detailed data and interpretations are provided in appendixes. 14 figs., 45 tabs.

  4. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  5. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    SciTech Connect (OSTI)

    Corradin, Michael; Anderson, M.; Muci, M.; Hassan, Yassin; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  6. Lagoon Seepage Testing Procedures for Central Facilities Area (CFA) Sewage Lagoons at Idaho National Laboratory Butte County, Idaho April 2014

    SciTech Connect (OSTI)

    Alan Giesbrecht

    2014-05-01

    The lagoon seepage testing procedures are documented herein as required by the Wastewater Rules (IDAPA 58.01.16.493). The Wastewater Rules and Wastewater Reuse Permit LA-000141-03 require that the procedure used for performing a seepage test be approved by IDEQ prior to conducting the seepage test. The procedures described herein are based on a seepage testing plan that was developed by J-U-B ENGINEERS, Inc. (J-U-B) and has been accepted by several IDEQ offices for lagoons in Idaho.

  7. SAVANNAH RIVER SITE R REACTOR DISASSEMBLY BASIN IN SITU DECOMMISSIONING

    SciTech Connect (OSTI)

    Langton, C.; Blankenship, J.; Griffin, W.; Serrato, M.

    2009-12-03

    The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate if from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,424 cubic meters or 31,945 cubic yards. Portland cement-based structural fill materials were design and tested for the reactor ISD project and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and work flow considerations, the recommended maximum lift height is 5 feet with 24 hours between lifts. Pertinent data and information related to the SRS 105-R-Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material designs and testing, and fill placement strategy. This information is applicable to decommissioning both the 105-P and 105-R facilities. The ISD process for the entire 105-P and 105-R reactor facilities will require approximately 250,000 cubic yards (191,140 cubic meters) of grout and 2,400 cubic yards (1,840 cubic meters) of structural concrete which will be placed over a twelve month period to meet the accelerated schedule ISD schedule. The status and lessons learned in the SRS Reactor Facility ISD process will be described.

  8. Review of light water reactor safety

    SciTech Connect (OSTI)

    Cheng, H.S.

    1980-12-01

    A review of the present status of light water reactor (LWR) safety is presented. The review starts with a brief discussion of the outstanding accident scenarios concerning LWRs. Where possible the areas of present technological uncertainties are stressed. To provide a better perspective of reactor safety, it then reviews the probabilistic assessment of the outstanding LWR accidents considered in the Reactor Safety Study (WASH-1400) and discusses the potential impact of the present technological uncertainties on WASH-1400.

  9. Closure Report for Corrective Action Unit 127: Areas 25 and 26 Storage Tanks, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2008-02-01

    CAU 127, Areas 25 and 26 Storage Tanks, consists of twelve CASs located in Areas 25 and 26 of the NTS. The closure alternatives included no further action, clean closure, and closure in place with administrative controls. The purpose of this Closure Report is to provide a summary of the completed closure activities, documentation of waste disposal, and analytical data to confirm that the remediation goals were met.

  10. New Remote Method for Estimation of Contamination Levels of Reactor Equipment - 13175

    SciTech Connect (OSTI)

    Danilovich, Alexey; Ivanov, Oleg; Potapov, Victor; Semenov, Sergey; Semin, Ilya; Smirnov, Sergey; Stepanov, Vyacheslav; Volkovich, Anatoly

    2013-07-01

    Projects for decommissioning of shutdown reactors and reactor facilities carried out in several countries, including Russia. In the National Research Centre 'Kurchatov Institute' decontamination and decommissioning of the research reactor MR (Material Testing Reactor) has been initiated. The research reactor MR has a long history and consists of nine loop facilities for experiments with different kinds of fuel. During the operation of main and auxiliary equipment of reactors it was subjected to strong radioactive contamination. The character of this contamination requires individual strategies for the decontamination work. This requires information about the character of the distribution of radioactive contamination of equipment in the premises. A detailed radiation survey of these premises using standard dosimetric equipment is almost impossible because of high levels of radiation and high-density of the equipment that does not allow identifying the most active fragments using standard tools of measurement. The problem can be solved using the method of remote measurements of distribution of radioactivity with help of the collimated gamma-ray detectors. For radiation surveys of the premises of loop installations remotely operated spectrometric collimated system was used [1, 2, 3]. As a result of the work, maps of the distribution of activity and dose rate for surveyed premises were plotted and superimposed on its photo. The new results of measurements in different areas of the reactor and at its loop installations, with emphasis on the radioactive survey of highly-contaminated samples, are presented. (authors)

  11. Corrective Action Decision Document/Corrective Action Plan for the 92-Acre Area and Corrective Action Unit 111: Area 5 WMD Retired Mixed Waste Pits, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2009-07-31

    This Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) has been prepared for the 92-Acre Area, the southeast quadrant of the Radioactive Waste Management Site, located in Area 5 of the Nevada Test Site (NTS). The 92-Acre Area includes Corrective Action Unit (CAU) 111, 'Area 5 WMD Retired Mixed Waste Pits.' Data Quality Objectives (DQOs) were developed for the 92-Acre Area, which includes CAU 111. The result of the DQO process was that the 92-Acre Area is sufficiently characterized to provide the input data necessary to evaluate corrective action alternatives (CAAs) without the collection of additional data. The DQOs are included as Appendix A of this document. This CADD/CAP identifies and provides the rationale for the recommended CAA for the 92-Acre Area, provides the plan for implementing the CAA, and details the post-closure plan. When approved, this CADD/CAP will supersede the existing Pit 3 (P03) Closure Plan, which was developed in accordance with Title 40 Code of Federal Regulations (CFR) Part 265, 'Interim Status Standards for Owners and Operators of Hazardous Waste Treatment, Storage, and Disposal Facilities.' This document will also serve as the Closure Plan and the Post-Closure Plan, which are required by 40 CFR 265, for the 92-Acre Area. After closure activities are complete, a request for the modification of the Resource Conservation and Recovery Act Permit that governs waste management activities at the NTS will be submitted to the Nevada Division of Environmental Protection to incorporate the requirements for post-closure monitoring. Four CAAs, ranging from No Further Action to Clean Closure, were evaluated for the 92-Acre Area. The CAAs were evaluated on technical merit focusing on performance, reliability, feasibility, safety, and cost. Based on the evaluation of the data used to develop the conceptual site model; a review of past, current, and future operations at the site; and the detailed and comparative analysis of the potential CAAs, Closure in Place with Administrative Controls is the preferred CAA for the 92-Acre Area. Closure activities will include the following: (1) Constructing an engineered evapotranspiration cover over the 92-Acre Area; (2) Installing use restriction (UR) warning signs, concrete monuments, and subsidence survey monuments; (3) Establishing vegetation on the cover; (4) Implementing a UR; and (5) Implementing post-closure inspections and monitoring. The Closure in Place with Administrative Controls alternative meets all requirements for the technical components evaluated, fulfills all applicable federal and state regulations for closure of the site, and will minimize potential future exposure pathways to the buried waste at the site.

  12. Closed Brayton cycle power conversion systems for nuclear reactors :

    SciTech Connect (OSTI)

    Wright, Steven A.; Lipinski, Ronald J.; Vernon, Milton E.; Sanchez, Travis

    2006-04-01

    This report describes the results of a Sandia National Laboratories internally funded research program to study the coupling of nuclear reactors to gas dynamic Brayton power conversion systems. The research focused on developing integrated dynamic system models, fabricating a 10-30 kWe closed loop Brayton cycle, and validating these models by operating the Brayton test-loop. The work tasks were performed in three major areas. First, the system equations and dynamic models for reactors and Closed Brayton Cycle (CBC) systems were developed and implemented in SIMULINKTM. Within this effort, both steady state and dynamic system models for all the components (turbines, compressors, reactors, ducting, alternators, heat exchangers, and space based radiators) were developed and assembled into complete systems for gas cooled reactors, liquid metal reactors, and electrically heated simulators. Various control modules that use proportional-integral-differential (PID) feedback loops for the reactor and the power-conversion shaft speed were also developed and implemented. The simulation code is called RPCSIM (Reactor Power and Control Simulator). In the second task an open cycle commercially available Capstone C30 micro-turbine power generator was modified to provide a small inexpensive closed Brayton cycle test loop called the Sandia Brayton test-Loop (SBL-30). The Capstone gas-turbine unit housing was modified to permit the attachment of an electrical heater and a water cooled chiller to form a closed loop. The Capstone turbine, compressor, and alternator were used without modification. The Capstone systems nominal operating point is 1150 K turbine inlet temperature at 96,000 rpm. The annular recuperator and portions of the Capstone control system (inverter) and starter system also were reused. The rotational speed of the turbo-machinery is controlled by adjusting the alternator load by using the electrical grid as the load bank. The SBL-30 test loop was operated at the manufacturers site (Barber-Nichols Inc.) and installed and operated at Sandia. A sufficiently detailed description of the loop is provided in this report along with the design characteristics of the turbo-alternator-compressor set to allow other researchers to compare their results with those measured in the Sandia test-loop. The third task consisted of a validation effort. In this task the test loop was operated and compared with the modeled results to develop a more complete understanding of this electrically heated closed power generation system and to validate the model. The measured and predicted system temperatures and pressures are in good agreement, indicating that the model is a reasonable representation of the test loop. Typical deviations between the model and the hardware results are less than 10%. Additional tests were performed to assess the capability of the Brayton engine to continue to remove decay heat after the reactor/heater is shutdown, to develop safe and effective control strategies, and to access the effectiveness of gas inventory control as an alternative means to provide load following. In one test the heater power was turned off to simulate a rapid reactor shutdown, and the turbomachinery was driven solely by the sensible heat stored in the heater for over 71 minutes without external power input. This is an important safety feature for CBC systems as it means that the closed Brayton loop will keep cooling the reactor without the need for auxiliary power (other than that needed to circulate the waste heat rejection coolant) provided the heat sink is available.

  13. Research and Medical Isotope Reactor Supply | Y-12 National Security...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Y-12 tops the short list of the world's most secure, reliable uranium feedstock suppliers for dozens of research and test reactors on six continents. These reactors can be...

  14. Design of Complex Systems to Achieve Passive Safety: Natural Circulation Cooling of Liquid Salt Pebble Bed Reactors

    E-Print Network [OSTI]

    Scarlat, Raluca Olga

    2012-01-01

    safety design criteria separate effects test steam generators small modular reactor San Onofre Nuclear

  15. Type A Accident Investigation Board Report on the August 13, 1996, Electrical Shock at TRA-609, Test Reactor Area, Idaho National Engineering Laboratory

    Broader source: Energy.gov [DOE]

    This report is an independent product of an electrical shock accident investigation report board appointed by John M. Wilcynski, Manager, Idaho Operations Office, U.S. Department of Energy.

  16. Lagoon Seepage Testing Report for Central Facilities Area (CFA) Sewage Lagoons at Idaho National Laboratory, Butte County, Idaho

    SciTech Connect (OSTI)

    Bridger Morrison

    2014-09-01

    J-U-B ENGINEERS, Inc. (J-U-B) performed seepage tests on the CFA Wastewater Lagoons 1, 2, and 3 between August 26th and September 22nd, 2014. The lagoons were tested to satisfy the Idaho Department of Environmental Quality (DEQ) Rules (IDAPA 58.01.16) that require all lagoons be tested at a frequency of every 10 years and the Compliance Activity CA-141-03 in the DEQ Wastewater Reuse Permit for the CFA Sewage Treatment Plant (LA-000141-03). The lagoons were tested to determine if the average seepage rates are less than 0.25 in/day, the maximum seepage rate allowed for lagoons built prior to April 15, 2007. The average seepage rates were estimated for each lagoon and are given in Table-ES1. The average seepage rates for Lagoons 1 and 2 are less than the allowable seepage rate of 0.25 in/day. Lagoon 1 and 2 passed the seepage test and will not have to be tested again until the year 20241. However, the average seepage rate for Lagoon 3 appears to exceed the allowable seepage rate of 0.25 in/day which means the potential source for the excessive leakage should be investigated further.

  17. Environmental assessment of SP-100 ground engineering system test site: Hanford Site, Richland, Washington

    SciTech Connect (OSTI)

    Not Available

    1988-12-01

    The US Department of Energy (DOE) proposes to modify an existing reactor containment building (decommissioned Plutonium Recycle Test Reactor (PRTR) 309 Building) to provide ground test capability for the prototype SP-100 reactor. The 309 Building (Figure 1.1) is located in the 300 Area on the Hanford Site in Washington State. The National Environmental Policy Act (NEPA) requires that Federal agencies assess the potential impacts that their actions may have on the environment. This Environmental Assessment describes the consideration given to environmental impacts during reactor concept and test site selection, examines the environmental effects of the DOE proposal to ground test the nuclear subsystem, describes alternatives to the proposed action, and examines radiological risks of potential SP-100 use in space. 73 refs., 19 figs., 7 tabs.

  18. Experimental Breeder Reactor I Preservation Plan

    SciTech Connect (OSTI)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  19. Laboratory and Field Studies Related to Radionuclide Migration at the Nevada Test Site in Support of the Underground Test Area and Hydrologic Resources Management Projects

    SciTech Connect (OSTI)

    D.L.Finnegan; J.L.Thompson

    2002-06-01

    This report details the work of Chemistry Division personnel from Los Alamos National Laboratory in FY 2001 for the U. S. Department of Energy National Nuclear Security Administration Nevada Operations Office (NNSA/NV) under its Defense Programs and Environmental Restoration divisions. Los Alamos is one of a number of agencies collaborating in an effort to describe the present and future movement of radionuclides in the underground environment of the Nevada Test Site. This fiscal year we collected and analyzed water samples from a number of expended test locations at the Nevada Test Site. We give the results of these analyses and summarize the information gained over the quarter century that we have been studying several of these sites. We find that by far most of the radioactive residues from a nuclear test are contained in the melt glass in the cavity. Those radionuclides that are mobile in water can be transported if the groundwater is moving due to hydraulic or thermal gradients. The extent to which they move is a function of their chemical speciation, with neutral or anionic materials traveling freely relative to cationic materials that tend to sorb on rock surfaces. However, radionuclides sorbed on colloids may be transported if the colloids are moving. Local conditions strongly influence the distribution and movement of radionuclides, and we continue to study sites such as Almendro, which is thermally quite hot, and Nash and Bourbon, where radionuclides had not been measured for 8 years. We collected samples from three characterization wells in Frenchman Flat to obtain baseline radiochemistry data for each well, and we analyzed eight wells containing radioactivity for {sup 237}Np, using our highly sensitive ICP/MS. We have again used our field probe that allows us to measure important groundwater properties in situ. We conclude our report by noting document reviews and publications produced in support of this program.

  20. Assessment of Geothermal Resource Potential at a High-Priority Area on the Utah Testing and Training Range–South (UTTR–S)

    SciTech Connect (OSTI)

    Richard P. Smith, PhD., PG; Robert P. Breckenridge, PhD.; Thomas R. Wood, PhD.

    2012-04-01

    Field investigations conducted during 2011 support and expand the conclusion of the original Preliminary Report that discovery of a viable geothermal system is possible in the northwestern part of the Utah Testing and Training Range-South (UTTR-S), referred to henceforth as Focus Area 1. The investigations defined the southward extent of the Wendover graben into and near Focus Area 1, enhanced the understanding of subsurface conditions, and focused further geothermal exploration efforts towards the northwestern-most part of Focus Area 1. Specifically, the detailed gravity survey shows that the Wendover graben, first defined by Cook et al. (1964) for areas north of Interstate Highway 80, extends and deepens southwest-ward to the northwest corner of Focus Area 1. At its deepest point, the intersection with a northwest-trending graben there is favorable for enhanced permeability associated with intersecting faults. Processing and modeling of the gravity data collected during 2011 provide a good understanding of graben depth and distribution of faults bounding the graben and has focused the interest area of the study. Down-hole logging of temperatures in wells made available near the Intrepid, Inc., evaporation ponds, just north of Focus Area 1, provide a good understanding of the variability of thermal gradients in that area and corroborate the more extensive temperature data reported by Turk (1973) for the depth range of 300-500 m. Moderate temperature gradients in the northern part of the Intrepid area increase to much higher gradients and bottom-hole temperatures southeastward, towards graben-bounding faults, suggesting upwelling geothermal waters along those faults. Water sampling, analysis, and temperature measurements of Blue Lakes and Mosquito Willey's springs, on the western boundary of Focus Area 1, also show elevated temperatures along the graben-bounding fault system. In addition, water chemistry suggests origin of those waters in limestone rocks beneath the graben in areas with temperatures as high as 140 C (284 F). In conclusion, all of the field data collected during 2011 and documented in the Appendices of this report indicate that there is reasonable potential for a viable geothermal resource along faults that bound the Wendover graben. Prospects for a system capable of binary electrical generation are especially good, and the possibility of a flash steam system is also within reason. The next steps should focus on securing the necessary funding for detailed geophysical surveys and for drilling a set of temperature gradient wells to further evaluate the resource, and to focus deep exploration efforts in the most promising areas.

  1. Modeling and Simulation for Nuclear Reactors Hub | Department...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    to help the nuclear industry make reactors more efficient through computer modeling and simulation. The Department's Energy Innovation Hubs are helping to advance promising areas...

  2. 2009 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site, Nye County, Nevada: Review of the Performance Assessments and Composite Analysis

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2010-03-15

    The U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office performed an annual review of the Area 3 and Area 5 Radioactive Wate Management Site (RWMS) Performance Assessments (PAs) and Composite Analyses (CAs) in fiscal year (FY) 2009. This annual summary report presents data and conclusions from the FY 2009 review, and determines the adequacy of the PAs and CAs. Operational factors (e.g., waste forms and containers, facility design, and waste receipts), closure plans, monitoring results, and research and development (R&D) activities were reviewed to determine the adequacy of the PAs. Likewise, the environmental restoration activities at the Nevada Test Site relevant to the sources of residual radioactive material that are considered in the CAs, the land-use planning, and the results of the environmental monitoring and R&D activities were reviewed to determine the adequacy of the CAs.

  3. Corrective Action Decision Document/Closure Report for Corrective Action Unit 383: Area E-Tunnel Sites, Nevada Test Site

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2010-03-15

    This Corrective Action Decision Document/Closure Report (CADD/CR) was prepared by the Defense Threat Reduction Agency (DTRA) for Corrective Action Unit (CAU) 383, Area 12 E-Tunnel Sites, which is the joint responsibility of DTRA and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (NNSA/NSO). This CADD/CR is consistent with the requirements of the Federal Facility Agreement and Consent Order (FFACO) agreed to by the State of Nevada, the DOE, and the U.S. Department of Defense. Corrective Action Unit 383 is comprised of three Corrective Action Sites (CASs) and two adjacent areas: • CAS 12-06-06, Muckpile • CAS 12-25-02, Oil Spill • CAS 12-28-02, Radioactive Material • Drainage below the Muckpile • Ponds 1, 2, and 3 The purpose of this CADD/CR is to provide justification and documentation to support the recommendation for closure with no further corrective action, by placing use restrictions at the three CASs and two adjacent areas of CAU 383.

  4. Structure of processes in flow reactor and closed reactor: Flow reactor

    E-Print Network [OSTI]

    Greifswald, Ernst-Moritz-Arndt-Universität

    Structure of processes in flow reactor and closed reactor: Flow reactor Closed reactor Active Zone -- chemical quasi- equilibria, similarity principles and macroscopic kinetics", in: Lectures on Plasma Physics

  5. The IAEA Coordinated Research Program on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis: Description of the Benchmark Test Cases and Phases

    SciTech Connect (OSTI)

    Frederik Reitsma; Gerhard Strydom; Bismark Tyobeka; Kostadin Ivanov

    2012-10-01

    The continued development of High Temperature Gas Cooled Reactors (HTGRs) requires verification of design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The uncertainties in the HTR analysis tools are today typically assessed with sensitivity analysis and then a few important input uncertainties (typically based on a PIRT process) are varied in the analysis to find a spread in the parameter of importance. However, one wish to apply a more fundamental approach to determine the predictive capability and accuracies of coupled neutronics/thermal-hydraulics and depletion simulations used for reactor design and safety assessment. Today there is a broader acceptance of the use of uncertainty analysis even in safety studies and it has been accepted by regulators in some cases to replace the traditional conservative analysis. Finally, there is also a renewed focus in supplying reliable covariance data (nuclear data uncertainties) that can then be used in uncertainty methods. Uncertainty and sensitivity studies are therefore becoming an essential component of any significant effort in data and simulation improvement. In order to address uncertainty in analysis and methods in the HTGR community the IAEA launched a Coordinated Research Project (CRP) on the HTGR Uncertainty Analysis in Modelling early in 2012. The project is built on the experience of the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity, but focuses specifically on the peculiarities of HTGR designs and its simulation requirements. Two benchmark problems were defined with the prismatic type design represented by the MHTGR-350 design from General Atomics (GA) while a 250 MW modular pebble bed design, similar to the INET (China) and indirect-cycle PBMR (South Africa) designs are also included. In the paper more detail on the benchmark cases, the different specific phases and tasks and the latest status and plans are presented.

  6. Predevelopment Water-Level Contours for Aquifers in the Rainier Mesa and Shoshone Mountain area of the Nevada Test Site, Nye County, Nevada

    SciTech Connect (OSTI)

    Joseph M. Fenelon; Randell J. Laczniak; and Keith J. Halford

    2008-06-24

    Contaminants introduced into the subsurface of the Nevada Test Site at Rainier Mesa and Shoshone Mountain by underground nuclear testing are of concern to the U.S. Department of Energy and regulators responsible for protecting human health and safety. Although contaminants were introduced into low-permeability rocks above the regional flow system, the potential for contaminant movement away from the underground test areas and into the accessible environment is greatest by ground-water transport. The primary hydrologic control on this transport is evaluated and examined through a series of contour maps developed to represent the water-level distribution within each of the major aquifers underlying the area. Aquifers were identified and their extents delineated by merging and analyzing multiple hydrostratigraphic framework models developed by other investigators from existing geologic information. The contoured water-level distribution in each major aquifer was developed from a detailed evaluation and assessment of available water-level measurements. Multiple spreadsheets that accompany this report provide pertinent water-level and geologic data by well or drill hole. Aquifers are mapped, presented, and discussed in general terms as being one of three aquifer types—volcanic aquifer, upper carbonate aquifer, or lower carbonate aquifer. Each of these aquifer types was subdivided and mapped as independent continuous and isolated aquifers, based on the continuity of its component rock. Ground-water flow directions, as related to the transport of test-generated contaminants, were developed from water-level contours and are presented and discussed for each of the continuous aquifers. Contoured water-level altitudes vary across the study area and range from more than 5,000 feet in the volcanic aquifer beneath a recharge area in the northern part of the study area to less than 2,450 feet in the lower carbonate aquifer in the southern part of the study area. Variations in water-level altitudes within any single continuous aquifer range from a few hundred feet in a lower carbonate aquifer to just more than 1,100 feet in a volcanic aquifer. Flow directions throughout the study area are dominantly southward with minor eastward or westward deviations. Primary exceptions are westward flow in the northern part of the volcanic aquifer and eastward flow in the eastern part of the lower carbonate aquifer. Northward flow in the upper and lower carbonate aquifers in the northern part of the study area is possible but cannot be substantiated because data are lacking. Interflow between continuous aquifers is evaluated and mapped to define major flow paths. These flow paths delineate tributary flow systems, which converge to form the regional ground-water flow system. The implications of these tributary flow paths in controlling transport away from the underground test areas at Rainier Mesa and Shoshone Mountain are discussed. The obvious data gaps contributing to uncertainties in the delineation of aquifers and development of water-level contours are identified and evaluated.

  7. Closure Report for Corrective Action Unit 326: Areas 6 and 27 Release Sites, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    K. B. Campbell

    2002-12-01

    This Closure Report (CR) documents the activities undertaken to close Corrective Action Unit (CAU) 326, Areas 6 and 27 Release Sites, in accordance with the Federal Facility Agreement and Consent Order (FFACO) of 1996. Site closure was performed in accordance with the Nevada Division of Environmental Protection (NDEP)-approved Streamlined Approach for Environmental Restoration Plan (SAFER) Plan for CAU 326 (US Department of Energy, Nevada Operations Office [DOE/NV, 2001]). CAU 326 consists of four Corrective Action Sites (CASs), 06-25-01, 06-25-02, 06-25-04, and 27-25-01. CAS 06-25-01 is a release site associated with an underground pipeline that carried heating oil from the heating oil underground storage tank (UST), Tank 6-CP-1, located to the west of Building CP-70 to the boiler in Building CP-1 located in the Area 6 Control Point (CP) compound. This site was closed in place administratively by implementing use restrictions. CAS 06-25-02 is a hydrocarbon release associated with an active heating oil UST, Tank 6-DAF-5, located west of Building 500 at the Area 6 Device Assembly Facility. This site was closed in place administratively by implementing use restrictions. CAS 06-25-04 was a hydrocarbon release associated with Tank 6-619-4. This site was successfully remediated when Tank 6-619-4 was removed. No further action was taken at this site. CAS 27-25-01 is an excavation that was created in an attempt to remove hydrocarbon-impacted soil from the Site Maintenance Yard in Area 27. Approximately 53 cubic meters (m{sup 3}) (70 cubic yards [yd{sup 3}]) of soil impacted by total petroleum hydrocarbons (TPH) and polychlorinated biphenyls (PCBs) was excavated from the site in August of 1994. Clean closure of this site was completed in 2002 by the excavation and disposal of approximately 160 m{sup 3} (210 yd{sup 3}) of PCB-impacted soil.

  8. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  9. Corrective Action Plan for Corrective Action Unit 321: Area 22 Weather Station Fuel Storage Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    D. S. Tobiason

    2000-06-01

    The purpose of this Corrective Action Plan (CAP) is to provide the strategy and methodology to close the Area 22 Weather Station Fuel Storage. The CAU will be closed following state and federal regulations and the FFACO (1996). Site characterization was done during February 1999. Soil samples were collected using a direct-push method. Soil samples were collected at 0.6-m (2-ft) intervals from the surface to 1.8 m (6 ft) below ground surface. The results of the characterization were reported in the Corrective Action Decision Document (CADD) (DOE, 1999b). Soil sample results indicated that two locations in the bermed area contain total petroleum hydrocarbons (TPH) as diesel at concentrations of 124 milligrams per kilogram (mg/kg) and 377 mg/kg. This exceeds the Nevada Division of Environmental Protection (NDEP) regulatory action level for TPH of 100 mg/kg (Nevada Administrative Code, 1996). The TPH-impacted soil will be removed and disposed as part of the corrective action.

  10. Corrective Action Investigation Plan for Corrective Action Unit 104: Area 7 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Patrick Matthews

    2011-08-01

    CAU 104 comprises the 15 CASs listed below: (1) 07-23-03, Atmospheric Test Site T-7C; (2) 07-23-04, Atmospheric Test Site T7-1; (3) 07-23-05, Atmospheric Test Site; (4) 07-23-06, Atmospheric Test Site T7-5a; (5) 07-23-07, Atmospheric Test Site - Dog (T-S); (6) 07-23-08, Atmospheric Test Site - Baker (T-S); (7) 07-23-09, Atmospheric Test Site - Charlie (T-S); (8) 07-23-10, Atmospheric Test Site - Dixie; (9) 07-23-11, Atmospheric Test Site - Dixie; (10) 07-23-12, Atmospheric Test Site - Charlie (Bus); (11) 07-23-13, Atmospheric Test Site - Baker (Buster); (12) 07-23-14, Atmospheric Test Site - Ruth; (13) 07-23-15, Atmospheric Test Site T7-4; (14) 07-23-16, Atmospheric Test Site B7-b; (15) 07-23-17, Atmospheric Test Site - Climax These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives (CAAs). Additional information will be obtained by conducting a corrective action investigation before evaluating CAAs and selecting the appropriate corrective action for each CAS. The results of the field investigation will support a defensible evaluation of viable CAAs that will be presented in the Corrective Action Decision Document. The sites will be investigated based on the data quality objectives (DQOs) developed on April 28, 2011, by representatives of the Nevada Division of Environmental Protection and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office. The DQO process was used to identify and define the type, amount, and quality of data needed to develop and evaluate appropriate corrective actions for CAU 104. The releases at CAU 104 consist of surface-deposited radionuclides from 30 atmospheric nuclear tests. The presence and nature of contamination at CAU 104 will be evaluated based on information collected from a field investigation. Radiological contamination will be evaluated based on a comparison of the total effective dose (TED) to the dose-based final action level (FAL). The presence of TED exceeding the FAL is considered a radiological contaminant of concern (COC). Anything identified as a COC will require corrective action. The TED will be calculated as the total of separate estimates of internal and external dose. Results from the analysis of soil samples will be used to calculate internal radiological dose. Thermoluminescent dosimeters will be used to measure external radiological dose. Based on process knowledge of the releases associated with the nuclear tests and radiological survey information about the location and shape of the resulting contamination plume, it was determined that the releases from the nuclear tests are co-located and will be investigated concurrently. A field investigation will be performed to define areas where TED exceeds the FAL and to determine whether other COCs are present at the site. The investigation will also collect information to determine the presence and nature of contamination associated with migration and excavation, as well as any potential releases discovered during the investigation. Appendix A provides a detailed discussion of the DQO methodology and the DQOs specific to each CAS.

  11. Horizontal baffle for nuclear reactors

    DOE Patents [OSTI]

    Rylatt, John A. (Monroeville, PA)

    1978-01-01

    A horizontal baffle disposed in the annulus defined between the core barrel and the thermal liner of a nuclear reactor thereby physically separating the outlet region of the core from the annular area below the horizontal baffle. The horizontal baffle prevents hot coolant that has passed through the reactor core from thermally damaging apparatus located in the annulus below the horizontal baffle by utilizing the thermally induced bowing of the horizontal baffle to enhance sealing while accommodating lateral motion of the baffle base plate.

  12. Characterization of Sediments from the Soil Desiccation Pilot Test (SDPT) Site in the BC Cribs and Trenches Area

    SciTech Connect (OSTI)

    Um, Wooyong; Truex, Michael J.; Valenta, Michelle M.; Iovin, Cristian; Kutnyakov, Igor V.; Chang, Hyun-shik; Clayton, Ray E.; Serne, R. Jeffrey; Ward, Anderson L.; Brown, Christopher F.; Geiszler, Keith N.; Clayton, Eric T.; Baum, Steven R.; Smith, David M.

    2009-09-25

    This technical report documents the results of laboratory geochemical and hydrologic measurements of sediments collected from new borehole 299-E13-65 (C7047) and comparison of the results with those of nearby borehole 299-13E-62 (C5923) both drilled in the BC Cribs and Trenches Area. The total and water-leachable concentrations of key contaminants will be used to update contaminant-distribution conceptual models and to provide more data for improving baseline risk predictions and remedial alternative selections. Improved understanding of subsurface conditions and methods to remediate these principal contaminants can be also used to evaluate the application of specific technologies to other contaminants across the Hanford Site.

  13. Generation IV Reactors Integrated Materials Technology Program Plan: Focus on Very High Temperature Reactor Materials

    SciTech Connect (OSTI)

    Corwin, William R; Burchell, Timothy D; Katoh, Yutai; McGreevy, Timothy E; Nanstad, Randy K; Ren, Weiju; Snead, Lance Lewis; Wilson, Dane F

    2008-08-01

    Since 2002, the Department of Energy's (DOE's) Generation IV Nuclear Energy Systems (Gen IV) Program has addressed the research and development (R&D) necessary to support next-generation nuclear energy systems. The six most promising systems identified for next-generation nuclear energy are described within this roadmap. Two employ a thermal neutron spectrum with coolants and temperatures that enable hydrogen or electricity production with high efficiency (the Supercritical Water Reactor-SCWR and the Very High Temperature Reactor-VHTR). Three employ a fast neutron spectrum to enable more effective management of actinides through recycling of most components in the discharged fuel (the Gas-cooled Fast Reactor-GFR, the Lead-cooled Fast Reactor-LFR, and the Sodium-cooled Fast Reactor-SFR). The Molten Salt Reactor (MSR) employs a circulating liquid fuel mixture that offers considerable flexibility for recycling actinides and may provide an alternative to accelerator-driven systems. At the inception of DOE's Gen IV program, it was decided to significantly pursue five of the six concepts identified in the Gen IV roadmap to determine which of them was most appropriate to meet the needs of future U.S. nuclear power generation. In particular, evaluation of the highly efficient thermal SCWR and VHTR reactors was initiated primarily for energy production, and evaluation of the three fast reactor concepts, SFR, LFR, and GFR, was begun to assess viability for both energy production and their potential contribution to closing the fuel cycle. Within the Gen IV Program itself, only the VHTR class of reactors was selected for continued development. Hence, this document will address the multiple activities under the Gen IV program that contribute to the development of the VHTR. A few major technologies have been recognized by DOE as necessary to enable the deployment of the next generation of advanced nuclear reactors, including the development and qualification of the structural materials needed to ensure their safe and reliable operation. The focus of this document will be the overall range of DOE's structural materials research activities being conducted to support VHTR development. By far, the largest portion of material's R&D supporting VHTR development is that being performed directly as part of the Next-Generation Nuclear Plant (NGNP) Project. Supplementary VHTR materials R&D being performed in the DOE program, including university and international research programs and that being performed under direct contracts with the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, will also be described. Specific areas of high-priority materials research that will be needed to deploy the NGNP and provide a basis for subsequent VHTRs are described, including the following: (1) Graphite: (a) Extensive unirradiated materials characterization and assessment of irradiation effects on properties must be performed to qualify new grades of graphite for nuclear service, including thermo-physical and mechanical properties and their changes, statistical variations from billot-to-billot and lot-to-lot, creep, and especially, irradiation creep. (b) Predictive models, as well as codification of the requirements and design methods for graphite core supports, must be developed to provide a basis for licensing. (2) Ceramics: Both fibrous and load-bearing ceramics must be qualified for environmental and radiation service as insulating materials. (3) Ceramic Composites: Carbon-carbon and SiC-SiC composites must be qualified for specialized usage in selected high-temperature components, such as core stabilizers, control rods, and insulating covers and ducting. This will require development of component-specific designs and fabrication processes, materials characterization, assessment of environmental and irradiation effects, and establishment of codes and standards for materials testing and design requirements. (4) Pressure Vessel Steels: (a) Qualification of short-term, high-temperature properties of light water rea

  14. Modular Pebble Bed Reactor March 22, 2000

    E-Print Network [OSTI]

    ;"Naturally" Safe Fuel · Shut Off All Cooling · Withdraw All Control Rods · No Emergency Cooling · No Operator · No melt down · No significant radiation release in accident · Demonstrate with actual test of reactor #12

  15. Corrective Action Decision Document/Closure Report for Corrective Action Unit 570: Area 9 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Matthews, Patrick

    2013-11-01

    This Corrective Action Decision Document/Closure Report presents information supporting the closure of Corrective Action Unit (CAU) 570: Area 9 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada. This complies with the requirements of the Federal Facility Agreement and Consent Order (FFACO) that was agreed to by the State of Nevada; U.S. Department of Energy (DOE), Environmental Management; U.S. Department of Defense; and DOE, Legacy Management. The purpose of the CADD/CR is to provide justification and documentation supporting the recommendation that no further corrective action is needed.

  16. Corrective Action Decision Document/Closure Report for Corrective Action Unit 105: Area 2 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada, Revision 1

    SciTech Connect (OSTI)

    Matthews, Patrick

    2014-01-01

    The purpose of this Corrective Action Decision Document/Closure Report is to provide justification and documentation supporting the recommendation that no further corrective action is needed for CAU 105 based on the implementation of the corrective actions. Corrective action investigation (CAI) activities were performed from October 22, 2012, through May 23, 2013, as set forth in the Corrective Action Investigation Plan for Corrective Action Unit 105: Area 2 Yucca Flat Atmospheric Test Sites; and in accordance with the Soils Activity Quality Assurance Plan, which establishes requirements, technical planning, and general quality practices.

  17. New Pump and Treat Facility Remedial Action Work Plan For Test Area North Final Groundwater Remediation, Operable Unit 1-07B

    SciTech Connect (OSTI)

    Nelson, L. O.

    2007-06-12

    This remedial action work plan identifies the approach and requirements for implementing the medial zone remedial action for Test Area North, Operable Unit 1-07B, at the Idaho National Laboratory. This plan details the management approach for the construction and operation of the New Pump and Treat Facility (NPTF). As identified in the remediatial design/remedial action scope of work, a separate remedial design/remedial action work plan will be prepared for each remedial component of the Operable Unit 1-07B remedial action.

  18. Record of Technical Change No.1 for ``Corrective Action Investigation Plan for Corrective Action Unit 321: Area 22 Weather Station Fuel Storage, Nevada Test Site, Nevada''

    SciTech Connect (OSTI)

    DOE /NV

    1999-02-16

    This Record of Technical Change provides updates to the technical information provided in ''Corrective Action Investigation Plan for Corrective Action Unit 321: Area 22 Weather Station Fuel Storage, Nevada Test Site, Nevada,'' Revision 0. The change specified is in Table 3-1 on page 11. The total lead analyte should specify a Minimum Reporting Limit for soil of 1.0 mg/kg instead of 0.3 mg/kg. The EMAX laboratory cannot meet the 0.3 mg/kg limit.

  19. Corrective Action Investigation Plan for Corrective Action Unit 490: Station 44 Burn Area, Tonopah Test Range, Nevada (with Record of Technical Change No.1)

    SciTech Connect (OSTI)

    U.S. Department of Energy, Nevada Operations Office

    2000-06-09

    This Corrective Action Investigation Plan (CAIP) contains the U.S. Department of Energy, Nevada Operations Office's approach to collect the data necessary to evaluate corrective action alternatives appropriate for the closure of Corrective Action Unit (CAU) 490 under the Federal Facility Agreement and Consent Order. Corrective Active Unit 490 consists of four Corrective Action Sites (CASs): 03-56-001-03BA, Fire Training Area (FTA); RG-56-001-RGBA, Station 44 Burn Area; 03-58-001-03FN, Sandia Service Yard; and 09-54-001-09L2, Gun Propellant Burn Area. These CASs are located at the Tonopah Test Range near Areas 3 and 9. Historically, the FTA was used for training exercises where tires and wood were ignited with diesel fuel. Records indicate that water and carbon dioxide were the only extinguishing agents used during these training exercises. The Station 44 Burn Area was used for fire training exercises and consisted of two wooden structures. The two burn areas (ignition of tires, wood, and wooden structures with diesel fuel and water) were limited to the building footprints (10 ft by 10 ft each). The Sandia Service Yard was used for storage (i.e., wood, tires, metal, electronic and office equipment, construction debris, and drums of oil/grease) from approximately 1979 to 1993. The Gun Propellant Burn Area was used from the 1960s to 1980s to burn excess artillery gun propellant, solid-fuel rocket motors, black powder, and deteriorated explosives; additionally, the area was used for the disposal of experimental explosive items. Based on site history, the focus of the field investigation activities will be to: (1) determine the presence of contaminants of potential concern (COPCs) at each CAS, (2) determine if any COPCs exceed field-screening levels and/or preliminary action levels, and (3) determine the nature and extent of contamination with enough certainty to support selection of corrective action alternatives for each CAS. The scope of this CAIP is to resolve the question of whether or not potentially hazardous wastes were generated at three of the four CASs within CAU 490, and whether or not potentially hazardous and radioactive wastes were generated at the fourth CAS in CAU 490 (CAS 09-54-001-09L2). Suspected CAS-specific COPCs include volatile organic compounds, semivolatile organic compounds, total petroleum hydrocarbons, polychlorinated biphenyls, pesticides, explosives, and uranium and plutonium isotopes. The results of this field investigation will support a defensible evaluation of corrective action alternatives in the corrective action decision document.

  20. Authorized Limits for the Release of a 25 Ton Locomotive, Serial Number 21547, at the Area 25 Engine Maintenance, Assembly, and Disassembly Facility, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    Jeremy Gwin and Douglas Frenette

    2010-04-08

    This document contains process knowledge and radiological data and analysis to support approval for release of the 25-ton locomotive, Serial Number 21547, at the Area 25 Engine Maintenance, Assembly, and Disassembly (EMAD) Facility, located on the Nevada Test Site (NTS). The 25-ton locomotive is a small, one-of-a-kind locomotive used to move railcars in support of the Nuclear Engine for Rocket Vehicle Application project. This locomotive was identified as having significant historical value by the Nevada State Railroad Museum in Boulder City, Nevada, where it will be used as a display piece. A substantial effort to characterize the radiological conditions of the locomotive was undertaken by the NTS Management and Operations Contractor, National Security Technologies, LLC (NSTec). During this characterization process, seven small areas on the locomotive had contamination levels that exceeded the NTS release criteria (limits consistent with U.S. Department of Energy [DOE] Order DOE O 5400.5, “Radiation Protection of the Public and the Environment”). The decision was made to perform radiological decontamination of these known accessible impacted areas to further the release process. On February 9, 2010, NSTec personnel completed decontamination of these seven areas to within the NTS release criteria. Although all accessible areas of the locomotive had been successfully decontaminated to within NTS release criteria, it was plausible that inaccessible areas of the locomotive (i.e., those areas on the locomotive where it was not possible to perform radiological surveys) could potentially have contamination above unrestricted release limits. To access the majority of these inaccessible areas, the locomotive would have to be disassembled. A complete disassembly for a full radiological survey could have permanently destroyed parts and would have ruined the historical value of the locomotive. Complete disassembly would also add an unreasonable financial burden for the contractor. A decision was reached between the NTS regulator and NSTec, opting for alternative authorized limits from DOE Headquarters. In doing so, NSTec personnel performed a dose model using the DOE-approved modeling code RESRAD-BUILD v3.5 to evaluate scenarios. The parameters used in the dose model were conservative. NSTec’s Radiological Engineering Calculation, REC-2010-001, “Public Dose Estimate from the EMAD 25 Ton Locomotive,” concluded that the four scenarios evaluated were below the 25-millirem per year limit, the “likely” dose scenarios met the “few millirem in a year” criteria, and that the EMAD 25-ton locomotive met the radiological requirements to be released with residual radioactivity to the public.

  1. Corrective Action Decision Document/Closure Report for Corrective Action Unit 569: Area 3 Yucca Flat Atmospheric Test Sites Nevada National Security Site, Nevada with ROTC 1, Revision 0

    SciTech Connect (OSTI)

    Sloop, Christy

    2013-04-01

    This Corrective Action Decision Document/Closure Report presents information supporting the closure of Corrective Action Unit (CAU) 569: Area 3 Yucca Flat Atmospheric Test Sites, Nevada National Security Site, Nevada. CAU 569 comprises the following nine corrective action sites (CASs): • 03-23-09, T-3 Contamination Area • 03-23-10, T-3A Contamination Area • 03-23-11, T-3B Contamination Area • 03-23-12, T-3S Contamination Area • 03-23-13, T-3T Contamination Area • 03-23-14, T-3V Contamination Area • 03-23-15, S-3G Contamination Area • 03-23-16, S-3H Contamination Area • 03-23-21, Pike Contamination Area The purpose of this Corrective Action Decision Document/Closure Report is to provide justification and documentation supporting the recommendation that no further corrective action is needed for CAU 569 based on the implementation of the corrective actions listed in Table ES-2.

  2. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Busboom, Herbert J. (San Jose, CA)

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

  3. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect (OSTI)

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  4. Corrective Action Investigation Plan for Corrective Action Unit 106: Areas 5, 11 Frenchman Flat Atmospheric Sites, Nevada Test Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Patrick Matthews

    2010-04-01

    Corrective Action Unit (CAU) 106 is located in Area 5 of the Nevada Test Site, which is approximately 65 miles northwest of Las Vegas, Nevada. Corrective Action Unit 106 comprises the five corrective action sites (CASs) listed below: •05-23-02, GMX Alpha Contaminated Area •05-23-05, Atmospheric Test Site - Able •05-45-01, Atmospheric Test Site - Hamilton •05-45-04, 306 GZ Rad Contaminated Area •05-45-05, 307 GZ Rad Contaminated Area These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives (CAAs). Additional information will be obtained by conducting a corrective action investigation before evaluating CAAs and selecting the appropriate corrective action for each CAS. The results of the field investigation will support a defensible evaluation of viable CAAs that will be presented in the Corrective Action Decision Document. The sites will be investigated based on the data quality objectives (DQOs) developed on January 19, 2010, by representatives of the Nevada Division of Environmental Protection and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office. The DQO process was used to identify and define the type, amount, and quality of data needed to develop and evaluate appropriate corrective actions for CAU 106. The presence and nature of contamination at CAU 106 will be evaluated based on information collected from a field investigation. The CAU includes land areas impacted by the release of radionuclides from a weapons-effect tower test (CAS 05-45-01), a weapons-related airdrop test (CAS 05-23-05), “equation of state” experiments (CAS 05-23-02), and unknown support activities at two sites (CAS 05-45-04 and CAS 05-45-05). Surface-deposited radiological contamination will be evaluated based on a comparison of the total effective dose (TED) at sample plot locations to the dose-based final action level. The TED will be calculated as the total of separate estimates of internal and external doses. Results from the analysis of soil samples collected from sample plots will be used to calculate internal radiological dose. Thermoluminescent dosimeters placed at the center of each sample plot will be used to measure external radiological dose. The presence and nature of contamination from other types of releases (such as migration and excavation as well as any potential releases discovered during the investigation) will be evaluated using soil samples collected from the locations most likely containing contamination, if present. Appendix A provides a detailed discussion of the DQO methodology and the DQOs specific to each CAS. The scope of the corrective action investigation for CAU 106 includes the following activities: •Conduct radiological surveys. •Collect and submit environmental samples for laboratory analysis to determine internal dose rates and the presence of contaminants of concern. •If contaminants of concern are present, collect additional samples to define the extent of the contamination and determine the area where TED at the site exceeds final action levels (i.e., corrective action boundary). •Collect samples of investigation-derived waste, as needed, for waste management purposes.

  5. Gas-cooled fast breeder reactor. Quarterly progress report, November 1, 1979 through January 31, 1980

    SciTech Connect (OSTI)

    Not Available

    1980-02-01

    Information is presented concerning the nuclear steam supply system; reactor core; systems engineering; safety and reliability; and circulator test facility.

  6. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  7. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  8. Addendum to the Closure Report for Corrective Action Unit 398: Area 25 Spill Sites, Nevada Test Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Lynn Kidman

    2008-10-01

    This document constitutes an addendum to the April 2003, Closure Report for Corrective Action Unit 398: Area 25 Spill Sites as described in the document Recommendations and Justifications for Modifications for Use Restrictions Established under the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office Federal Facility Agreement and Consent Order (UR Modification document) dated February 2008. The UR Modification document was approved by NDEP on February 26, 2008. The approval of the UR Modification document constituted approval of each of the recommended UR modifications. In conformance with the UR Modification document, this addendum consists of: • This cover page that refers the reader to the UR Modification document for additional information • The cover and signature pages of the UR Modification document • The NDEP approval letter • The corresponding section of the UR Modification document This addendum provides the documentation justifying the cancellation of the UR for CAS 25-25-17, Subsurface Hydraulic Oil Spill. This UR was established as part of a Federal Facility Agreement and Consent Order (FFACO) corrective action and is based on the presence of contaminants at concentrations greater than the action levels established at the time of the initial investigation (FFACO, 1996; as amended August 2006). Since this UR was established, practices and procedures relating to the implementation of risk-based corrective actions (RBCA) have changed. Therefore, this UR was re-evaluated against the current RBCA criteria as defined in the Industrial Sites Project Establishment of Final Action Levels (NNSA/NSO, 2006c). This re-evaluation consisted of comparing the original data (used to define the need for the UR) to risk-based final action levels (FALs) developed using the current Industrial Sites RBCA process. The re-evaluation resulted in a recommendation to remove the UR because contamination is not present at the site above the risk-based FALs. Requirements for inspecting and maintaining this UR will be canceled, and the postings and signage at this site will be removed. Fencing and posting may be present at this site that are unrelated to the FFACO UR such as for radiological control purposes as required by the NV/YMP Radiological Control Manual (NNSA/NSO, 2004f). This modification will not affect or modify any non-FFACO requirements for fencing, posting, or monitoring at this site.

  9. Safety of transformers, reactors, power supply units and similar products for supply voltages up to 1 100 V part 2-6 : particular requirements and test for safety isolating transformers and power supply units incorporating safety isolating transformers

    E-Print Network [OSTI]

    International Electrotechnical Commission. Geneva

    2009-01-01

    Safety of transformers, reactors, power supply units and similar products for supply voltages up to 1 100 V

  10. Safety of transformers, reactors, power supply units and similar products for supply voltages up to 1 100 V part 2-4 : particular requirements and tests for isolating transformers and power supply units incorporating isolating transformers

    E-Print Network [OSTI]

    International Electrotechnical Commission. Geneva

    2009-01-01

    Safety of transformers, reactors, power supply units and similar products for supply voltages up to 1 100 V

  11. Safety of transformers, reactors, power supply units and similar products for supply voltages up to 1 100 V part 2-13 : particular requirements and tests for auto transformers and power supply units incorporating auto transformers

    E-Print Network [OSTI]

    International Electrotechnical Commission. Geneva

    2009-01-01

    Safety of transformers, reactors, power supply units and similar products for supply voltages up to 1 100 V

  12. Safety of transformers, reactors, power supply units and similar products for supply voltages up to 1 100 V part 2-16 : particular requirements and tests for switch mode power supply units and transformers for switch mode power supply units

    E-Print Network [OSTI]

    International Electrotechnical Commission. Geneva

    2013-01-01

    Safety of transformers, reactors, power supply units and similar products for supply voltages up to 1 100 V

  13. Addendum to the Corrective Action Investigation Plan for Corrective Action Unit 321: Area 22 Weather Station Fuel Storage, Nevada Test Site, Nevada (Rev. 0, November 2000)

    SciTech Connect (OSTI)

    DOE /NV

    2000-11-03

    This addendum to the Corrective Action Investigation Plan (CAIP) contains the U.S. Department of Energy, Nevada Operations Office's approach to determine the extent of contamination existing at Corrective Action Unit (CAU) 321. This addendum was required when the extent of contamination exceeded the estimate in the original Corrective Action Decision Document (CADD). Located in Area 22 on the Nevada Test Site, Corrective Action Unit 321, Weather Station Fuel Storage, consists of Corrective Action Site 22-99-05, Fuel Storage Area, was used to store fuel and other petroleum products necessary for motorized operations at the historic Camp Desert Rock facility. This facility was operational from 1951 to 1958 and dismantled after 1958. Based on site history and earlier investigation activities at CAU 321, the contaminant of potential concern (COPC) was previously identified as total petroleum hydrocarbons (diesel-range organics). The scope of this corrective action investigation for the Fuel Storage Area will include the selection of biased sample locations to determine the vertical and lateral extent of contamination, collection of soil samples using rotary sonic drilling techniques, and the utilization of field-screening methods to accurately determine the extent of COPC contamination. The results of this field investigation will support a defensible evaluation of corrective action alternatives and be included in the revised CADD.

  14. Corrective Action Decision Document/Closure Report for Corrective Action Unit 309: Area 12 Muckpiles, Nevada Test Site, Nevada, Rev. No.: 0 with Errata Sheet

    SciTech Connect (OSTI)

    Alfred Wickline

    2005-12-01

    This Corrective Action Decision Document/Closure Report (CADD/CR) has been prepared for Corrective Action Unit (CAU) 309, Area 12 Muckpiles, Nevada Test Site (NTS), Nevada. The corrective actions proposed in this document are according to the ''Federal Facility Agreement and Consent Order'' (FFACO) that was agreed to by the State of Nevada, U.S. Department of Energy (DOE), and the U.S. Department of Defense (FFACO, 1996). The NTS is approximately 65 miles (mi) northwest of Las Vegas, Nevada (Figure 1-1). Corrective Action Unit 309 is comprised of the three Corrective Action Sites (CASs) (Figure 1-1) listed below: (1) CAS 12-06-09, Muckpile; (2) CAS 12-08-02, Contaminated Waste Dump (CWD); and (3) CAS 12-28-01, I-, J-, and K-Tunnel Debris. Corrective Action Sites 12-06-09 and 12-08-02 will be collectively referred to as muckpiles in this document. Corrective Action Site 12-28-01 will be referred to as the fallout plume because of the extensive lateral area of debris and fallout contamination resulting from the containment failures of the J- and K-Tunnels. A detailed discussion of the history of this CAU is presented in the ''Corrective Action Investigation Plan (CAIP) for Corrective Action Unit 309: Area 12 Muckpiles, Nevada Test Site (NTS), Nevada.'' (NNSA/NSO, 2004). This CADD/CR provides justification for the closure of CAU 309 without further corrective action. This justification is based on process knowledge and the results of the investigative activities conducted according to the CAIP (NNSA/NSO, 2004), which provides information relating to the history, planning, and scope of the investigation. Therefore, this information will not be repeated in this CADD/CR.

  15. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  16. ENGINEERING DEVELOPMENT OF SLURRY BUBBLE COLUMN REACTOR (SBCR) TECHNOLOGY

    SciTech Connect (OSTI)

    Bernard A. Toseland, Ph.D.

    1999-03-01

    The major technical objectives of this program are threefold: (1) to develop the design tools and a fundamental understanding of the fluid dynamics of a slurry bubble column reactor to maximize reactor productivity, (2) to develop the mathematical reactor design models and gain an understanding of the hydrodynamic fundamentals under industrially relevant process conditions, and (3) to develop an understanding of the hydrodynamics and their interaction with the chemistries occurring in the bubble column reactor. Successful completion of these objectives will permit more efficient usage of the reactor column and tighter design criteria, increase overall reactor efficiency, and ensure a design that leads to stable reactor behavior when scaling up to large diameter reactors. The past three months of research have been focused on two major areas of bubble column hydrodynamics: (1) pressure and temperature effects on gas holdup and (2) region transition using a sparger as a gas distributor.

  17. Summary of Natural Resources that Potentially Influence Human Intrusion at the Area 5 Radioactive Waste Management Site, Nevada Test Site, Nye County, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2007-06-01

    In 1993, Raytheon Services Nevada completed a review of natural resource literature and other sources to identify potentially exploitable resources and potential future land uses near the Area 5 Radioactive Waste Management Site (RWMS) of the Nevada Test Site (NTS), Nye County, Nevada, that could lead to future inadvertent human intrusion and subsequent release of radionuclides to the accessible environment. National Security Technologies, LLC, revised the original limited-distribution document to conform to current editorial standards and U.S. Department of Energy requirements for public release. The researchers examined the potential for future development of sand, gravel, mineral, petroleum, water resources, and rural land uses, such as agriculture, grazing, and hunting. The study was part of the performance assessment for Greater Confinement Disposal boreholes. Sand and gravel are not considered exploitable site resources because the materials are common throughout the area and the quality at the Area 5 RWMS is not ideal for typical commercial uses. Site information also indicates a very low mineral potential for the area. None of the 23 mining districts in southern Nye County report occurrences of economic mineral deposits in unconsolidated alluvium. The potential for oil and natural gas is low for southern Nye County. No occurrences of coal, tar sand, or oil shale on the NTS are reported in available literature. Several potential future uses of water were considered. Agricultural irrigation is impractical due to poor soils and existing water supply regulations. Use of water for geothermal energy development is unlikely because temperatures are too low for typical commercial applications using current technology. Human consumption of water has the most potential for cause of intrusion. The economics of future water needs may create a demand for the development of deep carbonate aquifers in the region. However, the Area 5 RWMS is not an optimal location for extraction of groundwater from the deep carbonate aquifer. Grazing and hunting are unlikely to be potential causes for inadvertent human intrusion into waste areas because of vegetation characteristics and lack of significant game animal populations.

  18. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

    1990-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  19. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA); Hui, Marvin M. (Sunnyvale, CA); Berglund, Robert C. (Saratoga, CA)

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  20. UCLA program in reactor studies: The ARIES tokamak reactor study

    SciTech Connect (OSTI)

    Not Available

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  1. A Hydrostratigraphic System for Modeling Groundwater Flow and Radionuclide Migration at the Corrective Action Unit Scale, Nevada Test Site and Surrounding Areas, Clark, Lincoln, and Nye Counties, Nevada

    SciTech Connect (OSTI)

    Prothro, Lance; Drellack Jr., Sigmund; Mercadante, Jennifer

    2009-01-31

    Underground Test Area (UGTA) corrective action unit (CAU) groundwater flow and contaminant transport models of the Nevada Test Site (NTS) and vicinity are built upon hydrostratigraphic framework models (HFMs) that utilize the hydrostratigraphic unit (HSU) as the fundamental modeling component. The delineation and three-dimensional (3-D) modeling of HSUs within the highly complex geologic terrain that is the NTS requires a hydrostratigraphic system that is internally consistent, yet flexible enough to account for overlapping model areas, varied geologic terrain, and the development of multiple alternative HFMs. The UGTA CAU-scale hydrostratigraphic system builds on more than 50 years of geologic and hydrologic work in the NTS region. It includes 76 HSUs developed from nearly 300 stratigraphic units that span more than 570 million years of geologic time, and includes rock units as diverse as marine carbonate and siliciclastic rocks, granitic intrusives, rhyolitic lavas and ash-flow tuffs, and alluvial valley-fill deposits. The UGTA CAU-scale hydrostratigraphic system uses a geology-based approach and two-level classification scheme. The first, or lowest, level of the hydrostratigraphic system is the hydrogeologic unit (HGU). Rocks in a model area are first classified as one of ten HGUs based on the rock’s ability to transmit groundwater (i.e., nature of their porosity and permeability), which at the NTS is mainly a function of the rock’s primary lithology, type and degree of postdepositional alteration, and propensity to fracture. The second, or highest, level within the UGTA CAU-scale hydrostratigraphic system is the HSU, which is the fundamental mapping/modeling unit within UGTA CAU-scale HFMs. HSUs are 3-D bodies that are represented in the finite element mesh for the UGTA groundwater modeling process. HSUs are defined systematically by stratigraphically organizing HGUs of similar character into larger HSUs designations. The careful integration of stratigraphic information in the development of HSUs is important to assure individual HSUs are internally consistent, correlatable, and mappable throughout all the model areas.

  2. Special Analysis of Transuranic Waste in Trench T04C at the Area 5 Radioactive Waste Management Site, Nevada Test Site, Nye County, Nevada, Revision 1

    SciTech Connect (OSTI)

    Greg Shott, Vefa Yucel, Lloyd Desotell

    2008-05-01

    This Special Analysis (SA) was prepared to assess the potential impact of inadvertent disposal of a limited quantity of transuranic (TRU) waste in classified Trench 4 (T04C) within the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site (NTS). The Area 5 RWMS is a low-level radioactive waste disposal site in northern Frenchman Flat on the Nevada Test Site (NTS). The Area 5 RWMS is regulated by the U.S. Department of Energy (DOE) under DOE Order 435.1 and DOE Manual (DOE M) 435.1-1. The primary objective of the SA is to evaluate if inadvertent disposal of limited quantities of TRU waste in a shallow land burial trench at the Area 5 RWMS is in compliance with the existing, approved Disposal Authorization Statement (DAS) issued under DOE M 435.1-1. In addition, supplemental analyses are performed to determine if there is reasonable assurance that the requirements of Title 40, Code of Federal Regulations (CFR), Part 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level, and Transuranic Radioactive Wastes, can be met. The 40 CFR 191 analyses provide supplemental information regarding the risk to human health and the environment of leaving the TRU waste in T04C. In 1989, waste management personnel reviewing classified materials records discovered that classified materials buried in trench T04C at the Area 5 RWMS contained TRU waste. Subsequent investigations determined that a total of 102 55-gallon drums of TRU waste from Rocky Flats were buried in trench T04C in 1986. The disposal was inadvertent because unclassified records accompanying the shipment indicated that the waste was low-level. The exact location of the TRU waste in T04C was not recorded and is currently unknown. Under DOE M 435.1-1, Chapter IV, Section P.5, low-level waste disposal facilities must obtain a DAS. The DAS specifies conditions that must be met to operate within the radioactive waste management basis, consisting of a performance assessment (PA), composite analysis (CA), closure plan, monitoring plan, waste acceptance criteria, and a PA/CA maintenance plan. The DOE issued a DAS for the Area 5 RWMS in 2000. The Area 5 RWMS DAS was, in part, based on review of a CA as required under DOE M 435.1-1, Chapter IV, Section P.(3). A CA is a radiological assessment required for DOE waste disposed before 26 September 1988 and includes the radiological dose from all sources of radioactive material interacting with all radioactive waste disposed at the Area 5 RWMS. The approved Area 5 RWMS CA, which includes the inventory of TRU waste in T04C, indicates that the Area 5 RWMS waste inventory and all interacting sources of radioactive material can meet the 0.3 mSv dose constraint. The composite analysis maximum annual dose for a future resident at the Area 5 RWMS was estimated to be 0.01 mSv at 1,000 years. Therefore, the inadvertent disposal of TRU in T04C is protective of the public and the environment, and compliant with all the applicable requirements in DOE M 435.1-1 and the DAS. The U.S. Environmental Protection Agency promulgated 40 CFR 191 to establish standards for the planned disposal of spent nuclear fuel, high level, and transuranic wastes in geologic repositories. Although not required, the National Nuclear Security Administration Nevada Site Office requested a supplemental analysis to evaluate the likelihood that the inadvertent disposal of TRU waste in T04C meets the requirements of 40 CFR 191. The SA evaluates the likelihood of meeting the 40 CFR 191 containment requirements (CRs), assurance requirements, individual protection requirements (IPRs), and groundwater protection standards. The results of the SA indicate that there is a reasonable expectation of meeting all the requirements of 40 CFR 191. The conclusion of the SA is that the Area 5 RWMS with the TRU waste buried in T04C is in compliance with all requirements in DOE M 435.1-1 and the DAS. Compliance with the DAS is demonstrated by the results of the Area 5 RWMS CA. Supplemental analyses in the SA indicate there is a

  3. Closure Report for Corrective Action Unit 118: Area 27 Super Kukla Facility, Nevada Test Site, Nevada with ROTC 1, Revision 0

    SciTech Connect (OSTI)

    Mark Burmeister

    2007-09-01

    This CR provides documentation and justification for the closure of CAU 118 without further corrective action. This justification is based on process knowledge and the results of the investigative and closure activities conducted in accordance with the CAU 118 SAFER Plan: Streamlined Approach for Environmental Restoration (SAFER) Plan for CAU 118: Area 27 Super Kukla Facility, Nevada Test Site, Nevada (NNSA/NSO, 2006). The SAFER Plan provides information relating to site history as well as the scope and planning of the investigation. This CR also provides the analytical and radiological survey data to confirm that the remediation goals were met as specified in the CAU 118 SAFER Plan (NNSA/NSO, 2006). The Nevada Division of Environmental Protection (NDEP) approved the CAU 118 SAFER Plan (Murphy, 2006), which recommends closure in place with use restrictions (URs).

  4. Post-Closure Inspection Report for Corrective Action Unit 424: Area 3 Landfill Complex, Tonopah Test Range, Nevada, Calendar Year 2000

    SciTech Connect (OSTI)

    K. B. Campbell

    2001-06-01

    Corrective Action Unit (CAU) 424, the Area 3 Landfill Complex at Tonopah Test Range, consists of eight landfill sites, Corrective Action Sites (CASS), seven of which are landfill cells that were closed previously by capping. (The eighth CAS, A3-7, was not used as a landfill site and was closed without taking any corrective action.) Figure 1 shows the locations of the landfill cells. CAU 424 closure activities included removing small volumes of soil containing petroleum hydrocarbons, repairing cell covers that were cracked or had subsided, and installing above-grade and at-grade monuments marking the comers of the landfill cells. Post-closure monitoring requirements for CAU 424 are detailed in Section 5.0, Post-Closure Inspection Plan contained, in the Closure Report for Corrective Action Unit 424: Area 3 Landfill Complex, Tonopah Test Range. Nevada, report number DOE/NV--283. The Closure Report (CR) was approved by the Nevada Division of Environmental Protection (NDEP) in July 1999. The CR includes compaction and permeability results of soils that cap the seven landfill cells. Post-closure monitoring consists of the following: (1) Site inspections done twice a year to evaluate the condition of the unit; (2) Verification that the site is secure; (3) Notice of any subsidence or deficiencies that may compromise the integrity of the unit; (4) Remedy of any deficiencies within 90 days of discovery; and (5) Preparation and submittal of an annual report. Site inspections were conducted on June 20, 2000, and November 20, 2000. The inspections were preformed after the NDEP approval of the CR. This report includes copies of the inspection checklist and photographs, and recommendations and conclusions. The Post-Closure Inspection Checklists are found in Attachment A, a copy of the field notes is found in Attachment B, and copies of the inspection photographs are found in Attachment C.

  5. Post-Closure Inspection Report for Corrective Action Unit 453: Area 9 UXO Landfill Tonopah Test Range, Nevada, Calendar Year 2000

    SciTech Connect (OSTI)

    K. B. Campbell

    2001-06-01

    Post-closure monitoring requirements for the Area 9 Unexploded Ordnance Landfill (Corrective Action Unit [CAU] 453) (Figure 1) are described in Closure Report for Corrective Action Unit 453: Area 9 UXO Landfill, Tonopah Test Range, Nevada, report number DOE/NV--284. The Closure Report (CR) was submitted to the Nevada Division of Environmental Protection (NDEP) on August 5,1999. The CR (containing the Post-Closure Monitoring Plan) was approved by the NDEP on September 10,1999. Post-closure monitoring at CAU 453 consists of the following: (1) Site inspections done twice a year to evaluate the condition of the unit; (2) Verification that the site is secure; (3) Notice of any subsidence or deficiencies that may compromise the integrity of the unit; (4) Remedy of any deficiencies within 90 days of discovery; and (5) Preparation and submittal of an annual report. Site inspections were conducted on June 20, 2000 and November 21, 2000. Both site inspections were conducted after NDEP approval of the CR, and in accordance with the Post-Closure Monitoring Plan in the NDEP-approved CR. This report includes copies of the inspection checklists, photographs, recommendations, and conclusions. The Post-Closure Inspection Checklists are found in Attachment A, a copy of the field notes is found in Attachment B, and inspection photographs are found in Attachment C.

  6. Post-Closure Inspection Report for Corrective Action Unit 453: Area 9 UXO Landfill Tonopah Test Range, Nevada, Calendar Year 2001

    SciTech Connect (OSTI)

    K. B. Campbell

    2002-01-01

    Post-closure monitoring requirements for the Area 9 Unexploded Ordinance Landfill (Corrective Action Unit [CAU] 453) (Figure 1) are described in Closure Report for Corrective Action Unit 453: Area 9 UXO Landfill, Tonopah Test Range, Nevada, report number DOE/NV--284, August 1999. The Closure Report (CR) was submitted to the Nevada Division of Environmental Protection (NDEP) on August 5 , 1999. The CR (containing the Post-Closure Monitoring Plan) was approved by the NDEP on September 10,1999. As stated in Section 5.0 of the NDEP-approved CR, post-closure monitoring at CAU 453 consists of the following: (1) Visual site inspections are conducted twice a year to evaluate the condition of the cover. (2) Verification that the site is secure and the condition of the fence and posted warning signs. (3) Notice of any subsidence, erosion, unauthorized excavation, etc., deficiencies that may compromise the integrity of the unit. (4) Remedy of any deficiencies within 90 days of discovery. (5) Preparation and submittal of an annual report. Site inspections were conducted on May 15, 2001 and November 6, 2001. Both site inspections were conducted in accordance with the Post-Closure Monitoring Plan in the NDEP-approved CR. This report includes copies of the inspection checklists, photographs, recommendations, and conclusions. The Post-Closure Inspection Checklists are found in Attachment A, a copy of the field notes is found in Attachment B, and inspection photographs are found in Attachment C.

  7. Post-Closure Inspection and Monitoring Report for Corrective Action Unit 417: Central Nevada Test Area Surface, Hot Creek Valley, Nevada

    SciTech Connect (OSTI)

    2009-01-01

    This report presents data collected during the annual post-closure site inspection conducted at the Central Nevada Test Area Surface Corrective Action Unit (CAU) 417 in May of 2008. The annual post-closure site inspection included inspections of the UC-1, UC-3, and UC-4 sites in accordance with the Post-Closure Monitoring Plan provided in the CAU 417 Closure Report (NNSA/NV 2001). The annual inspection conducted at the UC-1 Central Mud Pit (CMP) indicated that the site and soil cover were in good condition. Three new cracks or fractures were observed in the soil cover during the annual inspection and were immediately filled with bentonite chips. The vegetation on the soil cover was adequate, but showed signs of the area's ongoing drought. No issues were identified with the CMP fence, gate, or subsidence monuments. No issues were identified with the warning signs and monuments at the other two UC-1 locations. The annual subsidence survey was conducted at UC-1 CMP and UC-4 Mud Pit C in August 2008. The results of the subsidence surveys indicate that the covers are performing as expected, and no unusual subsidence was observed.

  8. Testing Surface Area Pravesh Kothari

    E-Print Network [OSTI]

    O'Donnell, Ryan

    Program of China Grant 2011CBA00300, 2011CBA00301, the National Natural Science Foundation of China Grant

  9. CENTRAL NEVPJJA SUPPLEMENTAL TEST AREA

    Office of Legacy Management (LM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield Municipal Gas &SCE-SessionsSouth DakotaRobbins and700, 1. .&. 'explainsBurrell,C: LI c1 T -r r

  10. Corrective Action Investigation Plan for Corrective Action Unit 570: Area 9 Yucca Flat Atmospheric Test Sites Nevada National Security Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Patrick Matthews

    2012-08-01

    CAU 570 comprises the following six corrective action sites (CASs): • 02-23-07, Atmospheric Test Site - Tesla • 09-23-10, Atmospheric Test Site T-9 • 09-23-11, Atmospheric Test Site S-9G • 09-23-14, Atmospheric Test Site - Rushmore • 09-23-15, Eagle Contamination Area • 09-99-01, Atmospheric Test Site B-9A These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives (CAAs). Additional information will be obtained by conducting a corrective action investigation before evaluating CAAs and selecting the appropriate corrective action for each CAS. The results of the field investigation will support a defensible evaluation of viable CAAs that will be presented in the Corrective Action Decision Document. The sites will be investigated based on the data quality objectives (DQOs) developed on April 30, 2012, by representatives of the Nevada Division of Environmental Protection and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office. The DQO process was used to identify and define the type, amount, and quality of data needed to develop and evaluate appropriate corrective actions for CAU 570. The site investigation process will also be conducted in accordance with the Soils Activity Quality Assurance Plan, which establishes requirements, technical planning, and general quality practices to be applied to this activity. The presence and nature of contamination at CAU 570 will be evaluated based on information collected from a field investigation. Radiological contamination will be evaluated based on a comparison of the total effective dose at sample locations to the dose-based final action level. The total effective dose will be calculated as the total of separate estimates of internal and external dose. Results from the analysis of soil samples will be used to calculate internal radiological dose. Thermoluminescent dosimeters placed near the center of each sample location will be used to measure external radiological dose. Appendix A provides a detailed discussion of the DQO methodology and the DQOs specific to each CAS.

  11. Tokamak reactor startup power

    SciTech Connect (OSTI)

    Weldon, D.M.; Murray, J.G.

    1983-01-01

    Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor (ETR).

  12. Attrition reactor system

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  13. Attrition reactor system

    DOE Patents [OSTI]

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  14. Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide

    SciTech Connect (OSTI)

    Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

    2012-09-01

    This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

  15. Measurement of transmission efficiency for 400 MeV proton beam through collimator at Fermilab MuCool Test Area using Chromox-6 scintillation screen

    SciTech Connect (OSTI)

    Jana, M. R.; Chung, M.; Leonova, M.; Moretti, A.; Palmer, M.; Schwarz, T.; Tollestrup, A.; Yonehara, K. [Fermi National Accelerator Laboratory, Batavia, Illinois 60510 (United States); Freemire, B.; Hanlet, P.; Torun, Y. [Illinois Institute of Technology, Chicago, Illinois 60616 (United States)

    2013-06-15

    The MuCool Test Area (MTA) at Fermilab is a facility to develop the technology required for ionization cooling for a future Muon Collider and/or Neutrino Factory. As part of this research program, feasibility studies of various types of RF cavities in a high magnetic field environment are in progress. As a unique approach, we have tested a RF cavity filled with a high pressure hydrogen gas with a 400 MeV proton beam in an external magnetic field (B = 3 T). Quantitative information about the number of protons passing through this cavity is an essential requirement of the beam test. The MTA is a flammable gas (hydrogen) hazard zone. Due to safety reasons, no active (energized) beam diagnostic instrument can be used. Moreover, when the magnetic field is on, current transformers (toroids) used for beam intensity measurements do not work due to the saturation of the ferrite material of the transformer. Based on these requirements, we have developed a passive beam diagnostic instrumentation using a combination of a Chromox-6 scintillation screen and CCD camera. This paper describes details of the beam profile and position obtained from the CCD image with B = 0 T and B = 3 T, and for high and low intensity proton beams. A comparison is made with beam size obtained from multi-wires detector. Beam transmission efficiency through a collimator with a 4 mm diameter hole is measured by the toroids and CCD image of the scintillation screen. Results show that the transmission efficiency estimated from the CCD image is consistent with the toroid measurement, which enables us to monitor the beam transmission efficiency even in a high magnetic field environment.

  16. Post-Closure Inspection and Monitoring Report for Corrective Action Unit 417: Central Nevada Test Area Surface, Hot Creek Valley, Nevada, for Calendar Year 2007

    SciTech Connect (OSTI)

    2008-09-01

    This report presents data collected during the annual post-closure site inspection conducted at the Central Nevada Test Area Surface Corrective Action Unit (CAU) 417 in May 2007. The annual post-closure site inspection included inspections of the UC-1, UC-3, and UC-4 sites in accordance with the Post-Closure Monitoring Plan provided in the CAU 417 Closure Report (NNSA/NV 2001). The annual inspection conducted at the UC-1 Central Mud Pit (CMP) indicated the site and soil cover were in good condition. No new cracks or fractures were observed in the soil cover during the annual inspection. A crack on the west portion of the cover was observed during the last quarterly inspection in December 2006. This crack was filled with bentonite as part of the maintenance activities conducted in February 2007 and will be monitored during subsequent annual inspections. The vegetation on the soil cover was adequate but showing signs of the area's ongoing drought. No issues were identified with the CMP fence, gate, or subsidence monuments. New DOE Office of Legacy Management signs with updated emergency phone numbers were installed as part of this annual inspection, no issues were identified with the warning signs and monuments at the other two UC-1 locations. The annual subsidence survey was conducted at UC-1 CMP and UC-4 Mud Pit C as part of the maintenance activities conducted in February 2007. The results of the subsidence surveys indicate that the covers are performing as expected, and no unusual subsidence was observed. A vegetation survey of the UC-1 CMP cover and adjacent areas was conducted as part of the annual inspection in May 2007. The vegetation survey indicated that revegetation continues to be successful, although stressed due to the area's prevailing drought conditions. The vegetation should continue to be monitored to document any changes in the plant community and to identify conditions that could potentially require remedial action to maintain a viable vegetation cover on the site. It is suggested that future vegetation surveys be conducted once every 2 years or as needed to help monitor the health of the vegetation.

  17. Nuclear Reactor Safety Design Criteria

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1993-01-19

    The order establishes nuclear safety criteria applicable to the design, fabrication, construction, testing, and performance requirements of nuclear reactor facilities and safety class structures, systems, and components (SSCs) within these facilities. Cancels paragraphs 8a and 8b of DOE 5480.6. Cancels DOE O 5480.6 in part. Supersedes DOE 5480.1, dated 1-19-93. Certified 11-18-10.

  18. Power-reactor fuel-pin thermomechanics

    SciTech Connect (OSTI)

    Tutnov, A.A.; Ul'yanov, A.I.

    1987-11-01

    The authors describe a method for determining the creep and elongation and other aspects of mechanical behavior of fuel pins and cans under the effects of irradiation and temperature encountered in reactors under loading and burnup conditions. An exhaustive method for testing for fuel-cladding interactions is described. The methodology is shown to be applicable to the design, fabrication, and loading of pins for WWER, SGHWR, and RBMK type reactors, from which much of the experimental data were derived.

  19. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  20. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  1. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  2. Reactor technology. Progress report, July-September 1980

    SciTech Connect (OSTI)

    Breslow, M.

    1980-12-01

    Progress in the Space Power Advanced Reactor (SPAR) Program includes indications that revision of the BeO reflector configuration can reduce system weight. Observed boiling limit restrictions on the performance of the annular-wick core heat pipe have accelerated transition to the development of the target-design arterial heat pipe. Successful bends of core heat pipes have been made with sodium as the mandrel material. With the phasing out of the GCFR Program, work on the Low Power Safety Experiments Program is now concentrated on completion of the third 37-rod Full Length Subgroup test. In the Reactor Safety/Structural Analysis area, effort on the Category I Structures Program is toward developing an experimental test plan focusing on a specific structural design. Buckling experiments on thin-walled cylindrical shells with circular cutouts are reported. Results of a three-dimensional analysis of thermal stresses in the Fort St. Vrain core support block are presented. Materials investigations and operation of a molybdenum-core SiC heat pipe are reported. Entrainment limits for gravity-assisted heat pipes and heat pipe configurations for application to energy conservation are being investigated. The new solution critical assembly, SHEBA, was completed. Godiva IV was temporarily relocated at TA-15. Influence of scattered radiations in the test vault on InRad measurements was determined from detector scans of the vault produced by /sup 252/Cf neutron and /sup 137/Cs gamma sources.

  3. Corrective Action Investigation Plan for Corrective Action Unit 105: Area 2 Yucca Flat Atmospheric Test Sites Nevada National Security Site, Nevada, Revision 0

    SciTech Connect (OSTI)

    Patrick Matthews

    2012-09-01

    Corrective Action Unit (CAU) 105 is located in Area 2 of the Nevada National Security Site, which is approximately 65 miles northwest of Las Vegas, Nevada. CAU 105 is a geographical grouping of sites where there has been a suspected release of contamination associated with atmospheric nuclear testing. This document describes the planned investigation of CAU 105, which comprises the following corrective action sites (CASs): • 02-23-04, Atmospheric Test Site - Whitney • 02-23-05, Atmospheric Test Site T-2A • 02-23-06, Atmospheric Test Site T-2B • 02-23-08, Atmospheric Test Site T-2 • 02-23-09, Atmospheric Test Site - Turk These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives (CAAs). Additional information will be obtained by conducting a corrective action investigation before evaluating CAAs and selecting the appropriate corrective action for each CAS. The results of the field investigation will support a defensible evaluation of viable CAAs that will be presented in the Corrective Action Decision Document. The sites will be investigated based on the data quality objectives (DQOs) developed on April 30, 2012, by representatives of the Nevada Division of Environmental Protection and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office. The DQO process was used to identify and define the type, amount, and quality of data needed to develop and evaluate appropriate corrective actions for CAU 105. The site investigation process will also be conducted in accordance with the Soils Activity Quality Assurance Plan, which establishes requirements, technical planning, and general quality practices to be applied to this activity. The potential contamination sources associated with all CAU 105 CASs are from atmospheric nuclear testing activities. The presence and nature of contamination at CAU 105 will be evaluated based on information collected from a field investigation. Radiological contamination will be evaluated based on a comparison of the total effective dose at sample locations to the dose-based final action level. The total effective dose will be calculated as the total of separate estimates of internal and external dose. Results from the analysis of soil samples will be used to calculate internal radiological dose. Thermoluminescent dosimeters placed at the center of each sample location will be used to measure external radiological dose. Appendix A provides a detailed discussion of the DQO methodology and the DQOs specific to each CAS. This Corrective Action Investigation Plan has been developed in accordance with the Federal Facility Agreement and Consent Order that was agreed to by the State of Nevada; DOE, Environmental Management; U.S. Department of Defense; and DOE, Legacy Management. Under the Federal Facility Agreement and Consent Order, this Corrective Action Investigation Plan will be submitted to the Nevada Division of Environmental Protection for approval. Fieldwork will be conducted after the plan is approved.

  4. SAVANNAH RIVER SITE R-REACTOR DISASSEMBLY BASIN IN-SITU DECOMMISSIONING -10499

    SciTech Connect (OSTI)

    Langton, C.; Serrato, M.; Blankenship, J.; Griffin, W.

    2010-01-04

    The US DOE concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate intact, structurally sound facilities that are no longer needed for their original purpose, i.e., generating (reactor facilities), processing(isotope separation facilities) or storing radioactive materials. The 105-R Disassembly Basin is the first SRS reactor facility to undergo the in-situ decommissioning (ISD) process. This ISD process complies with the 105-R Disassembly Basin project strategy as outlined in the Engineering Evaluation/Cost Analysis for the Grouting of the R-Reactor Disassembly Basin at the Savannah River Site and includes: (1) Managing residual water by solidification in-place or evaporation at another facility; (2) Filling the below grade portion of the basin with cementitious materials to physically stabilize the basin and prevent collapse of the final cap - Sludge and debris in the bottom few feet of the basin will be encapsulated between the basin floor and overlying fill material to isolate it from the environment; (3) Demolishing the above grade portion of the structure and relocating the resulting debris to another location or disposing of the debris in-place; and (4) Capping the basin area with a concrete slab which is part of an engineered cap to prevent inadvertent intrusion. The estimated total grout volume to fill the 105-R Reactor Disassembly Basin is 24,384 cubic meters or 31,894 cubic yards. Portland cement-based structural fill materials were designed and tested for the reactor ISD project, and a placement strategy for stabilizing the basin was developed. Based on structural engineering analyses and material flow considerations, maximum lift heights and differential height requirements were determined. Pertinent data and information related to the SRS 105-R Reactor Disassembly Basin in-situ decommissioning include: regulatory documentation, residual water management, area preparation activities, technology needs, fill material designs and testing, and fill placement strategy. This information is applicable to decommissioning both the 105-P and 105-R facilities. The ISD process for the entire 105-P and 105-R reactor facilities will require approximately 250,000 cubic yards (191,140 cubic meters) of grout and approximately 3,900 cubic yards (2,989 cubic meters) of structural concrete which will be placed over about an eighteen month period to meet the accelerated schedule ISD schedule. The status and lessons learned in the SRS Reactor Facility ISD process will be described.

  5. CORRRECTIVE ACTION DECISION DOCUMENT FOR CORRECTIVE ACTION UNIT 427: AREA 3 SEPTIC WASTE SYSTEMS 2 AND 6, TONOPAH TEST RANGE, NEVADA, REVISION 0, JUNE 1998

    SciTech Connect (OSTI)

    DOE /NV

    1998-06-23

    This Corrective Action Decision Document has been prepared for the Area 3 Septic Waste Systems 2 and 6 (Corrective Action Unit 427) in accordance with the Federal Facility Agreement and Consent Order of 1996 (FFACO, 1996). Corrective Action Unit 427 is located at the Tonopah Test Range, Nevada, and is comprised of the following Corrective Action Sites, each an individual septic waste system (DOE/NV, 1996a): (1) Septic Waste System 2 is Corrective Action Site Number 03-05-002-SW02. (2) Septic Waste System 6 is Corrective Action Site Number 03-05-002-SW06. The purpose of this Corrective Action Decision Document is to identify and provide a rationale for the selection of a recommended corrective action alternative for each Corrective Action Site. The scope of this Correction Action Decision Document consists of the following tasks: (1) Develop corrective action objectives. (2) Identify corrective action alternative screening criteria. (3) Develop corrective action alternatives. (4) Perform detailed and comparative evaluations of the corrective action alternatives in relation to the corrective action objectives and screening criteria. (5) Recommend and justify a preferred corrective action alternative for each CAS. From November 1997 through January 1998, a corrective action investigation was performed as set forth in the Corrective Action Investigation Plan for Corrective Action Unit No. 427: Area 3 Septic Waste System Numbers 2 and 6, Tonopah Test Range, Nevada (DOE/NV, 1997b). Details can be found in Appendix A of this document. The results indicated that contamination is present in some portions of the CAU and not in others as described in Table ES-1 and shown in Figure A.2-2 of Appendix A. Based on the potential exposure pathways, the following corrective action objectives have been identified for Corrective Action Unit 427: (1) Prevent or mitigate human exposure to subsurface soils containing TPH at concentrations greater than 100 milligrams per kilogram (NAC, 1996b). (2) Close Sep tic Tank 33-5 in accordance with Nevada Administrative Code 459 (NAC, 1996c). (3) Prevent adverse impacts to groundwater quality. Based on the review of existing data, future land use, and current operations at the Tonopah Test Range, the following alternatives were developed for consideration at the Area 3 Septic Waste Systems 2 and 6: Alternative 1 - No Further Action; Alternative 2 - Closure of Septic Tank 33-5 and Administrative Controls; Alternative 3 - Closure of Septic Tank 33-5, Excavation, and Disposal The corrective action alternatives were evaluated based on four general corrective action standards and five remedy selection decision factors. Based on the results of this evaluation, the preferred alternative for Corrective Action Unit 427 is Alternative 2, Closure of Septic Tank 33-5 and Administrative Controls. The preferred corrective action alternative was evaluated on technical merit, focusing on performance, reliability, feasibility, and safety. The alternative was judged to meet all requirements for the technical components evaluated. The alternative meets all applicable state and federal regulations for closure of the site and will reduce potential future exposure pathways to the contaminated soils. During corrective action implementation, this alternative will present minimal potential threat to site workers who come in contact with the waste. However, procedures will be developed and implemented to ensure worker health and safety.

  6. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  7. Corrective Action Investigation Plan for Corrective Action Unit 34: Area 3 Contaminated Waste Site, Nevada Test Site, Nevada (Rev. 0, March 2001)

    SciTech Connect (OSTI)

    U.S. Department of Energy, National Nuclear Security Administration Nevada Operations Office

    2001-03-27

    This Corrective Action Investigation Plan contains the U.S. Department of Energy, National Nuclear Security Administration Nevada Operations Office's approach to collect the data necessary to evaluate corrective action alternatives appropriate for the closure of Corrective Action Unit (CAU) 34 under the Federal Facility Agreement and Consent Order. Corrective Action Unit 34 consists of four Corrective Action Sites (CASs). The CAU is located within the Area 3 Compound at the Nevada Test Site (NTS) in the vicinity of the Mud Plant Facility in Yucca Valley. Historically, CAS 03-09-07, Mud Pit, was used for disposal of excess mud from washing drilling equipment from 1968 to 1974, at which time it began to be used for excess mud disposal (currently inactive); CAS 03-44-01, Chromium Contamination Spill, was used to store additives used in the formulation of drilling mud from the early 1960s to the mid-1990s; CAS 03-47-02, Area 3 Mud Plant Pond, was used as a freshwater storage reservoir for the mud plant as well as supplied water for a number of activities including the mixing of mud, the rinsing and cleaning of tanks, and various washdowns from the 1960s through 1990s; and CAS 03-09-06, Mud Disposal Crater, was created in 1962 by an underground nuclear detonation (i.e., Chinchilla test) and was used to mix and store mud, dispose of receiving waste from the mud plant floor drains and excess drilling mud, and clean/flush mix tanks through the mid-1990s. Based on site history, the scope of this plan is to identify potentially contaminated ground soil at each of the four CASs and determine the quantity, nature, and extent of contaminants of potential concern (COPCs). The investigation will include systematic and biased surface and subsurface soil and mud sampling using hand-auguring and direct-push techniques; visual, video, and/or electromagnetic surveys of pipes; field screening for volatile organic compounds (VOCs) and alpha/beta-emitting radionuclides; and laboratory analysis to characterize any investigation-derived waste for disposal both on site at NTS and at off-site locations. Historical information provided by former NTS employees indicates that COPCs include VOCs, semivolatile organic compounds, Resource Conservation and Recovery Act metals, petroleum hydrocarbons, gamma-emitting radionuclides, isotopic plutonium, and strontium-90. The results of this field investigation will support a defensible evaluation of corrective action alternatives in the corrective action decision document.

  8. Post-Closure Inspection Report for Corrective Action Unit 424: Area 3 Landfill Complexes Tonopah Test Range, Nevada Calendar Year 2001

    SciTech Connect (OSTI)

    K. B. Campbell

    2002-02-01

    Corrective Action Unit (CAU) 424, the Area 3 Landfill Complexes at Tonopah Test Range, consists of eight Corrective Action Sites (CASs), seven of which are landfill cells that were closed previously by capping. (The eighth CAS, A3-7, was not used as a landfill site and was closed without taking any corrective action.) Figure 1 shows the general location of the landfill cells. Figure 2 shows in more detail the location of the eight landfill cells. CAU 424 closure activities included removing small volumes of soil containing petroleum hydrocarbons, repairing cell covers that were cracked or had subsided, and installing above-grade and at-grade monuments marking the comers of the landfill cells. Post-closure monitoring requirements for CAU 424 are detailed in Section 5.0, Post-Closure Inspection Plan, contained in the Closure Report for Corrective Action Unit 424: Area 3 Landfill Complexes, Tonopah Test Range, Nevada, report number DOE/NV--283, July 1999. The Closure Report (CR) was approved by the Nevada Division of Environmental Protection (NDEP) in July 1999. The CR includes compaction and permeability results of soils that cap the seven landfill cells. As stated in Section 5.0 of the NDEP-approved CR, post-closure monitoring at CAU 424 consists of the following: (1) Site inspections conducted twice a year to evaluate the condition of the unit. (2) Verification that landfill markers and warning signs are in-place, intact, and readable. (3) Notice of any subsidence, erosion, unauthorized use, or deficiencies that may compromise the integrity of the landfill covers. (4) Remedy of any deficiencies within 90 days of discovery. (5) Preparation and submittal of an annual report. Site inspections were conducted on May 16, 2001, and November 6, 2001. The inspections were preformed after the NDEP approval of the CR. This report includes copies of the inspection checklist, photographs, recommendations, and conclusions. The Post-Closure Inspection Checklists are found in Attachment A, a copy of the field notes is found in Attachment B, and copies of the inspection photographs are found in Attachment C.

  9. Sub-crop geologic map of pre-Tertiary rocks in the Yucca Flat and northern Frenchman Flat areas, Nevada Test Site, southern Nevada

    SciTech Connect (OSTI)

    Cole, J.C.; Harris, A.G.; Wahl, R.R.

    1997-10-02

    This map displays interpreted structural and stratigraphic relations among the Paleozoic and older rocks of the Nevada Test Site region beneath the Miocene volcanic rocks and younger alluvium in the Yucca Flat and northern Frenchman Flat basins. These interpretations are based on a comprehensive examination and review of data for more than 77 drillholes that penetrated part of the pre-Tertiary basement beneath these post-middle Miocene structural basins. Biostratigraphic data from conodont fossils were newly obtained for 31 of these holes, and a thorough review of all prior microfossil paleontologic data is incorporated in the analysis. Subsurface relationships are interpreted in light of a revised regional geologic framework synthesized from detailed geologic mapping in the ranges surrounding Yucca Flat, from comprehensive stratigraphic studies in the region, and from additional detailed field studies on and around the Nevada Test Site. All available data indicate the subsurface geology of Yucca Flat is considerably more complicated than previous interpretations have suggested. The western part of the basin, in particular, is underlain by relics of the eastward-vergent Belted Range thrust system that are folded back toward the west and thrust by local, west-vergent contractional structures of the CP thrust system. Field evidence from the ranges surrounding the north end of Yucca Flat indicate that two significant strike-slip faults track southward beneath the post-middle Miocene basin fill, but their subsurface traces cannot be closely defined from the available evidence. In contrast, the eastern part of the Yucca Flat basin is interpreted to be underlain by a fairly simple north-trending, broad syncline in the pre-Tertiary units. Far fewer data are available for the northern Frenchman Flat basin, but regional analysis indicates the pre-Tertiary structure there should also be relatively simple and not affected by thrusting. This new interpretation has implications for ground water flow through pre-Tertiary rocks beneath the Yucca Flat and northern Frenchman Flat areas, and has consequences for ground water modeling and model validation. Our data indicate that the Mississippian Chainman Shale is not laterally extensive confining unit in the western part of the basin because it is folded back onto itself by the convergent structures of the Belted Range and CP thrust systems. Early and Middle Paleozoic limestone and dolomite are present beneath most of both basins and, regardless of structural complications, are interpreted to form a laterally continuous and extensive carbonate aquifer. Structural culmination that marks the French Peak accommodation zone along the topographic divide between the two basins provides a lateral pathway through highly fractured rock between the volcanic aquifers of Yucca Flat and the regional carbonate aquifer. This pathway may accelerate the migration of ground-water contaminants introduced by underground nuclear testing toward discharge areas beyond the Nevada Test Site boundaries. Predictive three-dimensional models of hydrostratigraphic units and ground-water flow in the pre-Tertiary rocks of subsurface Yucca Flat are likely to be unrealistic due to the extreme structural complexities. The interpretation of hydrologic and geochemical data obtained from monitoring wells will be difficult to extrapolate through the flow system until more is known about the continuity of hydrostratigraphic units. 1 plate

  10. High Flux Metallic Membranes for Hydrogen Recovery and Membrane Reactors

    SciTech Connect (OSTI)

    Buxbaum, Robert

    2010-06-30

    We made and tested over 250 new alloys for use as lower cost, higher flux hydrogen extraction membrane materials. Most of these were intermetallic, or contained significant intermetallic content, particularly based on B2 alloy compositions with at least one refractory component; B2 intermetallics resemble BCC alloys, in structure, but the atoms have relatively fixed positions, with one atom at the corners of the cube, the other at the centers. The target materals we were looking for would contain little or no expensive elements, no strongly toxic or radioactive elements, would have high flux to hydrogen, while being fabricable, brazable, and relatively immune to hydrogen embrittlement and corrosion in operation. The best combination of properties of the membrane materials we developed was, in my opinion, a Pd-coated membrane consisting of V -9 atomic % Pd. This material was relatively cheap, had 5 times the flux of Pd under the same pressure differential, was reasonably easy to fabricate and braze, and not bad in terms of embrittlement. Based on all these factors we project, about 1/3 the cost of Pd, on an area basis for a membrane designed to last 20 years, or 1/15 the cost on a flux basis. Alternatives to this membrane replaced significant fractions of the Pd with Ni and or Co. The cost for these membranes was lower, but so was the flux. We produced successful brazed products from the membrane materials, and made them into flat sheets. We tested, unsuccessfully, several means of fabricating thematerials into tubes, and eventually built a membrane reactor using a new, flat-plate design: a disc and doughnut arrangement, a design that seems well- suited to clean hydrogen production from coal. The membranes and reactor were tested successfully at Western Research. A larger equipment company (Chart Industries) produced similar results using a different flat-plate reactor design. Cost projections of the membrane are shown to be attractive.

  11. Spinning fluids reactor

    DOE Patents [OSTI]

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  12. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  13. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M. (San Jose, CA); Taft, William E. (Los Gatos, CA)

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  14. A Study and Comparison of SCR Reaction Kinetics from Reactor and Engine Experimental Data

    Broader source: Energy.gov [DOE]

    Presents experimental study of a Cu-zeolite SCR in both reactor and engine test cell, and comparison of the model parameters between the SCR reactor and engine model

  15. Corrective Action Investigation Plan for Corrective Action Unit 552: Area 12 Muckpile and Ponds, Nevada Test Site, Nevada, Rev. No.: 1 with ROTC 1 and 2

    SciTech Connect (OSTI)

    David A. Strand

    2005-01-01

    This Corrective Action Investigation Plan (CAIP) contains project-specific information including facility descriptions, environmental sample collection objectives, and criteria for conducting site investigation activities at Corrective Action Unit (CAU) 552: Area 12 Muckpile and Ponds, Nevada Test Site (NTS), Nevada. This CAIP has been developed in accordance with the ''Federal Facility Agreement and Consent Order'' (FFACO) (1996) that was agreed to by the State of Nevada, the U.S. Department of Energy (DOE), and the U.S. Department of Defense. The NTS is approximately 65 miles (mi) northwest of Las Vegas, Nevada (Figure 1-1). Corrective Action Unit 552 is comprised of the one Corrective Action Site which is 12-23-05, Ponds. One additional CAS, 12-06-04, Muckpile (G-Tunnel Muckpile), was removed from this CAU when it was determined that the muckpile is an active site. A modification to the FFACO to remove CAS 12-06-04 was approved by the Nevada Division of Environmental Protection (NDEP) on December 16, 2004. The G-Tunnel ponds were first identified in the 1991 Reynolds Electrical & Engineering Co., Inc. document entitled, ''Nevada Test Site Inventory of Inactive and Abandoned Facilities and Waste Sites'' (REECo, 1991). Corrective Action Unit 552 is being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives. Therefore, additional information will be obtained by conducting a corrective action investigation (CAI) prior to evaluating and selecting the corrective action alternatives for the site. The CAI will include field inspections, radiological surveys, and sampling of appropriate media. Data will also be obtained to support investigation-derived waste (IDW) disposal and potential future waste management decisions.

  16. USE OF CEMENTITIOUS MATERIALS FOR SRS REACTOR FACILITY IN-SITU DECOMMISSIONING - 11620

    SciTech Connect (OSTI)

    Langton, C.; Stefanko, D.; Serrato, M.; Blankenship, J.; Griffin, W.; Waymer, J.; Matheny, D.; Singh, D.

    2010-12-07

    The United States Department of Energy (US DOE) concept for facility in-situ decommissioning (ISD) is to physically stabilize and isolate in tact, structurally sound facilities that are no longer needed for their original purpose of, i.e., producing (reactor facilities), processing (isotope separation facilities) or storing radioactive materials. The Savannah River Site 105-P and 105-R Reactor Facility ISD requires about 250,000 cubic yards of grout to fill the below grade structure. The fills are designed to prevent subsidence, reduce water infiltration, and isolate contaminated materials. This work is being performed as a Comprehensive Environmental Response, Compensations and Liability Act (CERCLA) action and is part of the overall soil and groundwater completion projects for P- and R-Areas. Cementitious materials were designed for the following applications: (1) Below grade massive voids/rooms: Portland cement-based structural flowable fills for - Bulk filling, Restricted placement and Underwater placement. (2) Special below grade applications for reduced load bearing capacity needs: Cellular portland cement lightweight fill (3) Reactor vessel fills that are compatible with reactive metal (aluminum metal) components in the reactor vessels: Calcium sulfoaluminate flowable fill, and Magnesium potassium phosphate flowable fill. (4) Caps to prevent water infiltration and intrusion into areas with the highest levels of radionuclides: Portland cement based shrinkage compensating concrete. A system engineering approach was used to identify functions and requirements of the fill and capping materials. Laboratory testing was performed to identify candidate formulations and develop final design mixes. Scale-up testing was performed to verify material production and placement as well as fresh and cured properties. The 105-P and 105-R ISD projects are currently in progress and are expected to be complete in 2012. The focus of this paper is to describe the (1) grout mixes for filling the reactor vessels, and (2) a specialty grout mix to fill a selected portion of the P-Reactor Disassembly Basin. Details of the grout mixes designed for ISD of he SRS Reactor Disassembly Basins and below grade portions of the 105-Buildings was described elsewhere. Material property test results, placement strategies, full-scale production and delivery systems will also be described.

  17. Addendum to the Corrective Action Decision Document/Closure Report for Corrective Action Unit 529: Area 25 Contaminated Materials, Nevada Test Site, Nevada, Revision 1

    SciTech Connect (OSTI)

    Krauss, Mark J

    2013-10-01

    This document constitutes an addendum to the Corrective Action Decision Document/Closure Report for Corrective Action Unit 529: Area 25 Contaminated Materials, Nevada Test Site, Nevada as described in the document Recommendations and Justifications To Remove Use Restrictions Established under the U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office Federal Facility Agreement and Consent Order dated September 2013. The Use Restriction (UR) Removal document was approved by the Nevada Division of Environmental Protection on October 16, 2013. The approval of the UR Removal document constituted approval of each of the recommended UR removals. In conformance with the UR Removal document, this addendum consists of: This page that refers the reader to the UR Removal document for additional information The cover, title, and signature pages of the UR Removal document The NDEP approval letter The corresponding section of the UR Removal document This addendum provides the documentation justifying the cancellation of the UR for CAS 25-23-17, Contaminated Wash (Parcel H). This UR was established as part of FFACO corrective actions and was based on the presence of total petroleum hydrocarbon diesel-range organics contamination at concentrations greater than the NDEP action level at the time of the initial investigation.

  18. Corrective Action Decision Document for Corrective Action Unit 168: Area 25 and 26 Contaminated Materials and Waste Dumps, Nevada Test Site, Nevada, Rev. No.: 2 with Errata Sheet

    SciTech Connect (OSTI)

    Wickline, Alfred

    2006-12-01

    This Corrective Action Decision Document has been prepared for Corrective Action Unit (CAU) 168: Area 25 and 26, Contaminated Materials and Waste Dumps, Nevada Test Site, Nevada. The purpose of this Corrective Action Decision Document is to identify and provide a rationale for the selection of a recommended corrective action alternative for each corrective action site (CAS) within CAU 168. The corrective action investigation (CAI) was conducted in accordance with the ''Corrective Action Investigation Plan for Corrective Action Unit 168: Area 25 and 26, Contaminated Materials and Waste Dumps, Nevada Test Site, Nevada'', as developed under the ''Federal Facility Agreement and Consent Order'' (1996). Corrective Action Unit 168 is located in Areas 25 and 26 of the Nevada Test Site, Nevada and is comprised of the following 12 CASs: CAS 25-16-01, Construction Waste Pile; CAS 25-16-03, MX Construction Landfill; CAS 25-19-02, Waste Disposal Site; CAS 25-23-02, Radioactive Storage RR Cars; CAS 25-23-13, ETL - Lab Radioactive Contamination; CAS 25-23-18, Radioactive Material Storage; CAS 25-34-01, NRDS Contaminated Bunker; CAS 25-34-02, NRDS Contaminated Bunker; CAS 25-99-16, USW G3; CAS 26-08-01, Waste Dump/Burn Pit; CAS 26-17-01, Pluto Waste Holding Area; and CAS 26-19-02, Contaminated Waste Dump No.2. Analytes detected during the CAI were evaluated against preliminary action levels (PALs) to determine contaminants of concern (COCs) for CASs within CAU 168. Radiological measurements of railroad cars and test equipment were compared to unrestricted (free) release criteria. Assessment of the data generated from the CAI activities revealed the following: (1) Corrective Action Site 25-16-01 contains hydrocarbon-contaminated soil at concentrations exceeding the PAL. The contamination is at discrete locations associated with asphalt debris. (2) No COCs were identified at CAS 25-16-03. Buried construction waste is present in at least two disposal cells contained within the landfill boundaries. (3) No COCs were identified at CAS 25-19-02. (4) Radiological surveys at CAS 25-23-02 identified 13 railroad cars that exceeded the NV/YMP Radiological Control Manual limits for free release. Six railroad cars were below these limits and therefore met the free-release criteria. (5) An In-Situ Object Counting System survey taken at CAS 25-23-02 identified two railroad cars possibly containing fuel fragments; both exceeded the NV/YMP Radiological Control Manual free release criteria. (6) Corrective Action Site 25-23-18 contains total petroleum hydrocarbons-diesel-range organics, Aroclor-1260, uranium-234, uranium-235, strontium-90, and cesium-137 that exceed PALs. (7) Radiological surveys at CAS 25-34-01 indicate that there were no total contamination readings that exceeded the NV/YMP Radiological Control Manual limits for free release. (8) Radiological surveys at CAS 25-34-02 indicate that there were no total contamination readings that exceeded the NV/YMP Radiological Control Manual limits for free release. (9) Radiological surveys at CAS 25-23-13 identified six pieces of equipment that exceed the NV/YMP Radiological Control Manual limits for free release. (10) Corrective Action Site 25-99-16 was not investigated. A review of historical documentation and current site conditions showed that no further characterization was required to select the appropriate corrective action. (11) Corrective Action Site 26-08-01 contains hydrocarbon-contaminated soil at concentrations exceeding the PAL. The contamination is at discrete locations associated with asphalt debris. (12) Corrective Action Site 26-17-01 contains total petroleum hydrocarbons-diesel-range organics and Aroclor-1260 exceeding the PALs. (13) Radiological surveys at CAS 26-19-02 identified metallic debris that exceeded the NV/YMP Radiological Control Manual limits for free release. Concentrations of radiological or chemical constituents in soil did not exceed PALs.

  19. Post-Closure Inspection and Monitoring Report for Corrective Action Unit 417: Central Nevada Test Area Surface, Hot Creek Valley, Nevada

    SciTech Connect (OSTI)

    None

    2013-03-01

    This report presents results of data collected during the annual post-closure site inspections conducted at the Central Nevada Test Area surface Corrective Action Unit (CAU) 417 in May 2011 and July 2012. The annual post-closure site inspections included inspections of the UC-1, UC-3, and UC-4 sites in accordance with the Post-Closure Monitoring Plan provided in the CAU 417 Closure Report (NNSA/NV 2001). The annual inspections conducted at the UC-1 Central Mud Pit (CMP) indicated that the site and soil cover were in good condition. No new fractures or extension of existing fractures were observed and no issues with the fence or gate were identified. The vegetation on the cover continues to look healthy, but the biennial vegetation survey conducted during the 2012 inspection indicated that the total foliar cover was slightly higher in 2009 than in 2012. This may be indicative of a decrease in precipitation observed during the 2-year monitoring period. The precipitation totaled 9.9 inches from July 1, 2010, through June 30, 2011, and 5 inches from July 1, 2011, through June 30, 2012. This decrease in precipitation is also evident in the soil moisture data obtained from the time domain reflectometry sensors. Soil moisture content data show that the UC-1 cover is performing as designed, and evapotranspiration is effectively removing water from the cover.

  20. Relap5-3d model validation and benchmark exercises for advanced gas cooled reactor application 

    E-Print Network [OSTI]

    Moore, Eugene James Thomas

    2006-08-16

    HTTR High Temperature engineering Test Reactor INET Institute of Nuclear Energy Technology LWR Light Water Reactor OKBM Test Design Bureau for Machine Building ORNL Oak Ridge National Laboratory RCCS Reactor Cavity Cooling System.... This code used a card-based input deck and was designed to simulate transients in light water reactors (LWR) including, but not limited to, a loss of coolant accident or a station blackout. In the thirty years since, the code has been continuously...

  1. 100-B area technical baseline report

    SciTech Connect (OSTI)

    Carpenter, R.W.

    1994-09-01

    This document supports the environmental remediation effort of the 100-B Area by providing remediation planners with key data that characterize the 100-B and 100-C Reactor sites. It provides operational histories of the 100-B and 100-C Reactors and each of their associated liquid and solid waste sites.

  2. Liquid metal cooled nuclear reactors with passive cooling system

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Fanning, Alan W. (San Jose, CA)

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  3. TESTING AND ACCEPTANCE OF FUEL PLATES FOR RERTR FUEL DEVELOPMENT EXPERIMENTS

    SciTech Connect (OSTI)

    J.M. Wight; G.A. Moore; S.C. Taylor

    2008-10-01

    This paper discusses how candidate fuel plates for RERTR Fuel Development experiments are examined and tested for acceptance prior to reactor insertion. These tests include destructive and nondestructive examinations (DE and NDE). The DE includes blister annealing for dispersion fuel plates, bend testing of adjacent cladding, and microscopic examination of archive fuel plates. The NDE includes Ultrasonic (UT) scanning and radiography. UT tests include an ultrasonic scan for areas of “debonds” and a high frequency ultrasonic scan to determine the "minimum cladding" over the fuel. Radiography inspections include identifying fuel outside of the maximum fuel zone and measurements and calculations for fuel density. Details of each test are provided and acceptance criteria are defined. These tests help to provide a high level of confidence the fuel plate will perform in the reactor without a breach in the cladding.

  4. Generating unstructured nuclear reactor core meshes in parallel

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore »examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a Ľ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

  5. Generating unstructured nuclear reactor core meshes in parallel

    SciTech Connect (OSTI)

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a Ľ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.

  6. Pressurized melt ejection into scaled reactor cavities

    SciTech Connect (OSTI)

    Tarbell, W.W.; Pilch, M.; Brockmann, J.E.; Ross, J.W.; Gilbert, D.W.

    1986-10-01

    This report describes four tests performed in the High-Pressure Melt Streaming Program (HIPS) using linear-scaled cavities of the Zion Nuclear Power Plant. These experiments were conducted to study the phenomena involved in high-pressure ejection of core debris into the cavity beneath the reactor pressure vessel. One-tenth and one-twentieth linear scale models of reactor cavities were constructed and instrumented. The first test used an apparatus constructed of alumina firebrick to minimize the potential interaction between the ejected melt and cavity material. The remaining three experiments used scaled representations of the Zion nuclear plant geometry, constructed of prototypic concrete composition.

  7. Site Environmental Report for Calendar Year 2009. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

    SciTech Connect (OSTI)

    Liu, Ning; Rutherford, Phil; Amar, Ravnesh

    2010-09-01

    This Annual Site Environmental Report (ASER) for 2009 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2009 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

  8. Annual Site Environmental Report, Department of Energy Operations at the Energy Technology Engineering Center – Area IV, Santa Susana Field Laboratory

    SciTech Connect (OSTI)

    Frazee, Brad; Hay, Scott; Wondolleck, John; Sorrels, Earl; Rutherford, Phil; Dassler, David; Jones, John

    2015-05-01

    This Annual Site Environmental Report (ASER) for 2014 describes the environmental conditions related to work performed for the DOE at Area IV of the Santa Susana Field Laboratory (SSFL). The ETEC, a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

  9. Site Environmental Report for Calendar Year 2010. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

    SciTech Connect (OSTI)

    Liu, Ning; Rutherford, Phil; Amar, Ravnesh

    2011-09-01

    This Annual Site Environmental Report (ASER) for 2010 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2010 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

  10. Site Environmental Report For Calendar Year 2012. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

    SciTech Connect (OSTI)

    Liu, Ning; Rutherford, Phil; Dassler, David

    2013-09-01

    This Annual Site Environmental Report (ASER) for 2012 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2012 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

  11. Site Environmental Report for Calendar Year 2011. DOE Operations at The Boeing Company Santa Susana Field Laboratory, Area IV

    SciTech Connect (OSTI)

    Liu, Ning; Rutherford, Phil; Dassler, David

    2012-09-01

    This Annual Site Environmental Report (ASER) for 2011 describes the environmental conditions related to work performed for the Department of Energy (DOE) at Area IV of Boeing’s Santa Susana Field Laboratory (SSFL). The Energy Technology Engineering Center (ETEC), a government-owned, company-operated test facility, was located in Area IV. The operations in Area IV included development, fabrication, operation and disassembly of nuclear reactors, reactor fuel, and other radioactive materials. Other activities in the area involved the operation of large-scale liquid metal facilities that were used for testing non-nuclear liquid metal fast breeder reactor components. All nuclear work was terminated in 1988, and all subsequent radiological work has been directed toward environmental restoration and decontamination and decommissioning (D&D) of the former nuclear facilities and their associated sites. Liquid metal research and development ended in 2002. Since May 2007, the D&D operations in Area IV have been suspended by the DOE, but the environmental monitoring and characterization programs have continued. Results of the radiological monitoring program for the calendar year 2011 continue to indicate that there are no significant releases of radioactive material from Area IV of SSFL. All potential exposure pathways are sampled and/or monitored, including air, soil, surface water, groundwater, direct radiation, transfer of property (land, structures, waste), and recycling.

  12. POST-CLOSURE INSPECTION AND MONITORING REPORT FOR CORRECTIVE ACTION UNIT 112: AREA 23 HAZARDOUS WASTE TRENCHES, NEVADA TEST SITE, NEVADA; FOR THE PERIOD OCTOBER 2003 - SEPTEMBER 2004

    SciTech Connect (OSTI)

    BECHTEL NEVADA

    2004-12-01

    Corrective Action Unit (CAU) 112, Area 23 Hazardous Waste Trenches, Nevada Test Site (NTS), Nevada, is a Resource Conservation and Recovery Act (RCRA) unit located in Area 23 of the NTS. This annual Post-Closure Inspection and Monitoring Report provides the results of inspections and monitoring for CAU 112. This report includes a summary and analysis of the site inspections, repair and maintenance, meteorological information, and neutron soil moisture monitoring data obtained at CAU 112 for the current monitoring period, October 2003 through September 2004. Inspections of the CAU 112 RCRA unit were performed quarterly to identify any significant physical changes to the site that could impact the proper operation of the waste unit. The overall condition of the covers and facility was good, and no significant findings were observed. The annual subsidence survey of the elevation markers was conducted on August 23, 2004, and the results indicated that no cover subsidence4 has occurred at any of the markers. The elevations of the markers have been consistent for the past 11 years. The total precipitation for the current reporting period, october 2003 to September 2004, was 14.0 centimeters (cm) (5.5 inches [in]) (National Oceanographic and Atmospheric Administration, Air Resources Laboratory, Special Operations and Research Division, 2004). This is slightly below the average rainfall of 14.7 cm (5.79 in) over the same period from 1972 to 2004. Post-closure monitoring verifies that the CAU 112 trench covers are performing properly and that no water is infiltrating into or out of the waste trenches. Sail moisture measurements are obtained in the soil directly beneath the trenches and compared to baseline conditions for the first year of post-closure monitoring, which began in october 1993. neutron logging was performed twice during this monitoring period along 30 neutron access tubes to obtain soil moisture data and detect any changes that may indicate moisture movement beneath each trench. Soil moisture results obtained to date indicate that the compliance criterion of less than 5% Residual Volumetric Moisture Content was met. Soil conditions remain dry and stable beneath the trenches, and the cover is functioning as designed within the compliance limits.

  13. Corrective Action Decision Document for Corrective Action Unit 271: Areas 25, 26, and 27 Septic Systems, Nevada Test Site, Nevada, Rev. 0

    SciTech Connect (OSTI)

    NNSA /NV

    2002-09-16

    This corrective action decision document (CADD) identifies and rationalizes the U.S. Department of Energy, National Nuclear Security Administration Nevada Operations Office's selection of a recommended corrective action alternative (CAA) appropriate to facilitate the closure of Corrective Action Unit (CAU) 271, Areas 25, 26, and 27 Septic Systems, Nevada Test Site (NTS), Nevada, under the Federal Facility Agreement and Consent Order (FFACO). Located on the NTS approximately 65 miles northwest of Las Vegas, CAU 271 consists of fifteen Corrective Action Sites (CASs). The CASs consist of 13 septic systems, a radioactive leachfield, and a contaminated reservoir. The purpose of this CADD is to identify and provide a rationale for the selection of a recommended CAA for each CAS within CAU 271. Corrective action investigation (CAI) activities were performed from October 29, 2001, through February 22, 2002, and April 29, 2002, through June 25, 2002. Analytes detected during the CAI were evaluated against preliminary action levels and regulatory disposal limits to determine contaminants of concern (COC) for each CAS. It was determined that contaminants of concern included hydrocarbon-contaminated media, polychlorinated biphenyls, and radiologically-contaminated media. Three corrective action objectives were identified for these CASs, and subsequently three CAAs developed for consideration based on a review of existing data, future use, and current operations in Areas 25, 26, and 27 of the NTS. These CAAs were: Alternative 1 - No Further Action, Alternative 2 - Clean Closure, and Alternative 3 - Closure in Place with Administrative Controls. Alternative 2, Clean Closure, was chosen as the preferred CAA for all but two of the CASs (25-04-04 and 27-05-02) because Nevada Administrative Control 444.818 requires clean closure of the septic tanks involved with these CASs. Alternative 3, Closure in Place, was chosen for the final two CASs because the short-term risks of transportation were eliminated; long-term health risks were minimized by preventing public and worker access to contaminated media through physical and administrative controls; all waste will be managed in accordance with federal, state, and local requirements; and a cost-effective method for achieving protection and meeting closure requirements is provided. All CAAs were evaluated on technical merit focusing on performance, reliability, feasibility, and safety. These two alternatives were judged to meet all applicable state and federal regulations for closure of the site and reduced the potential for future exposure pathways to the contaminated soils at this site.

  14. Testing the scaling of thermal transport models: predicted and measured temperatures in the Tokamak Fusion Test

    E-Print Network [OSTI]

    in the Tokamak Fusion Test Reactor dimensionless scaling experiments D. R. Mikkelsen, S. D. Scott Princeton the Tokamak Fusion Test Reactor [D. J. Grove and D. M. Meade, Nucl. Fusion 25, 1167 (1985)] nondimensional to International Tokamak Experimental Reactor [2] (ITER) class tokamaks. This paper compares the predictions

  15. Space reactor safety, 1985--1995 lessons learned

    SciTech Connect (OSTI)

    Marshall, A.C.

    1995-12-31

    Space reactor safety activities and decisions have evolved over the last decade. Important safety decisions have been made in the SP-100, Space Exploration Initiative, NEPSTP, SNTP, and Bimodal Space Reactor programs. In addition, international guidance on space reactor safety has been instituted. Space reactor safety decisions and practices have developed in the areas of inadvertent criticality, reentry, radiological release, orbital operation, programmatic, and policy. In general, the lessons learned point out the importance of carefully reviewing previous safety practices for appropriateness to space nuclear programs in general and to the specific mission under consideration.

  16. Addendum to the Closure Report for Corrective Action Unit 326: Areas 6 and 27 Release Sites, Nevada Test Site, Nevada, Revision 1

    SciTech Connect (OSTI)

    Grant Evenson

    2009-05-01

    This document constitutes an addendum to the Closure Report for Corrective Action Unit 326: Areas 6 and 27 Release Sites, Nevada Test Site, Nevada (Revision 1), December 2002 as described in the document Supplemental Investigation Report for FFACO Use Restrictions, Nevada Test Site, Nevada (SIR) dated November 2008. The SIR document was approved by NDEP on December 5, 2008. The approval of the SIR document constituted approval of each of the recommended UR removals. In conformance with the SIR document, this addendum consists of: • This page that refers the reader to the SIR document for additional information • The cover, title, and signature pages of the SIR document • The NDEP approval letter • The corresponding section of the SIR document This addendum provides the documentation justifying the cancellation of the UR for CAS 06-25-01, CP-1 Heating Oil Release. This UR was established as part of a Federal Facility Agreement and Consent Order (FFACO) corrective action and is based on the presence of contaminants at concentrations greater than the action levels established at the time of the initial investigation (FFACO, 1996). Since this UR was established, practices and procedures relating to the implementation of risk-based corrective actions (RBCA) have changed. Therefore, this UR was reevaluated against the current RBCA criteria as defined in the Industrial Sites Project Establishment of Final Action Levels (NNSA/NSO, 2006). This re-evaluation consisted of comparing the original data (used to define the need for the UR) to risk-based final action levels (FALs) developed using the current Industrial Sites RBCA process. The re-evaluation resulted in a recommendation to remove the UR because contamination is not present at the site above the risk-based FALs. Requirements for inspecting and maintaining this UR will be canceled, and the postings and signage at this site will be removed. Fencing and posting may be present at this site that are unrelated to the FFACO UR such as for radiological control purposes as required by the NV/YMP Radiological Control Manual (NNSA/NSO, 2004). This modification will not affect or modify any non-FFACO requirements for fencing, posting, or monitoring at this site.

  17. Addendum 2 to the Closure Report for Corrective Action Unit 358: Areas 18, 19, 20 Cellars/Mud Pits, Nevada Test Site, Nevada, Revison 0

    SciTech Connect (OSTI)

    Grant Evenson

    2009-05-01

    This document constitutes an addendum to the Closure Report for Corrective Action Unit 358: Areas 18, 19, 20 Cellars/Mud Pits, Nevada Test Site, Nevada, January 2004 as described in the document Supplemental Investigation Report for FFACO Use Restrictions, Nevada Test Site, Nevada (SIR) dated November 2008. The SIR document was approved by NDEP on December 5, 2008. The approval of the SIR document constituted approval of each of the recommended UR removals. In conformance with the SIR document, this addendum consists of: • This page that refers the reader to the SIR document for additional information • The cover, title, and signature pages of the SIR document • The NDEP approval letter • The corresponding section of the SIR document This addendum provides the documentation justifying the cancellation of the UR for CAS 19-09-05, Mud Pit. This UR was established as part of a Federal Facility Agreement and Consent Order (FFACO) corrective action and is based on the presence of contaminants at concentrations greater than the action levels established at the time of the initial investigation (FFACO, 1996). Since this UR was established, practices and procedures relating to the implementation of risk-based corrective actions (RBCA) have changed. Therefore, this UR was reevaluated against the current RBCA criteria as defined in the Industrial Sites Project Establishment of Final Action Levels (NNSA/NSO, 2006). This re-evaluation consisted of comparing the original data (used to define the need for the UR) to risk-based final action levels (FALs) developed using the current Industrial Sites RBCA process. The re-evaluation resulted in a recommendation to remove the UR because contamination is not present at the site above the risk-based FALs. Requirements for inspecting and maintaining this UR will be canceled, and the postings and signage at this site will be removed. Fencing and posting may be present at this site that are unrelated to the FFACO UR such as for radiological control purposes as required by the NV/YMP Radiological Control Manual (NNSA/NSO, 2004). This modification will not affect or modify any non-FFACO requirements for fencing, posting, or monitoring at this site.

  18. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  19. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

  20. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01

    A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.