Sample records for test reactor area

  1. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    SciTech Connect (OSTI)

    NONE

    1994-10-01T23:59:59.000Z

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition.

  2. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01T23:59:59.000Z

    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. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28T23:59:59.000Z

    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.

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

    Broader source: Energy.gov (indexed) [DOE]

    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 land...

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

    Energy Savers [EERE]

    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...

  6. SRS reactor stack plume marking tests

    SciTech Connect (OSTI)

    Petry, S.F.

    1992-03-01T23:59:59.000Z

    Tests performed in 105-K in 1987 and 1988 demonstrated that the stack plume can successfully be made visible (i.e., marked) by introducing smoke into the stack breech. The ultimate objective of these tests is to provide a means during an emergency evacuation so that an evacuee can readily identify the stack plume and evacuate in the opposite direction, thus minimizing the potential of severe radiation exposure. The EPA has also requested DOE to arrange for more tests to settle a technical question involving the correct calculation of stack downwash. New test canisters were received in 1988 designed to produce more smoke per unit time; however, these canisters have not been evaluated, because normal ventilation conditions have not been reestablished in K Area. Meanwhile, both the authorization and procedure to conduct the tests have expired. The tests can be performed during normal reactor operation. It is recommended that appropriate authorization and procedure approval be obtained to resume testing after K Area restart.

  7. Advanced Test Reactor Capabilities and Future Irradiation Plans

    SciTech Connect (OSTI)

    Frances M. Marshall

    2006-10-01T23:59:59.000Z

    The Advanced Test Reactor (ATR), located at the Idaho National Laboratory (INL), is one of the most versatile operating research reactors in the Untied States. The ATR has a long history of supporting reactor fuel and material research for the US government and other test sponsors. The INL is owned by the US Department of Energy (DOE) and currently operated by Battelle Energy Alliance (BEA). The ATR is the third generation of test reactors built at the Test Reactor Area, now named the Reactor Technology Complex (RTC), whose mission is to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The current experiments in the ATR are for a variety of customers--US DOE, foreign governments and private researchers, and commercial companies that need neutrons. The ATR has several unique features that enable the reactor to perform diverse simultaneous tests for multiple test sponsors. The ATR has been operating since 1967, and is expected to continue operating for several more decades. The remainder of this paper discusses the ATR design features, testing options, previous experiment programs, future plans for the ATR capabilities and experiments, and some introduction to the INL and DOE's expectations for nuclear research in the future.

  8. STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver

    Broader source: Energy.gov (indexed) [DOE]

    Advance Test Reactor Class Waiver W(C)-2008-004 The Advanced Test Reactor (A TR) is a pressurized water test reactor at the Idaho National Laboratory (INL) that operates at low...

  9. Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01T23:59:59.000Z

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  10. Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment

    Broader source: Energy.gov [DOE]

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

  11. The advanced test reactor strategic evaluation program

    SciTech Connect (OSTI)

    Buescher, B.J.; Majumdar, D.; Croucher, D.W.

    1989-01-01T23:59:59.000Z

    Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed.

  12. Startup testing of Romania dual-core test reactor

    SciTech Connect (OSTI)

    Whittemore, W.L. [General Atomic Co., San Diego, CA (United States)

    1980-07-01T23:59:59.000Z

    Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

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

    SciTech Connect (OSTI)

    S. Blaine Grover

    2004-10-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

  14. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    SciTech Connect (OSTI)

    Grover, S.B.

    2004-10-06T23:59:59.000Z

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations.

  15. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    SciTech Connect (OSTI)

    Bragg-Sitton, Shannon M. [NASA Marshall Space Flight Center, Nuclear and Advanced Propulsion Branch, ER-11, MSFC, AL 35812 (United States); Morton, T. J. [Department of Chemical and Nuclear Engineering, University of New Mexico, Albuquerque, NM 87131 (United States)

    2006-01-20T23:59:59.000Z

    Non-nuclear testing can be a valuable tool in the development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe (HP) cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system. Reactivity feedback calculations were then based on a bulk reactivity feedback coefficient and measured average core temperature. This paper presents preliminary results from similar dynamic testing of a direct drive gas cooled reactor system (DDG), demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. Although the HP and DDG designs both utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility. Planned system upgrades to allow implementation of higher fidelity dynamic testing are also discussed. Proposed DDG testing will utilize a higher fidelity point kinetics model to control core power transients, and reactivity feedback will be based on localized feedback coefficients and several independent temperature measurements taken within the core block. This paper presents preliminary test results and discusses the methodology that will be implemented in follow-on DDG testing and the additional instrumentation required to implement high fidelity dynamic testing.

  16. HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2011-10-13T23:59:59.000Z

    The Heavy Water Components Test Reactor (HWCTR) Decommissioning Project was initiated in 2009 as a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) Removal Action with funding from the American Recovery and Reinvestment Act (ARRA). This paper summarizes the history prior to 2009, the major D&D activities, and final end state of the facility at completion of decommissioning in June 2011. The HWCTR facility was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In the early 1990s, DOE began planning to decommission HWCTR. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. In 2009 the $1.6 billion allocation from the ARRA to SRS for site footprint reduction at SRS reopened the doors to HWCTR - this time for final decommissioning. Alternative studies concluded that the most environmentally safe, cost effective option for final decommissioning was to remove the reactor vessel, both steam generators, and all equipment above grade including the dome. The transfer coffin, originally above grade, was to be placed in the cavity vacated by the reactor vessel and the remaining below grade spaces would be grouted. Once all above equipment including the dome was removed, a concrete cover was to be placed over the remaining footprint and the groundwater monitored for an indefinite period to ensure compliance with environmental regulations.

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

    SciTech Connect (OSTI)

    S. Blaine Grover

    2006-10-01T23:59:59.000Z

    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.

  18. Advanced Test Reactor Testing Experience: Past, Present and Future

    SciTech Connect (OSTI)

    Frances M. Marshall

    2005-04-01T23:59:59.000Z

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 48" long and 5.0" diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors -- US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, wherein the target material is placed in a capsule, or plate form, and the capsule is in direct contact with the primary coolant. The next level of complexity of an experiment is an instrumented lead experiment, which allows for active monitoring and control of experiment conditions during the irradiation. The highest level of complexity of experiment is the pressurized water loop experiment, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans.

  19. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect (OSTI)

    Not Available

    1988-05-01T23:59:59.000Z

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  20. 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

    2009-09-01T23:59:59.000Z

    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.

  1. 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

    2008-12-16T23:59:59.000Z

    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.

  2. Beryllium Use in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Glen R. Longhurst

    2007-12-01T23:59:59.000Z

    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.

  3. application test reactor: Topics by E-print Network

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

    flow reactors (OFRs). A micro-bioreactor (working (more) Reis, N. 2006-01-01 4 Web Application Testing with Customized Test Requirements Computer Technologies and...

  4. Decommissioning of the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    E. Perry; J. Chrzanowski; C. Gentile; R. Parsells; K. Rule; R. Strykowsky; M. Viola

    2003-10-28T23:59:59.000Z

    The Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory was operated from 1982 until 1997. The last several years included operations with mixtures of deuterium and tritium. In September 2002, the three year Decontamination and Decommissioning (D&D) Project for TFTR was successfully completed. The need to deal with tritium contamination as well as activated materials led to the adaptation of many techniques from the maintenance work during TFTR operations to the D&D effort. In addition, techniques from the decommissioning of fission reactors were adapted to the D&D of TFTR and several new technologies, most notably the development of a diamond wire cutting process for complex metal structures, were developed. These techniques, along with a project management system that closely linked the field crews to the engineering staff who developed the techniques and procedures via a Work Control Center, resulted in a project that was completed safely, on time, and well below budget.

  5. Automated Test Coverage Measurement for Reactor Protection System Software

    E-Print Network [OSTI]

    Automated Test Coverage Measurement for Reactor Protection System Software Implemented in Function in implementing safety critical systems such as nuclear reactor protection systems. We have defined new structural- ing a case study using test cases prepared by domain experts for reactor protection system software

  6. Space power reactor ground test in the Experimental Gas Cooled Reactor (EGCR) at Oak Ridge

    SciTech Connect (OSTI)

    Fontana, M.H.; Holcomb, R.S.; Cooper, R.H.

    1992-08-01T23:59:59.000Z

    The Experimental Gas Cooled Reactor (EGCR) facility and the supporting technical infrastructure at the Oak Ridge National Laboratory have the capabilities of performing ground tests of space nuclear power reactor systems. A candidate test would be a 10 MWt lithium cooled reactor, generating potassium vapor that would drive a power turbine. The facility is a large containment vessel originally intended to test the EGCR. Large, contained, and shielded spaces are available for testing, assembly, disassembly, and post-test examination.

  7. Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

    SciTech Connect (OSTI)

    Wright, A.E. (Argonne National Lab., IL (USA)); Dutt, D.S. (Westinghouse Hanford Co., Richland, WA (USA)); Harrison, L.J. (Argonne National Lab., Idaho Falls, ID (USA))

    1990-01-01T23:59:59.000Z

    Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs.

  8. TREAT (Transient Reactor Test Facility) reactor control rod scram system simulations and testing

    SciTech Connect (OSTI)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01T23:59:59.000Z

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs.

  9. Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

    1996-09-01T23:59:59.000Z

    The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it 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. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. 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. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions.

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

    SciTech Connect (OSTI)

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

    1991-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Vinson, Dennis

    2010-06-01T23:59:59.000Z

    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.

  12. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    SciTech Connect (OSTI)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)] [Oak Ridge National Lab., TN (United States)

    1994-01-01T23:59:59.000Z

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  13. I. INTRODUCTION The Tokamak Fusion Test Reactor (TFTR) is a

    E-Print Network [OSTI]

    research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D-site complex as a non-nuclear facility as defined in DOE Order 420.1 (Ă?Facility SafetyĂ?) and e) provide data The Decommissioning and Decontamination of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics

  14. I. INTRODUCTION The Tokamak Fusion Test Reactor (TFTR) is a

    E-Print Network [OSTI]

    research reactor that ceased operation in April 1997. The Decontamination and Decommissioning (D­site complex as a non­nuclear facility as defined in DOE Order 420.1 (``Facility Safety'') and e) provide data The Decommissioning and Decontamination of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics

  15. Testing of SP-100 reactor control approaches in the NAT

    SciTech Connect (OSTI)

    Rhow, S.K.; Meyer, R.A.; Wong, K.K.; Halfen, F.J. (General Electric Company, SP-100 Programs, Astro Space Division, San Jose, California 95153-5354 (US))

    1991-01-01T23:59:59.000Z

    The Generic Flight System (GFS) reactor control approach will be tested at the SP-100 Ground Engineering System (GES) test site as part of the Nuclear Assembly Test (NAT) program. A control scheme for the NAT has been developed to emulate the GFS conditions for testing of the GFS reactor control approach in the NAT. Comparisons between the GFS simulation results and the NAT simulation results show that a reasonably good emulation of the GFS conditions can be achieved in the NAT using the control scheme described in this paper.

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

    SciTech Connect (OSTI)

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

    1992-07-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1992-07-01T23:59:59.000Z

    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.

  18. 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.

    1984-06-01T23:59:59.000Z

    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.

  19. Technology Options for a Fast Spectrum Test Reactor

    SciTech Connect (OSTI)

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

    2006-06-01T23:59:59.000Z

    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.

  20. Reduced enrichment for research and test reactors: Proceedings

    SciTech Connect (OSTI)

    Not Available

    1993-07-01T23:59:59.000Z

    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.

  1. advanced test reactor: Topics by E-print Network

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

    test reactor First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Advancing Toward Test Automation through...

  2. aircraft shield test reactor: Topics by E-print Network

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

    aircraft shield test reactor First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Experimental Tests of...

  3. NEUTRON ACTIVATION COOLDOWN OF THE TOKAMAK FUSION TEST REACTOR

    E-Print Network [OSTI]

    involved the safe handling and processing about 100g of tritium. This resulted in manageable long concrete Test Cell showing the relative locations of the vessel, neutral beam injection systems, the vacuum. INTRODUCTION The Tokamak Fusion Test Reactor (TFTR) began high power deuterium­tritium (D­T) fueled operations

  4. 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

    2012-09-01T23:59:59.000Z

    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.

  5. Preliminary design studies on the Broad Application Test Reactor

    SciTech Connect (OSTI)

    Terry, W.J. [ed.; Terry, W.K.; Ryskamp, J.M.; Jahshan, S.N.; Fletcher, C.D.; Moore, R.L.; Leyse, C.F.; Ottewitte, E.H.; Motloch, C.G.; Lacy, J.M.

    1992-08-01T23:59:59.000Z

    This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented.

  6. Reduced-Enrichment Research and Test Reactor Program: Environmental assessment

    SciTech Connect (OSTI)

    Not Available

    1980-05-01T23:59:59.000Z

    The principal program objective and principal part of the proposed action is to improve the proliferation resistance of nuclear fuels used in research and test reactors by providing the technical means (through technical development, design, and testing) for reducing the uranium enrichment requirements of these fuels to substantially less than the 90 to 93% enrichment currently used. Operator acceptance of the reduced-enrichment-uranium (REU) fuel alternative will require minimizing of reactor performance reduction, fuel cycle cost increases, the number of new safety and licensing issues raised, and reactor and facility modifications. The other part of the proposed action is to assure the capability for commercial production and supply of REU fuel for use both in the US and abroad. The RERTR Program scope is limited to generic design studies, technical support to reactor operating organizations in preparing for conversions to REU fuels, fuel development, fuel demonstrations, and technical support for commercialization of REU fuels. This environmental assessment addresses the environmental consequences of RERTR Program activities and of specific conversions of typical reactors (the Ford Nuclear Reactor and one or two other to-be-designated demonstrations) to REU-fuel cycles, including domestic and international shipments of enriched uranium pertinent to the conduct of RERTR Program activities.

  7. Design of the Advanced Gas Reactor Fuel Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2005-10-01T23:59:59.000Z

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight particle fuel tests in the Advanced Test Reactor (ATR) located at the newly formed Idaho National Laboratory (INL) to support development of the next generation Very High Temperature Reactor (VHTR) in the United States. The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments will be irradiated in an inert sweep gas atmosphere with on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The final design phase has just been completed on the first experiment (AGR-1) in this series and the support systems and fission product monitoring system that will monitor and control the experiment during irradiation. This paper discusses the development of the experimental hardware and support system designs and the status of the experiment.

  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)

    2013-07-01T23:59:59.000Z

    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. Advanced Test Reactor probabilistic risk assessment methodology and results summary

    SciTech Connect (OSTI)

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

    1992-01-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 report documents a comprehensive and state-of-the-art study to establish and reduce the risk associated with operation of the ATR, expressed as a mean frequency of fuel damage. The ATR Level 1 PRA effort is unique and outstanding because of its consistent and state-of-the-art treatment of all facets of the risk study, its comprehensive and cost-effective risk reduction effort while the risk baseline was being established, and its thorough and comprehensive documentation. The PRA includes many improvements to the state-of-the-art, including the following: establishment of a comprehensive generic data base for component failures, treatment of initiating event frequencies given significant plant improvements in recent years, performance of efficient identification and screening of fire and flood events using code-assisted vital area analysis, identification and treatment of significant seismic-fire-flood-wind interactions, and modeling of large loss-of-coolant accidents (LOCAs) and experiment loop ruptures leading to direct damage of the ATR core. 18 refs.

  10. Heavy Water Components Test Reactor Decommissioning - Major Component Removal

    SciTech Connect (OSTI)

    Austin, W.; Brinkley, D.

    2010-05-05T23:59:59.000Z

    The Heavy Water Components Test Reactor (HWCTR) facility (Figure 1) was built in 1961, operated from 1962 to 1964, and is located in the northwest quadrant of the Savannah River Site (SRS) approximately three miles from the site boundary. The HWCTR facility is on high, well-drained ground, about 30 meters above the water table. The HWCTR was a pressurized heavy water test reactor used to develop candidate fuel designs for heavy water power reactors. It was not a defense-related facility like the materials production reactors at SRS. The reactor was moderated with heavy water and was rated at 50 megawatts thermal power. In December of 1964, operations were terminated and the facility was placed in a standby condition as a result of the decision by the U.S. Atomic Energy Commission to redirect research and development work on heavy water power reactors to reactors cooled with organic materials. For about one year, site personnel maintained the facility in a standby status, and then retired the reactor in place. In 1965, fuel assemblies were removed, systems that contained heavy water were drained, fluid piping systems were drained, deenergized and disconnected and the spent fuel basin was drained and dried. The doors of the reactor facility were shut and it wasn't until 10 years later that decommissioning plans were considered and ultimately postponed due to budget constraints. In the early 1990s, DOE began planning to decommission HWCTR again. Yet, in the face of new budget constraints, DOE deferred dismantlement and placed HWCTR in an extended surveillance and maintenance mode. The doors of the reactor facility were welded shut to protect workers and discourage intruders. The $1.6 billion allocation from the American Recovery and Reinvestment Act to SRS for site clean up at SRS has opened the doors to the HWCTR again - this time for final decommissioning. During the lifetime of HWCTR, 36 different fuel assemblies were tested in the facility. Ten of these experienced cladding failures as operational capabilities of the different designs were being established. In addition, numerous spills of heavy water occurred within the facility. Currently, radiation and radioactive contamination levels are low within HWCTR with most of the radioactivity contained within the reactor vessel. There are no known insults to the environment, however with the increasing deterioration of the facility, the possibility exists that contamination could spread outside the facility if it is not decommissioned. An interior panoramic view of the ground floor elevation taken in August 2009 is shown in Figure 2. The foreground shows the transfer coffin followed by the reactor vessel and control rod drive platform in the center. Behind the reactor vessel is the fuel pool. Above the ground level are the polar crane and the emergency deluge tank at the top of the dome. Note the considerable rust and degradation of the components and the interior of the containment building. Alternative studies have concluded that the most environmentally safe, cost effective option for final decommissioning is to remove the reactor vessel, steam generators, and all equipment above grade including the dome. Characterization studies along with transport models have concluded that the remaining below grade equipment that is left in place including the transfer coffin will not contribute any significant contamination to the environment in the future. The below grade space will be grouted in place. A concrete cover will be placed over the remaining footprint and the groundwater will be monitored for an indefinite period to ensure compliance with environmental regulations. The schedule for completion of decommissioning is late FY2011. This paper describes the concepts planned in order to remove the major components including the dome, the reactor vessel (RV), the two steam generators (SG), and relocating the transfer coffin (TC).

  11. advanced test idaho reactor: Topics by E-print Network

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

    test idaho reactor First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 DISPERSAL AND HARVEST OF SAGE GROUSE...

  12. 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.

    1982-03-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Gaubatz, D.C.

    1996-12-17T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Gaubatz, Donald C. (Cupertino, CA)

    1996-01-01T23:59:59.000Z

    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.

  15. Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

    SciTech Connect (OSTI)

    Not Available

    1993-07-01T23:59:59.000Z

    The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

  16. 100 Area soil washing treatability test plan

    SciTech Connect (OSTI)

    Not Available

    1993-03-01T23:59:59.000Z

    This test plan describes specifications, responsibilities, and general methodology for conducting a soil washing treatability study as applied to source unit contamination in the 100 Area. The objective ofthis 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. The purpose of separating these fractions is to minimize the volume of soil requiring permanent disposal. It is anticipated that this treatability study will be performed in two phases of testing, a remedy screening phase and a remedy selection phase. The remedy screening phase consists of laboratory- and bench-scale studies performed by Battelle Pacific Northwest laboratories (PNL) under a work order issued by Westinghouse Hanford Company (Westinghouse Hanford). This phase will be used to provide qualitative evaluation of the potential effectiveness of the soil washing technology. The remedy selection phase, consists of pilot-scale testing performed under a separate service contract to be competitively bid under Westinghouse Hanford direction. The remedy selection phase will provide data to support evaluation of the soil washing technology in future feasibility studies for Interim Remedial Measures (IRMs) or final operable unit (OU) remedies. Performance data from these tests will indicate whether applicable or relevant and appropriate requirements (ARARs) or cleanup goals can be met at the site(s) by application of soil washing. The remedy selection tests wig also allow estimation of costs associated with implementation to the accuracy required for the Feasibility Study.

  17. Results of Aquifer Tests Performed Near R-Area, Savannah River Site

    SciTech Connect (OSTI)

    Hiergesell, R.A.

    2001-01-31T23:59:59.000Z

    The aquifer testing described in this report was conducted in response to USEPA comments (WSRC, 1998) on the Rev. 0 R-Reactor Seepage Basins RFI/RI Report (WSRC, 1998a), Appendix G, Groundwater Contaminant Transport Modeling for the R-Reactor Seepage Basins (RRSB)/108-4R Overflow Basin Operable Unit. The R-area regional flow model described in Appendix G of the RFI/RI is based on small-scale and/or indirect measures of hydraulic conductivity, including laboratory tests, slug tests, cone penetration testing (CPT) and lithologic core descriptions. The USEPA proposed and SRS- agreed that large-scale conductivity estimates from multiple well pumping tests would be beneficial for validating the model conductivity field. Overall, the aquifer test results validate the 1998 R-area regional groundwater flow model.

  18. Abnormal operating procedures for ATR (Advanced Test Reactor's) experiment loops

    SciTech Connect (OSTI)

    Auflick, J.L.

    1989-09-01T23:59:59.000Z

    This paper outlines the background from the TMI accident which resulted in the definition and development of function-oriented procedures. It also explains how function-oriented procedures were applied in a task for the Advanced Test Reactor's (ATR) NR experiment loops. Human performance design discrepancies were identified for existing procedures, and were corrected by upgrading them according to current NRC and DOE standards. Finally, specific recommendations are made with respect to future ATR control room and loop improvements, as they relate to the revision of operating procedures within INEL's power reactor program. 8 refs., 4 figs.

  19. Human factors evaluation of the engineering test reactor control room

    SciTech Connect (OSTI)

    Banks, W.W.; Boone, M.P.

    1981-03-01T23:59:59.000Z

    The Reactor and Process Control Rooms at the Engineering Test Reactor were evaluated by a team of human factors engineers using available human factors design criteria. During the evaluation, ETR, equipment and facilities were compared with MIL-STD-1472-B, Human Engineering design Criteria for Military Systems. The focus of recommendations centered on: (a) displays and controls; placing displays and controls in functional groups; (b) establishing a consistent color coding (in compliance with a standard if possible); (c) systematizing annunciator alarms and reducing their number; (d) organizing equipment in functional groups; and (e) modifying labeling and lines of demarcation.

  20. Study of fueling requirements for the Engineering Test Reactor

    SciTech Connect (OSTI)

    Ho, S.K.; Perkins, L.J.

    1987-10-16T23:59:59.000Z

    An assessment of the fueling requirement for the TIBER Engineering Test Reactor is studied. The neutral shielding pellet ablation model with the inclusion of the effects of the alpha particles is used for our study. The high electron temperature in a reactor-grade plasma makes pellet penetration very difficult. The launch length has to be very large (several tens of meters) in order to avoid pellet breakage due to the low inertial strength of DT ''ice.'' The minimum repetition rate corresponding to the largest allowable pellet, is found to be about 1 Hz. A brief survey is done on the various operational and conceptual pellet injection schemes for plasma fueling. The underlying conclusion is that an alternative fueling scheme of coaxial compact-toroid plasma gun is very likely needed for effective central fueling of reactor-grade plasmas. 16 refs.

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

    SciTech Connect (OSTI)

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

    2011-06-01T23:59:59.000Z

    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.

  2. Enhanced in-pile instrumentation at the advanced test reactor

    SciTech Connect (OSTI)

    Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2011-07-01T23:59:59.000Z

    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 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 realtime 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. (authors)

  3. 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

    2012-08-01T23:59:59.000Z

    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.

  4. Draft environmental assessment -- Test Area North pool stabilization project update

    SciTech Connect (OSTI)

    NONE

    1997-06-01T23:59:59.000Z

    The purpose of this Environmental Assessment (EA) is to update the ``Test Area North Pool Stabilization Project`` EA (DOE/EA-1050) and finding of no significant impact (FONSI) issued May 6, 1996. This update analyzes the environmental and health impacts of a drying process for the Three Mile Island (TMI) nuclear reactor core debris canisters now stored underwater in a facility on the Idaho National Engineering and Environmental Laboratory (INEEL). A drying process was analyzed in the predecision versions of the EA released in 1995 but that particular process was determined to be ineffective and dropped form the Ea/FONSI issued May 6, 1996. The origin and nature of the TMI core debris and the proposed drying process are described and analyzed in detail in this EA. As did the 1996 EA, this update analyzes the environmental and health impacts of removing various radioactive materials from underwater storage, dewatering these materials, constructing a new interim dry storage facility, and transporting and placing the materials into the new facility. Also, as did the 1996 EA, this EA analyzes the removal, treatment and disposal of water from the pool, and placement of the facility into a safe, standby condition. The entire action would take place within the boundaries of the INEEL. The materials are currently stored underwater in the Test Area North (TAN) building 607 pool, the new interim dry storage facility would be constructed at the Idaho Chemical Processing Plant (ICPP) which is about 25 miles south of TAN.

  5. Advanced Test Reactor National Scientific User Facility Partnerships

    SciTech Connect (OSTI)

    Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

    2012-03-01T23:59:59.000Z

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin-Madison; (8) Illinois Institute of Technology (IIT) Materials Research Collaborative Access Team (MRCAT) beamline at Argonne National Laboratory's Advanced Photon Source; and (9) Nanoindenter in the University of California at Berkeley (UCB) Nuclear Engineering laboratory Materials have been analyzed for ATR NSUF users at the Advanced Photon Source at the MRCAT beam, the NIST Center for Neutron Research in Gaithersburg, MD, the Los Alamos Neutron Science Center, and the SHaRE user facility at Oak Ridge National Laboratory (ORNL). Additionally, ORNL has been accepted as a partner facility to enable ATR NSUF users to access the facilities at the High Flux Isotope Reactor and related facilities.

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

    SciTech Connect (OSTI)

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

    2011-09-01T23:59:59.000Z

    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).

  7. Fracture toughness test results of thermal aged reactor vessel materials

    SciTech Connect (OSTI)

    DeVan, M.J.; Lowe, A.L. Jr. [B and W Nuclear Technologies Inc., Lynchburg, VA (United States). Nuclear Engineering Dept.; Hall, J.B. [Babcock and Wilcox Co., Alliance, OH (United States)

    1996-12-31T23:59:59.000Z

    Thermal-aged surveillance materials consisting of Sa-533, Grade B, Class 1 plate material; SA-508, Class 2 forging material; and 2 Mn-Mo-Ni/Linde 80 weld metals were removed from two commercial reactor pressure vessels. The material from the first reactor vessel received a thermal exposure of approximately 103,000 hours at 282 C, while the material from the second reactor vessel received a thermal exposure of approximately 93,000 hours at 282 C. Tensile and 1/2 T compact fracture toughness specimens were fabricated from these materials and tested. In addition, to examine the effects of annealing, selected thermal-aged and unaged specimens were annealed at 454 C (850 F) and tested. Varying responses in the fracture toughness properties were observed for all materials after exposure to the thermal-aging temperature. The base metal plate had an observed decrease in J-values after its respective aging exposure, while no significant difference in the J-values were observed for the Linde 80 weld metals. No significant difference was seen in the J-data for the aged/annealed materials, but because of the small number of test specimens available, no conclusion could be determined for the response to annealing.

  8. 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

    2005-02-01T23:59:59.000Z

    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.

  9. 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

    2008-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2006-09-15T23:59:59.000Z

    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.

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

    Open Energy Info (EERE)

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

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

    Open Energy Info (EERE)

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

  13. Advanced Test Reactor National Scientific User Facility Progress

    SciTech Connect (OSTI)

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to 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, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives cannot be met using the INL facilities. The ATR NSUF program includes a robust education program enabling students to participate in their research at INL and the partner facilities, attend the ATR NSUF annual User Week, and compete for prizes at sponsored conferences. Development of additional research capabilities is also a key component of the ATR NSUF Program; researchers are encouraged to propose research projects leading to these enhanced capabilities. Some ATR irradiation experiment projects irradiate more specimens than are tested, resulting in irradiated materials available for post irradiation examination by other researchers. These “extra” specimens comprise the ATR NSUF Sample Library. This presentation will highlight the ATR NSUF Sample Library and the process open to researchers who want to access these materials and how to propose research projects using them. This presentation will provide the current status of all the ATR NSUF Program elements. Many of these were not envisioned in 2007, when DOE established the ATR NSUF.

  14. Electron beams in material testing for fusion reactors

    SciTech Connect (OSTI)

    Ganley, J.T.; McDonald, J.M.; Youchison, D.L. [Sandia National Laboratories, Albuquerque, NM (United States)

    1995-12-31T23:59:59.000Z

    The Plasma Material Test Facility at Sandia National Laboratory is currently involved in a multi-year program of thermally testing materials and high heat flux components for eventual use in fusion reactors. The major test system in the facility is the EB-1200 System, which uses two EH600S Von Ardenne electron guns to provide up to 1.2 MW of heating for large material samples and components. In this paper, the advantages of using electron guns for this type of testing will be discussed. Gun performance characteristics will be presented, as well as electronic techniques used to provide the rapidly varying power levels and complex scan patterns required for complicated targets. Diagnostic capabilities of importance to the experimental program will also be discussed.

  15. Replacement of the Advanced Test Reactor control room

    SciTech Connect (OSTI)

    Durney, J.L.; Klingler, W.B. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1989-01-01T23:59:59.000Z

    The control room for the Advanced Test Reactor has been replaced to provide modern equipment utilizing current standards and meeting the current human factors requirements. The control room was designed in the early 1960 era and had not been significantly upgraded since the initial installation. The replacement did not change any of the safety circuits or equipment but did result in replacement of some of the recorders that display information from the safety systems. The replacement was completed in concert with the replacement of the control room simulator which provided important feedback on the design. The design successfully incorporates computer-based systems into the display of the plant variables. This improved design provides the operator with more information in a more usable form than was provided by the original design. The replacement was successfully completed within the scheduled time thereby minimizing the down time for the reactor. 1 fig., 1 tab.

  16. Test Area North Pool Stabilization Project: Environmental assessment

    SciTech Connect (OSTI)

    NONE

    1996-05-01T23:59:59.000Z

    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.

  17. 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

    2014-04-01T23:59:59.000Z

    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.

  18. Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

    SciTech Connect (OSTI)

    A. B. Culp

    2007-01-26T23:59:59.000Z

    This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

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

    E-Print Network [OSTI]

    Rodriguez, Judy N

    2013-01-01T23:59:59.000Z

    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 ...

  20. Design of a Gas Test Loop Facility for the Advanced Test Reactor

    SciTech Connect (OSTI)

    C. A. Wemple

    2005-09-01T23:59:59.000Z

    The Office of Nuclear Energy within the U.S. Department of Energy (DOE-NE) has identified the need for irradiation testing of nuclear fuels and materials, primarily in support of the Generation IV (Gen-IV) and Advanced Fuel Cycle Initiative (AFCI) programs. These fuel development programs require a unique environment to test and qualify potential reactor fuel forms. This environment should combine a high fast neutron flux with a hard neutron spectrum and high irradiation temperature. An effort is presently underway at the Idaho National Laboratory (INL) to modify a large flux trap in the Advanced Test Reactor (ATR) to accommodate such a test facility [1,2]. The Gas Test Loop (GTL) Project Conceptual Design was initiated to determine basic feasibility of designing, constructing, and installing in a host irradiation facility, an experimental vehicle that can replicate with reasonable fidelity the fast-flux test environment needed for fuels and materials irradiation testing for advanced reactor concepts. Such a capability will be needed if programs such as the AFCI, Gen-IV, the Next Generation Nuclear Plant (NGNP), and space nuclear propulsion are to meet development objectives and schedules. These programs are beginning some irradiations now, but many call for fast flux testing within this decade.

  1. 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.

    1998-12-14T23:59:59.000Z

    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.

  2. CALMOS: Innovative device for the measurement of nuclear heating in material testing reactors

    SciTech Connect (OSTI)

    Carcreff, H. [Alternative Energies and Atomic Energy Commission CEA, Saclay Center, DEN/DANS/DRSN/SIREN, Gif Sur Yvette, 91191 (France)

    2011-07-01T23:59:59.000Z

    An R and D program has been carried out since 2002 in order to improve gamma heating measurements in the 70 MWth OSIRIS Material Testing Reactor operated by CEA's Nuclear Energy Div. at the Saclay research center. Throughout this program an innovative calorimetric probe associated to a specific handling system has been designed in order to make measurements both along the fissile height and on the upper part of the core, where nuclear heating rates still remain high. Two mock-ups of the probe were manufactured and tested in 2005 and 2009 in ex-core area of OSIRIS reactor for the process validation, while a displacement system has been especially designed to move the probe axially. A final probe has been designed thanks to modeling results and to preliminary measurements obtained with mock-ups irradiated to a heating level of 2W/g, This paper gives an overview of the development, describes the calorimetric probe, and expected advantages such as the possibility to use complementary methods to get the nuclear heating measurement. Results obtained with mock-ups irradiated in ex-core area of the reactor are presented and discussed. (authors)

  3. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect (OSTI)

    Sharp, G.L.; McCracken, R.T.

    2003-05-13T23:59:59.000Z

    The regulatory requirement to develop an upgraded safety basis for a DOE Nuclear Facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830). Subpart B of 10 CFR 830, ''Safety Basis Requirements,'' requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements. 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, ''Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants'' as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  4. Reactor Accident Analysis Methodology for the Advanced Test Reactor Critical Facility Documented Safety Analysis Upgrade

    SciTech Connect (OSTI)

    Gregg L. Sharp; R. T. McCracken

    2003-06-01T23:59:59.000Z

    The regulatory requirement to develop an upgraded safety basis for a DOE nuclear facility was realized in January 2001 by issuance of a revision to Title 10 of the Code of Federal Regulations Section 830 (10 CFR 830).1 Subpart B of 10 CFR 830, “Safety Basis Requirements,” requires a contractor responsible for a DOE Hazard Category 1, 2, or 3 nuclear facility to either submit by April 9, 2001 the existing safety basis which already meets the requirements of Subpart B, or to submit by April 10, 2003 an upgraded facility safety basis that meets the revised requirements.1 10 CFR 830 identifies Nuclear Regulatory Commission (NRC) Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants”2 as a safe harbor methodology for preparation of a DOE reactor documented safety analysis (DSA). The regulation also allows for use of a graded approach. This report presents the methodology that was developed for preparing the reactor accident analysis portion of the Advanced Test Reactor Critical Facility (ATRC) upgraded DSA. The methodology was approved by DOE for developing the ATRC safety basis as an appropriate application of a graded approach to the requirements of 10 CFR 830.

  5. Plan for decommissioning the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    Spampinato, P.T.; Walton, G.R. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Commander, J.C. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1993-12-31T23:59:59.000Z

    The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D&D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D&D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D&D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D&D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates.

  6. Plan for decommissioning the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    Spampinato, P.T.; Walton, G.R. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Commander, J.C. (EG and G Idaho, Inc., Idaho Falls, ID (United States))

    1993-01-01T23:59:59.000Z

    The Tokamak Fusion Test Reactor (TFTR) Project is in the planning phase of developing a decommissioning project. A Preliminary Decontamination and Decommissioning (D D) Plan has been developed which provides a framework for the baseline approach, and the cost and schedule estimates. TFTR will become activated and contaminated with tritium after completion of the deuterium-tritium (D-T) experiments. Hence some of the D D operations will require remote handling. It is expected that all of the waste generated will be low level radioactive waste (LLW). The objective of the D D Project is to make TFTR Test Cell available for use by a new fusion experiment. This paper discusses the D D objectives, the facility to be decommissioned, estimates of activation, the technical (baseline) approach, and the assumptions used to develop cost and schedule estimates.

  7. R-Area Reactor 1993 annual groundwater monitoring report

    SciTech Connect (OSTI)

    Not Available

    1994-09-01T23:59:59.000Z

    Groundwater was sampled and analyzed during 1993 from wells monitoring the following locations in R Area: Well cluster P20 east of R Area (one well each in the water table and the McBean formation), the R-Area Acid/Caustic Basin (the four water-table wells of the RAC series), the R-Area Ash Basin/Coal Pile (one well of the RCP series in the Congaree formation and one in the water table), the R-Area Disassembly Basin (the three water-table wells of the RDB series), the R-Area Burning/Rubble Pits (the four water-table wells of the RRP series), and the R-Area Seepage Basins (numerous water-table wells in the RSA, RSB, RSC, RSD, RSE, and RSF series). Lead was the only constituent detected above its 50{mu}g/L standard in any but the seepage basin wells; it exceeded that level in one B well and in 23 of the seepage basin wells. Cadmium exceeded its drinking water standard (DWS) in 30 of the seepage basin wells, as did mercury in 10. Nitrate-nitrite was above DWS once each in two seepage basin wells. Tritium was above DWS in six seepage basin wells, as was gross alpha activity in 22. Nonvolatile beta exceeded its screening standard in 29 wells. Extensive radionuclide analyses were requested during 1993 for the RCP series and most of the seepage basin wells. Strontium-90 in eight wells was the only specific radionuclide other than tritium detected above DWS; it appeared about one-half of the nonvolatile beta activity in those wells.

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

    Open Energy Info (EERE)

    Activity Details Location Blue Mountain Geothermal Area Exploration Technique Flow Test Activity Date 2002 - 2002 Usefulness not useful DOE-funding Unknown Exploration Basis...

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

    Open Energy Info (EERE)

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

  10. Flow Test At Valles Caldera - Sulphur Springs Geothermal Area...

    Open Energy Info (EERE)

    Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Valles Caldera - Sulphur Springs Geothermal Area (Musgrave, Et Al., 1989)...

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

    Open Energy Info (EERE)

    Details Location Long Valley Caldera Geothermal Area Exploration Technique Injectivity Test Activity Date 1999 - 1999 Usefulness not useful DOE-funding Unknown Notes A second...

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

    Open Energy Info (EERE)

    GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Aerial Photography At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Exploration Activity Details Location...

  13. Final Site-Specific Decommissioning Inspection Report for the University of Washington Research and Test Reactor

    SciTech Connect (OSTI)

    Sarah Roberts

    2006-10-18T23:59:59.000Z

    Report of site-specific decommissioning in-process inspection activities at the University of Washington Research and Test Reactor Facility.

  14. A review of experiments and results from the transient reactor test (TREAT) facility.

    SciTech Connect (OSTI)

    Deitrich, L. W.

    1998-07-28T23:59:59.000Z

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

  15. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    SciTech Connect (OSTI)

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22T23:59:59.000Z

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

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

    SciTech Connect (OSTI)

    Pierce, B.F.

    1986-07-01T23:59:59.000Z

    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.

  17. INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS BROOKHAVEN NATIONAL LABORATORY UPTON, NEW YORK

    SciTech Connect (OSTI)

    P.C. Weaver

    2010-12-15T23:59:59.000Z

    5098-SR-03-0 FINAL REPORT- INDEPENDENT VERIFICATION SURVEY OF THE HIGH FLUX BEAM REACTOR DECOMMISSIONING PROJECT OUTSIDE AREAS, BROOKHAVEN NATIONAL LABORATORY

  18. Testing of an advanced thermochemical conversion reactor system

    SciTech Connect (OSTI)

    Not Available

    1990-01-01T23:59:59.000Z

    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.

  19. 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

    2012-07-01T23:59:59.000Z

    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.

  20. 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

  1. INITIAL IRRADIATION OF THE FIRST ADVANCED GAS REACTOR FUEL DEVELOPMENT AND QUALIFICATION EXPERIMENT IN THE ADVANCED TEST REACTOR

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2007-09-01T23:59:59.000Z

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate 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 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 AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  2. area tonopah test: Topics by E-print Network

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

    area tonopah test First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Testing Surface Area Pravesh Kothari...

  3. Core design studies for advanced burner test reactor.

    SciTech Connect (OSTI)

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

    2008-01-01T23:59:59.000Z

    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

  4. Capabilities and Facilities Available at the Advanced Test Reactor to Support Development of the Next Generation Reactors

    SciTech Connect (OSTI)

    S. Blaine Grover; Raymond V. Furstenau

    2005-10-01T23:59:59.000Z

    The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. It is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The Irradiation Test Vehicle (ITV) installed in 1999 enhanced these capabilities by providing a built in experiment monitoring and control system for instrumented and/or temperature controlled experiments. This built in control system significantly reduces the cost for an actively monitored/temperature controlled experiments by providing the thermocouple connections, temperature control system, and temperature control gas supply and exhaust systems already in place at the irradiation position. Although the ITV in-core hardware was removed from the ATR during the last core replacement completed in early 2005, it (or a similar facility) could be re-installed for an irradiation program when the need arises. The proposed Gas Test Loop currently being designed for installation in the ATR will provide additional capability for testing of not only gas reactor materials and fuels but will also include enhanced fast flux rates for testing of materials and fuels for other next generation reactors including preliminary testing for fast reactor fuels and materials. This paper discusses the different irradiation capabilities available and the cost benefit issues related to each capability.

  5. Deterministic Modeling of the High Temperature Test Reactor

    SciTech Connect (OSTI)

    Ortensi, J.; Cogliati, J. J.; Pope, M. A.; Ferrer, R. M.; Ougouag, A. M.

    2010-06-01T23:59:59.000Z

    Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine INL’s current prismatic reactor deterministic analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 column thin annular core, and the fully loaded core critical condition with 30 columns. Special emphasis is devoted to the annular core modeling, which shares more characteristics with the NGNP base design. The DRAGON code is used in this study because it offers significant ease and versatility in modeling prismatic designs. Despite some geometric limitations, the code performs quite well compared to other lattice physics codes. DRAGON can generate transport solutions via collision probability (CP), method of characteristics (MOC), and discrete ordinates (Sn). A fine group cross section library based on the SHEM 281 energy structure is used in the DRAGON calculations. HEXPEDITE is the hexagonal z full core solver used in this study and is based on the Green’s Function solution of the transverse integrated equations. In addition, two Monte Carlo (MC) based codes, MCNP5 and PSG2/SERPENT, provide benchmarking capability for the DRAGON and the nodal diffusion solver codes. The results from this study show a consistent bias of 2–3% for the core multiplication factor. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.

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

    SciTech Connect (OSTI)

    Bradley K. Heath

    2014-03-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Smith, R.J.

    1995-09-11T23:59:59.000Z

    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

  8. aerospace system test reactor: Topics by E-print Network

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

    methods of nulcear fusion: inertial confinement fusion (ICF) and magnetic confinement fusion (MCF). Existing thermonuclear reactors are very complex, expensive, large, and...

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

    SciTech Connect (OSTI)

    Carlson, Bryan J [Los Alamos National Laboratory; Trujillo, Stephen [Los Alamos National Laboratory; Willms, R. Scott [Los Alamos National Laboratory

    2010-01-01T23:59:59.000Z

    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.

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

    Office of Environmental Management (EM)

    December 12, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation Nevada National Security Site Underground Test Area...

  11. Regional groundwater flow model for C, K. L. and P reactor areas, Savannah River Site, Aiken, SC

    SciTech Connect (OSTI)

    Flach, G.P.

    2000-02-11T23:59:59.000Z

    A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K, L, and P reactor areas has been developed. The reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department. The model provides a quantitative understanding of groundwater flow on a regional scale within the near surface aquifers and deeper semi-confined to confined aquifers. The model incorporates historical and current field characterization data up through Spring 1999. Model preprocessing is automated so that future updates and modifications can be performed quickly and efficiently. The CKLP regional reactor model can be used to guide characterization, perform scoping analyses of contaminant transport, and serve as a common base for subsequent finer-scale transport and remedial/feasibility models for each reactor area.

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

    SciTech Connect (OSTI)

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

    2013-12-01T23:59:59.000Z

    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.

  13. Completing the Design of the Advanced Gas Reactor Fuel Development and Qualification Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2006-10-01T23:59:59.000Z

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the newly formed 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 ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new 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 AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new 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 six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation.

  14. A New Interpretation of Alpha-Particle-Driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    A New Interpretation of Alpha-Particle-Driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor

  15. area field test: Topics by E-print Network

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

    area field test First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Smart Parking Management Field Test: A...

  16. area nevada test: Topics by E-print Network

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

    area nevada test First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Renewable Energy and the Nevada Test...

  17. areas nevada test: Topics by E-print Network

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

    areas nevada test First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Renewable Energy and the Nevada Test...

  18. area treatability test: Topics by E-print Network

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

    treatability test First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Testing Surface Area Pravesh Kothari...

  19. Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2008-10-01T23:59:59.000Z

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate 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 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. These AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control for each capsule. The swept gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The design of the first experiment (designated AGR-1) was completed in 2005, and the fabrication and assembly of the test train as well as the support systems and fission product monitoring system that monitor and control the experiment during irradiation were completed in September 2006. The experiment was inserted in the ATR in December 2006, and is serving as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed and the status of the experiment is provided.

  20. A risk characterization of safety research areas for Integral Fast Reactor program planning

    SciTech Connect (OSTI)

    Mueller, C.J.; Cahalan, J.E.; Hill, D.J.; Kramer, J.M.; Marchaterre, J.F.; Pedersen, D.R.; Sevy, R.H.; Tilbrook, R.W.; Wei, T.Y.; Wright, A.E.

    1988-01-01T23:59:59.000Z

    This paper characterizes the areas of Integral Fast Reactor (IFR) safety research in terms of their importance in addressing the risk of core disruption sequences for innovative designs. Such sequences have traditionally been determined to constitute the primary risk to public health and safety. All core disruption sequences are folded into four fault categories: classic unprotected (unscrammed) events; loss of decay heat; local fault propagation; and failure of critical reactor structures. Event trees are used to describe these sequences and the areas in the IFR Safety and related Base Technology research programs are discussed with respect to their relevance in addressing the key issues in preventing or delimiting core disruptive sequences. Thus a measure of potential for risk reduction is obtained for guidance in establishing research priorites.

  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

    2014-04-01T23:59:59.000Z

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

    SciTech Connect (OSTI)

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

    2014-08-01T23:59:59.000Z

    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.

  3. 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

    2011-03-30T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Romano, T.

    1997-09-29T23:59:59.000Z

    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.

  5. 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

    2012-02-01T23:59:59.000Z

    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.

  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

    2012-02-01T23:59:59.000Z

    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. Preliminary requirements for a Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR)

    SciTech Connect (OSTI)

    Massie, M.; Forsberg, C.; Forget, B. [Dept. of Nuclear Science and Engineering, Massachusetts Inst. of Technology, Cambridge, MA 02139 (United States); Hu, L. W. [Nuclear Reactor Laboratory, Massachusetts Inst. of Technology, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

    2012-07-01T23:59:59.000Z

    A Fluoride Salt-Cooled High-Temperature Test Reactor (FHTR) design is being developed at MIT to provide the first demonstration and test of a salt-cooled reactor using high-temperature fuel. The first step is to define the requirements. The top level requirements are (1) provide the confidence that a larger demonstration reactor is warranted and (2) develop the necessary data for a larger-scale reactor. Because requirements will drive the design of the FHTR, a significant effort is being undertaken to define requirements and understand the tradeoffs that will be required for a practical design. The preliminary requirements include specifications for design parameters and necessary tests of major reactor systems. Testing requirements include demonstration of components, systems, and procedures for refueling, instrumentation, salt temperature control to avoid coolant freezing, salt chemistry and volume control, tritium monitoring and control, and in-service inspection. Safety tests include thermal hydraulics, neutronics - including intrinsic core shutdown mechanisms such as Doppler feedback - and decay heat removal systems. Materials and coolant testing includes fuels (including mechanical wear and fatigue) and system corrosion behavior. Preliminary analysis indicates a thermal power output below 30 MW, an initial core using pebble-bed or prismatic-block fuel, peak outlet temperatures of at least 700 deg. C, and use of FLi{sup 7}Be ({sup 7}LiF-BeF{sub 2}) coolant. The option to change-out the reactor core, fuel type, and major components is being investigated. While the FHTR will be used for materials testing, its primary mission is as a reactor system performance test to enable the design and licensing of a FHR demonstration power reactor. (authors)

  8. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover; John Maki; David Petti

    2010-10-01T23:59:59.000Z

    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). 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 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 completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during irradiation will be discussed and the results of the experiment will be presented. The second experiment (AGR-2) is currently being assembled, and the status as well as the new fuel and irradiation conditions for that experiment will also be discussed.

  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

    2010-01-01T23:59:59.000Z

    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. Issues and test requirements in radiation shielding of fusion reactors

    SciTech Connect (OSTI)

    Nakagawa, M.; Abdou, M.A.

    1986-11-01T23:59:59.000Z

    Radiation shield issues for fusion reactors have been investigated and the experiments and facilities required to resolve the issues have been identified and characterized as part of the FINESSE program. This paper summarizes the recommended approach to fusion shield research and development, provides a summary of the necessary experiments and facilities, and presents the results of technical analyses involved.

  11. Issues and test requirements in radiation shielding of fusion reactors

    SciTech Connect (OSTI)

    Nakagawa, M.; Abdou, M.A.

    1986-01-01T23:59:59.000Z

    Radiation shield issues for fusion reactors have been investigated and the experiments and facilities required to resolve the issues have been identified and characterized as part of the FINESSE program. This paper summarizes the recommended approach to fusion shield R and D, provides a summary of the necessary experiments and facilities, and presents the results of technical analyses involved.

  12. Non Nuclear Testing of Reactor Systems In The Early Flight Fission Test Facilities (EFF-TF)

    SciTech Connect (OSTI)

    Van Dyke, Melissa; Martin, James [Marshall Space Flight Center, National Aeronautics and Space Administration, Huntsville, Alabama, 35812 (United States)

    2004-07-01T23:59:59.000Z

    The Early Flight Fission-Test Facility (EFF-TF) can assist in the design and development of systems through highly effective non-nuclear testing of nuclear systems when technical issues associated with near-term space fission systems are 'non-nuclear' in nature (e.g. system's nuclear operations are understood). For many systems, thermal simulators can be used to closely mimic fission heat deposition. Axial power profile, radial power profile, and fuel pin thermal conductivity can be matched. In addition to component and subsystem testing, operational and lifetime issues associated with the steady state and transient performance of the integrated reactor module can be investigated. Instrumentation at the EFF-TF allows accurate measurement of temperature, pressure, strain, and bulk core deformation (useful for accurately simulating nuclear behavior). Ongoing research at the EFF-TF is geared towards facilitating research, development, system integration, and system utilization via cooperative efforts with DOE laboratories, industry, universities, and other Nasa centers. This paper describes the current efforts for the latter portion of 2003 and beginning of 2004. (authors)

  13. 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

    2012-05-01T23:59:59.000Z

    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.

  14. Closure Report for Corrective Action Unit 143: Area 25 Contaminated Waste Dumps, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    D. S. Tobiason

    2002-03-01T23:59:59.000Z

    This Closure Report (CR) has been prepared for the Area 25 Contaminated Waste Dumps (CWD), Corrective Action Unit (CAU) 143 in accordance with the Federal Facility Agreement and Consent Order [FFACO] (FFACO, 1996) and the Nevada Division of Environmental Protection (NDEP)-approved Corrective Action Plan (CAP) for CAU 143: Area 25, Contaminated Waste Dumps, Nevada Test Site, Nevada. CAU 143 consists of two Corrective Action Sites (CASs): 25-23-09 CWD No.1, and 25-23-03 CWD No.2. The Area 25 CWDs are historic disposal units within the Area 25 Reactor Maintenance, Assembly, and Disassembly (R-MAD), and Engine Maintenance, Assembly, and Disassembly (E-MAD) compounds located on the Nevada Test Site (NTS). The R-MAD and E-MAD facilities originally supported a portion of the Nuclear Rocket Development Station in Area 25 of the NTS. CWD No.1 CAS 25-23-09 received solid radioactive waste from the R-MAD Compound (East Trestle and West Trench Berms) and 25-23-03 CWD No.2 received solid radioactive waste from the E-MAD Compound (E-MAD Trench).

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

    SciTech Connect (OSTI)

    Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

    2011-10-31T23:59:59.000Z

    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.

  16. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2009-09-01T23:59:59.000Z

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight 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). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new 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 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 ten years to demonstrate and qualify new 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 the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the two experiments will be compared and the irradiation results to date on the first experiment will be presented.

  17. 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))

    1993-01-15T23:59:59.000Z

    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.

  18. Preliminary investigation Area 12 fleet operations steam cleaning discharge area Nevada Test Site

    SciTech Connect (OSTI)

    NONE

    1996-07-01T23:59:59.000Z

    This report documents the characterization activities and findings of a former steam cleaning discharge area at the Nevada Test Site. The former steam cleaning site is located in Area 12 east of Fleet Operations Building 12-16. The characterization project was completed as a required condition of the ``Temporary Water Pollution Control Permit for the Discharge From Fleet Operations Steam Cleaning Facility`` issued by the Nevada Division of Environmental Protection. The project objective was to collect shallow soil samples in eight locations in the former surface discharge area. Based upon field observations, twelve locations were sampled on September 6, 1995 to better define the area of potential impact. Samples were collected from the surface to a depth of approximately 0.3 meters (one foot) below land surface. Discoloration of the surface soil was observed in the area of the discharge pipe and in localized areas in the natural drainage channel. The discoloration appeared to be consistent with the topographically low areas of the site. Hydrocarbon odors were noted in the areas of discoloration only. Samples collected were analyzed for bulk asbestos, Toxicity Characteristic Leaching Procedure (TCLP) metals, total petroleum hydrocarbons (TPHs), volatile organic compounds (VOCs), semi-volatile organic compounds (Semi-VOCs), and gamma scan.

  19. argonne tank research and test reactor-aarr: Topics by E-print...

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

    argonne tank research and test reactor-aarr First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Tests using...

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

    E-Print Network [OSTI]

    involved the safe handling and processing about 100g of tritium. This resulted in manageable long concrete Test Cell showing the relative locations of the vessel, neutral beam injection systems, the vacuum. INTRODUCTION The Tokamak Fusion Test Reactor (TFTR) began high power deuterium-tritium (D-T) fueled operations

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

    E-Print Network [OSTI]

    Yonehara, Katsuya

    2012-01-01T23:59:59.000Z

    In order to develop an RF cavity that is applicable for a muon beam cooling channel, a new facility, called Mucool Test Area (MTA) has been built at Fermilab. MTA is a unique facility whose purpose is to test RF cavities in various conditions. There are 201 and 805 MHz high power sources, a 4-Tesla solenoid magnet, a cryogenic system including a Helium liquifier, an explosion proof apparatus to operate gaseous/liquid Hydrogen, and a beam transport line to send an intense H- beam from the Fermilab Linac accelerator to the MTA hall. Recent activities at MTA will be discussed in this document.

  2. Fabrication, inspection, and test plan for the Advanced Test Reactor (ATR) Mixed-Oxide (MOX) fuel irradiation project

    SciTech Connect (OSTI)

    Wachs, G.W.

    1997-11-01T23:59:59.000Z

    The Department of Energy (DOE) Fissile Materials Disposition Materials Disposition Program (FMDP) has announced that reactor irradiation of MOX fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The MOX fuel test will be irradiated in the ATR to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. In addition, the test will contribute experience with irradiation of gallium-containing fuel to the data base required for resolution of generic CLWR fuel design issues (ORNL/MD/LTR-76). This Fabrication, Inspection, and Test Plan (FITP) is a level 2 document as defined in the FMDP LWR MOX Fuel Irradiation Test Project Plan (ORNL/MD/LTR-78).

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

    SciTech Connect (OSTI)

    Yvonne Townsend

    2000-05-01T23:59:59.000Z

    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.

  4. 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

    2010-03-01T23:59:59.000Z

    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”.

  5. 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

    2006-01-09T23:59:59.000Z

    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.

  6. 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.

    2009-09-30T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Mourao, Rogerio P.; Leite da Silva, Luiz [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil)] [Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte (Brazil); Miranda, Carlos A.; Mattar Neto, Miguel [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil)] [Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo (Brazil); Quintana, Jose F.A.; Saliba, Roberto O. [Comision Nacional de Energia Atomica, Bariloche (Argentina)] [Comision Nacional de Energia Atomica, Bariloche (Argentina); Novara, Oscar E. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)] [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    2013-07-01T23:59:59.000Z

    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)

  8. 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

    2010-08-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2000-02-01T23:59:59.000Z

    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.

  10. Installation and Final Testing of an On-Line, Multi-Spectrometer Fission Product Monitoring System (FPMS) to Support Advanced Gas Reactor (AGR) Fuel Testing and Qualification in the Advanced Test Reactor

    SciTech Connect (OSTI)

    J. K. Hartwell; D. M. Scates; M. W. Drigert; J. B. Walter

    2006-10-01T23:59:59.000Z

    The US Department of Energy (DOE) is initiating tests of reactor fuel for use in an Advanced Gas Reactor (AGR). The AGR will use helium coolant, a low-power-density ceramic core, and coated-particle fuel. A series of eight (8) fuel irradiation tests are planned for the Idaho National Laboratory’s (INL’s) Advanced Test Reactor (ATR). One important measure of fuel performance in these tests is quantification of the fission gas releases over the nominal 2-year duration of each irradiation experiment. This test objective will be met using the AGR Fission Product Monitoring System (FPMS) which includes seven (7) on-line detection stations viewing each of the six test capsule effluent lines (plus one spare). Each station incorporates both a heavily-shielded high-purity germanium (HPGe) gamma-ray spectrometer for quantification of the isotopic releases, and a NaI(Tl) scintillation detector to monitor the total count rate and identify the timing of the releases. The AGR-1 experiment will begin irradiation after October 1, 2006. To support this experiment, the FPMS has been completely assembled, tested, and calibrated in a laboratory at the INL, and then reassembled and tested in its final location in the ATR reactor basement. This paper presents the details of the equipment performance, the control and acquisition software, the test plan for the irradiation monitoring, and the installation in the ATR basement. Preliminary on-line data may be available by the Conference date.

  11. 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.)

    2004-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2011-06-15T23:59:59.000Z

    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).

  13. An overview of FFTF (Fast Flux Test Facility) contributions to Liquid Metal Reactor Safety

    SciTech Connect (OSTI)

    Waltar, A.E.; Padilla, A. Jr.

    1990-11-01T23:59:59.000Z

    The Fast Flux Test Facility has provided a very useful framework for testing the advances in Liquid Metal Reactor Safety Technology. During the licensing phase, the switch from a nonmechanistic bounding technique to the mechanistic approach was developed and implemented. During the operational phase, the consideration of new tests and core configurations led to use of the anticipated-transients-without-scram approach for beyond design basis events and the move towards passive safety. The future role of the Fast Flux Test Facility may involve additional passive safety and waste transmutation tests. 26 refs.

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

    SciTech Connect (OSTI)

    Cliff B. Davis

    2007-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    HALGREN DL

    2010-03-12T23:59:59.000Z

    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.

  16. A review of two recent occurrences at the Advanced Test Reactor involving subcontractor activities

    SciTech Connect (OSTI)

    Dahlke, H.J.; Jensen, N.C.; Vail, J.A.

    1997-11-01T23:59:59.000Z

    This report documents the results of a brief, unofficial investigation into two incidents at the Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR) facility, reported on October 25 and 31, 1997. The first event was an unanticipated breach of confinement. The second involved reactor operation with an inoperable seismic scram subsystem, violating the reactor`s Technical Specifications. These two incidents have been found to be unrelated. A third event that occurred on December 16, 1996, is also discussed because of its similarities to the first event listed above. Both of these incidents were unanticipated breaches of confinement, and both involved the work of construction subcontractor personnel. The cause for the subcontractor related occurrences is a work control process that fails to effectively interface with LMITCO management. ATR Construction Project managers work sufficient close with construction subcontractor personnel to understand planned day-to-day activities. They also have sufficient training and understanding of reactor operations to ensure adherence to applicable administrative requirements. However, they may not be sufficiently involved in the work authorization and control process to bridge an apparent communications gap between subcontractor employees and Facility Operations/functional support personnel for work inside the reactor facility. The cause for the inoperable seismic scram switch (resulting from a disconnected lead) is still under investigation. It does not appear to be subcontractor related.

  17. High poloidal beta long-pulse experiments in the Tokamak Fusion Test Reactor*

    E-Print Network [OSTI]

    Mauel, Michael E.

    High poloidal beta long-pulse experiments in the Tokamak Fusion Test Reactor* J. Kesner+ Plasma stability and confinement. As the current profile evolved, a significantly reduced beta limit was observed after the current ramp-down carried negative current. At later times in lower flN discharges, beta

  18. 2012 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2013-02-01T23:59:59.000Z

    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.

  19. 2013 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2014-02-01T23:59:59.000Z

    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.

  20. 2010 Radiological Monitoring Results Associated with the Advance Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    mike lewis

    2011-02-01T23:59:59.000Z

    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.

  1. 2011 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2012-02-01T23:59:59.000Z

    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.

  2. advanced test reactor critical facility: Topics by E-print Network

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

    test reactor critical facility First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Powerline Conductor...

  3. 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

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

    SciTech Connect (OSTI)

    Lowry, N.J.

    1998-10-21T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Redi, Martha H.

    on the Tokamak Fusion Test Reactor M. H. Redia , S. H. Batha, R. V. Budny, D. S. Darrow, F. M. Levinton, D. C. Mc Test Reactor [Phys. Plas. 5, 1577 (1998)] are found to be in good agreement with measurements includes the neoclassical transport processes, a recent first-principles model for stochastic ripple loss

  6. 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

  7. 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

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

    SciTech Connect (OSTI)

    G. N. Doyle

    2002-02-01T23:59:59.000Z

    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.

  9. Hanford Immobilized LAW Product Acceptance Testing: Tanks Focus Area Results

    SciTech Connect (OSTI)

    Vienna, John D.; Hrma, Pavel R.; Jiricka, Antonin; Smith, Donald E.; Lorier, Troy H.; Reamer, Irene A.; Schulz, Rebecca L.

    2001-12-31T23:59:59.000Z

    Immobilizing low-activity waste (LAW) stored at Hanford site will result in approximately 200 000 m3 of waste glass. It must be demonstrated that this glass can adequately retain radionuclides and prevent contamination of the surrounding environment. A study is being performed to determine the effect of glass composition on its capability to withstand the conditions in the Hanford site burial scenario. To predict the long-term corrosion behavior of waste glass, it is necessary to study the composition and properties of alteration products. The vapor hydration test (VHT) and product consistency test (PCT) were selected as the methods to accelerate the corrosion process and to form alteration products. VHT and PCT was performed on 75 glasses, of which 45 were designed to systematically vary the composition. VHTs were conducted at temperatures ranging from 90?C to 300?C. Alteration rates for most glasses are being determined at 200?C. Selected glasses were tested at different temperatures to determine the effect of temperature on the assemblage of alteration products and the apparent alteration rates. PCTs were performed at a glass surface area to solution volume ratio (S/V) of 2000 m-1 for 7 d and at a S/V of 20 000 m-1 for 10, 100, 1000, 5000, and 10000 h all at 90?C.

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

    SciTech Connect (OSTI)

    Reilly, Raymond W.

    2012-07-30T23:59:59.000Z

    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.

  11. Tests of Lorentz and CPT Violation in the Medium Baseline Reactor Antineutrino Experiment

    E-Print Network [OSTI]

    Yu-Feng Li; Zhen-hua Zhao

    2014-11-21T23:59:59.000Z

    Tests of Lorentz and CPT violation in the medium baseline reactor antineutrino experiment are presented in the framework of the Standard Model Extension (SME). Both the spectral distortion and sidereal variation are employed to derive the limits of Lorentz violation (LV) coefficients. We do the numerical analysis of the sensitivity of LV coefficients by taking the Jiangmen Underground Neutrino Observatory (JUNO) as an illustration, which can improve the sensitivity by more than two orders of magnitude compared with the current limits from reactor antineutrino experiments.

  12. 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.

    2010-12-01T23:59:59.000Z

    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.

  13. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    SciTech Connect (OSTI)

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01T23:59:59.000Z

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production.

  14. 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

    2006-02-01T23:59:59.000Z

    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.

  15. Results of the DF-4 BWR (boiling water reactor) control blade-channel box test

    SciTech Connect (OSTI)

    Gauntt, R.O.; Gasser, R.D.

    1990-10-01T23:59:59.000Z

    The DF-4 in-pile fuel damage experiment investigated the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered reactor accident. This experiment, which was carried out in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories, was performed under the USNRC's internationally sponsored severe fuel damage (SFD) program. The DF-4 test is described herein and results from the experiment are presented. Important findings from the DF-4 test include the low temperature melting of the stainless steel control blade caused by reaction with the B{sub 4}C, and the subsequent low temperature attack of the Zr-4 channel box by the relocating molten blade components. Hydrogen generation was found to continue throughout the experiment, diminishing slightly following the relocation of molten oxidizing zircaloy to the lower extreme of the test bundle. A large blockage which was formed from this material continued to oxidize while steam was being fed into the the test bundle. The results of this test have provided information on the initial stages of core melt progression in BWR geometry involving the heatup and cladding oxidation stages of a severe accident and terminating at the point of melting and relocation of the metallic core components. The information is useful in modeling melt progression in BWR core geometry, and provides engineering insight into the key phenomena controlling these processes. 12 refs., 12 figs.

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

    SciTech Connect (OSTI)

    Not Available

    1993-07-01T23:59:59.000Z

    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.

  17. 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

    2011-10-10T23:59:59.000Z

    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

  18. Cryogenics for the MuCool Test Area (MTA)

    SciTech Connect (OSTI)

    Darve, Christine; Norris, Barry; Pei, Liu-Jin; /Fermilab

    2005-09-01T23:59:59.000Z

    MuCool Test Area (MTA) is a complex of buildings at Fermi National Accelerator Laboratory, which are dedicated to operate components of a cooling cell to be used for Muon Collider and Neutrino Factory R&D. The long-term goal of this facility is to test ionization cooling principles by operating a 25-liter liquid hydrogen (LH{sub 2}) absorber embedded in a 5 Tesla superconducting solenoid magnet. The MTA solenoid magnet will be used with RF cavities exposed to a high intensity beam. Cryogens used at the MTA include LHe, LN{sub 2} and LH{sub 2}. The latter dictates stringent system design for hazardous locations. The cryogenic plant is a modified Tevatron refrigerator based on the Claude cycle. The implementation of an in-house refrigerator system and two 300 kilowatt screw compressors is under development. The helium refrigeration capacity is 500 W at 14 K. In addition the MTA solenoid magnet will be batch-filled with LHe every 2 days using the same cryo-plant. This paper reviews cryogenic systems used to support the Muon Collider and Neutrino Factory R&D programs and emphasizes the feasibility of handling cryogenic equipment at MTA in a safe manner.

  19. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    SciTech Connect (OSTI)

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01T23:59:59.000Z

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed a Readiness Assessment in November 2009. The HSIS is a key component of the ATR National Scientific User Facility (NSUF) operated by Battelle Energy Alliance, LLC and is available to a wide variety of university researchers for nuclear fuels and materials experiments as well as medical isotope research and production.

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

    SciTech Connect (OSTI)

    Ludowise, J.D.

    1994-05-01T23:59:59.000Z

    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.

  1. 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 experiment "Blind" Test Approach · Conduct high quality field and laboratory tests to characterize Geological Survey Turkey Flat, USA Site Effects Test Area: "Blind" Test of Predicted Ground Response

  2. The Advanced Test Reactor Irradiation Capabilities Available as a National Scientific User Facility

    SciTech Connect (OSTI)

    S. Blaine Grover

    2008-09-01T23:59:59.000Z

    The Advanced Test Reactor (ATR) is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These capabilities include simple capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. Monitoring systems have also been utilized to monitor different parameters such as fission gases for fuel experiments, to measure specimen performance during irradiation. ATR’s control system provides a stable axial flux profile throughout each reactor operating cycle, and allows the thermal and fast neutron fluxes to be controlled separately in different sections of the core. The ATR irradiation positions vary in diameter from 16 mm to 127 mm over an active core height of 1.2 m. This paper discusses the different irradiation capabilities with examples of different experiments and the cost/benefit issues related to each capability. The recent designation of ATR as a national scientific user facility will make the ATR much more accessible at very low to no cost for research by universities and possibly commercial entities.

  3. 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

    2013-12-30T23:59:59.000Z

    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.

  4. E-Print Network 3.0 - auxiliary reactor area-1 Sample Search...

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

    plans eyed for fusion project site Summary: said. The six parties involved in the Thermonuclear Experimental Reactor (ITER) project have been... unable to agree where to locate...

  5. The decontamination, decommissioning, and demolition of loss-of-fluid test reactor at the Idaho National Laboratory Site

    SciTech Connect (OSTI)

    Floerke, J.P.; Borschel, Th.F.; Rhodes, L.K. [CH2M-WG Idaho, Idaho Falls, ID (United States)

    2007-07-01T23:59:59.000Z

    In October 2006, CH2M-WG Idaho completed the decontamination, decommissioning and demolition of the Loss-of-Fluid Test (LOFT) facility. The 30-year-old research reactor, located at the Idaho National Laboratory site, posed significant challenges involving regulations governing the demolition of a historical facility, as well as worker safety issues associated with the removal of the reactor's domed structure. The LOFT facility was located at the west end of Test Area North (TAN), built in the 1950's to support the government's aircraft nuclear propulsion program. When President Kennedy cancelled the nuclear propulsion program in 1961, TAN began to host various other activities. The LOFT reactor became part of the new mission. The LOFT facility, constructed between 1965 and 1975, was a scaled-down version of a commercial pressurized water reactor. Its design allowed engineers, scientists, and operators to create or re-create loss-of-fluid accidents (reactor fuel meltdowns) under controlled conditions. The LOFT dome provided containment for a relatively small, mobile test reactor that was moved into and out of the facility on a railroad car. The dome was roughly 21 meters (70 feet) in diameter and 30 meters (98 feet) in height. The Nuclear Regulatory Commission received the results from the accident tests and incorporated the data into commercial reactor operating codes. The facility conducted 38 experiments, including several small loss-of-coolant experiments designed to simulate events such as the accident that occurred at Three Mile Island in Pennsylvania, before the LOFT facility was closed. Through formal survey and research, the LOFT facility was determined to be a DOE Signature Property, as defined by the 'INEEL Cultural Resource Management Plan', and thus eligible for inclusion in the National Register of Historic Places. Decontamination and decommissioning (D and D) of the facility constituted an adverse effect on the historic property that required resolution through the contractor (CH2M-WG Idaho), the U.S. Department of Energy, the Idaho State Historic Preservation Office (SHPO), and the Advisory Council on Historic Preservation. The project team identified multiple hazards that would result if conventional techniques were used to demolish the dome. The physical structure of the vessel containment facility reached 30 meters (98 feet) above grade, presenting significant worker safety hazards created by hoisting and rigging activities. The dome also included a polar crane, 19 meters (62 feet) above grade, that posed similar hazards to workers. The need to work on significantly elevated surfaces, and the thickness of the dome walls - 30 millimeters (1-3/16 inches) of carbon steel - would prove difficult with traditional arc plasma cutting tools. The dome's proximity to operating facilities with equipment sensitive to vibration added to the demolition challenges. To address cultural resource issues, the project team engaged all parties in negotiations and in mapping a path foreword. Open and frequent communication resulted in a Memorandum of Agreement, with stipulations that mitigated the adverse affects of the intended demolition action. The unique mitigating actions resulted in a favorable agreement being signed and issued. To mitigate hazards posed by the height of the facility, the project team had to abandon traditional D and D techniques and employ other methods to complete demolition safely. A different approach and a change in demolition sequence resulted in the safe and efficient removal of the one-of-a-kind containment facility. The approach reduced the use of aerial lifts, aboveground size reduction, and dangerous hoisting and rigging activities that could pose significant hazards to workers. (authors)

  6. 300 AREA URANIUM CONTAMINATION

    SciTech Connect (OSTI)

    BORGHESE JV

    2009-07-02T23:59:59.000Z

    {sm_bullet} Uranium fuel production {sm_bullet} Test reactor and separations experiments {sm_bullet} Animal and radiobiology experiments conducted at the. 331 Laboratory Complex {sm_bullet} .Deactivation, decontamination, decommissioning,. and demolition of 300 Area facilities

  7. Design of irradiation rig for reactor testing of prototype bolometers for ITER

    SciTech Connect (OSTI)

    Gusarov, A.; Huysmans, S. [SCK.CEN Belgian Nucrear Research Center, 2400 Mol (Belgium); Meister, H. [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstr. 2, D-85748 Garching b. Muenchen (Germany); Hodgson, E. [Euratom/CIEMAT Fusion Association, Avenida Complutense 22, 28040 Madrid (Spain)

    2011-07-01T23:59:59.000Z

    We describe the design of an experimental rig, which was developed to allow reactor testing at relevant conditions, i.e. vacuum and {approx}400 deg.C temperature, of prototype resistive bolometers, which will be used in ITER to acquire information on the radiated power distribution from the main plasma and in the diverter region. The main feature of the design is that the rig has no active temperature control. (authors)

  8. 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.

    1982-03-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    NA

    2006-03-01T23:59:59.000Z

    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.

  10. Use and Storage of Test and Operations Data from the High Temperature Test Reactor Acquired by the US Government from the Japan Atomic Energy Agency

    SciTech Connect (OSTI)

    Hans Gougar

    2010-02-01T23:59:59.000Z

    This document describes the use and storage of data from the High Temperature Test Reactor (HTTR) acquired from the Japan Atomic Energy Agency (JAEA) by the U.S. Government for high temperature reactor research under the Next Generation Nuclear Plant (NGNP) Project.

  11. D-area oil seepage basin bioventing optimization test plan

    SciTech Connect (OSTI)

    Berry, C.J.; Radway, J.C.; Alman, D.; Hazen, T.C.

    1998-12-31T23:59:59.000Z

    The D Area Oil Seepage Basin (DOSB) was used from 1952 to 1975 for disposal of petroleum-based products (waste oils), general office and cafeteria waste, and apparently some solvents [trichloroethylene (TCE)/tetrachloroethylene (PCE)]. Numerous analytical results have indicated the presence of TCE and its degradation product vinyl chloride in groundwater in and around the unit, and of petroleum hydrocarbons in soils within the unit. The DOSB is slated for additional assessment and perhaps for environmental remediation. In situ bioremediation represents a technology of demonstrated effectiveness in the reclamation of sites contaminated with petroleum hydrocarbons and chlorinated solvents, and has been retained as an alternative for the cleanup of the DOSB. The Savannah River Site is therefore proposing to conduct a field treatability study designed to demonstrate and optimize the effectiveness of in situ microbiological biodegradative processes at the DOSB. The introduction of air and gaseous nutrients via two horizontal injection wells (bioventing) is expected to enhance biodegradation rates of petroleum components and stimulate microbial degradation of chlorinated solvents. The data gathered in this test will allow a determination of the biodegradation rates of contaminants of concern in the soil and groundwater, allow an evaluation of the feasibility of in situ bioremediation of soil and groundwater at the DOSB, and provide data necessary for the functional design criteria for the final remediation system.

  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

    2012-08-01T23:59:59.000Z

    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. 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

    2008-04-01T23:59:59.000Z

    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.

  14. 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

    2012-09-01T23:59:59.000Z

    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.

  15. Subtask 7.4 - Power River Basin Subbituminous Coal-Biomass Cogasification Testing in a Transport Reactor

    SciTech Connect (OSTI)

    Michael Swanson; Daniel Laudal

    2009-03-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the Kellogg Brown and Root transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 3600 hours of operation on 17 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air- and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 90% have also been obtained and are highly dependent on the oxygen-coal ratio. Higher-reactivity (low-rank) coals appear to perform better in a transport reactor than the less reactive bituminous coals. Factors that affect TRDU product gas quality appear to be coal type, temperature, and oxygen/fuel ratios. During this series of tests, a previously tested baseline Powder River Basin (PRB) subbituminous coal from the Peabody Energy North Antelope Rochelle Mine near Gillette, Wyoming was mixed with 20 wt% biomass. Two types of biomass were used - wood waste and switchgrass. Gas and particulate sampling at several locations in the riser provided information on coal devolatilization and cracking chemistry as a function of residence time, transport gas, and mode of operation. The goal of these tests was to compare the operating data and sample chemistry of the coal-biomass mixture to the PRB coal, with a focus on Fischer-Tropsch coal-to-liquid production in oxygen-blown mode. Data are to be provided to DOE to determine kinetic rates of devolatilization and tar cracking.

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

    SciTech Connect (OSTI)

    M.E. Lumia; C.A. Gentile

    2002-01-18T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Lumia, M E

    2002-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Yoder Jr, Graydon L [ORNL] [ORNL; Aaron, Adam M [ORNL] [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)] [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL] [ORNL; Holcomb, David Eugene [ORNL] [ORNL; Kisner, Roger A [ORNL] [ORNL; Peretz, Fred J [ORNL] [ORNL; Robb, Kevin R [ORNL] [ORNL; Wilgen, John B [ORNL] [ORNL; Wilson, Dane F [ORNL] [ORNL

    2014-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Craig Wise

    2011-12-01T23:59:59.000Z

    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.”

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

    SciTech Connect (OSTI)

    Donna P. Guillen

    2012-07-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Hakan Ozaltun & Herman Shen

    2011-11-01T23:59:59.000Z

    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.

  2. 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

    2013-02-01T23:59:59.000Z

    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.

  3. Tracer Testing for Estimating Heat Transfer Area in Fractured Reservoirs

    E-Print Network [OSTI]

    Pruess, Karsten; van Heel, Ton; Shan, Chao

    2004-01-01T23:59:59.000Z

    Heat Flow in Fractured Reservoirs, SPE Advanced TechnologyTransfer Area in Fractured Reservoirs Karsten Pruess 1 , Tonbehavior arises in fractured reservoirs. As cold injected

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

    SciTech Connect (OSTI)

    S. Blaine Grover; David A. Petti

    2014-05-01T23:59:59.000Z

    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

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

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    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.

  6. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    SciTech Connect (OSTI)

    T. R. Allen; J. B. Benson; J. A. Foster; F. M. Marshall; M. K. Meyer; M. C. Thelen

    2009-05-01T23:59:59.000Z

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team projects and faculty/staff exchanges. In June of 2008, the first week-long ATR NSUF Summer Session was attended by 68 students, university faculty and industry representatives. The Summer Session featured presentations by 19 technical experts from across the country and covered topics including irradiation damage mechanisms, degradation of reactor materials, LWR and gas reactor fuels, and non-destructive evaluation. High impact research results from leveraging the entire research infrastructure, including universities, industry, small business, and the national laboratories. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. Current partner facilities include the MIT Reactor, the University of Michigan Irradiated Materials Testing Laboratory, the University of Wisconsin Characterization Laboratory, and the University of Nevada, Las Vegas transmission Electron Microscope User Facility. Needs for irradiation of material specimens at tightly controlled temperatures are being met by dedication of a large in-pile pressurized water loop facility for use by ATR NSUF users. Several environmental mechanical testing systems are under construction to determine crack growth rates and fracture toughness on irradiated test systems.

  7. Thermally Simulated 32kW Direct-Drive Gas-Cooled Reactor: Design, Assembly, and Test

    SciTech Connect (OSTI)

    Godfroy, Thomas J.; Bragg-Sitton, Shannon M. [NASA Marshall Space Flight Center, TD40, Huntsville, Alabama, 35812 (United States); University of Michgan, Dept. of Nuclear Engineering and Radiological Sciences, Ann Arbor MI 48109 (United States); Kapernick, Richard J. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2004-02-04T23:59:59.000Z

    One of the power systems under consideration for nuclear electric propulsion is a direct-drive gas-cooled reactor coupled to a Brayton cycle. In this system, power is transferred from the reactor to the Brayton system via a circulated closed loop gas. To allow early utilization, system designs must be relatively simple, easy to fabricate, and easy to test using non-nuclear heaters to closely mimic heat from fission. This combination of attributes will allow pre-prototypic systems to be designed, fabricated, and tested quickly and affordably. The ability to build and test units is key to the success of a nuclear program, especially if an early flight is desired. The ability to perform very realistic non-nuclear testing increases the success probability of the system. In addition, the technologies required by a concept will substantially impact the cost, time, and resources required to develop a successful space reactor power system. This paper describes design features, assembly, and test matrix for the testing of a thermally simulated 32kW direct-drive gas-cooled reactor in the Early Flight Fission - Test Facility (EFF-TF) at Marshall Space Flight Center. The reactor design and test matrix are provided by Los Alamos National Laboratories.

  8. Improving the AGR Fuel Testing Power Density Profile Versus Irradiation-Time in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Gray S. Chang; David A. Petti; John T. Maki; Misti A. Lillo

    2009-05-01T23:59:59.000Z

    The Very High Temperature gas-cooled Reactor (VHTR), which is currently being developed, achieves simplification of safety through reliance on ceramic-coated fuel particles. Each TRISO-coated fuel particle has its own containment which serves as the principal barrier against radionuclide release under normal operating and accident conditions. These fuel particles, in the form of graphite fuel compacts, are currently undergoing a series of irradiation tests in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) to support the Advanced Gas-Cooled Reactor (AGR) fuel qualification program. A representive coated fuel particle with an 235U enrichment of 19.8 wt% was used in this analysis. The fuel burnup analysis tool used to perform the neutronics study reported herein, couples the Monte Carlo transport code MCNP, with the radioactive decay and burnup code ORIGEN2. The fuel burnup methodology known as Monte-Carlo with ORIGEN2 (MCWO) was used to evaluate the AGR experiment assembly and demonstrate compliance with ATR safety requirements. For the AGR graphite fuel compacts, the MCWO-calculated fission power density (FPD) due to neutron fission in 235U is an important design parameter. One of the more important AGR fuel testing requirements is to maintain the peak fuel compact temperature close to 1250°C throughout the proposed irradiation campaign of 550 effective full power days (EFPDs). Based on the MCWO-calculated FPD, a fixed gas gap size was designed to allow regulation of the fuel compact temperatures throughout the entire fuel irradiation campaign by filling the gap with a mixture of helium and neon gases. The chosen fixed gas gap can only regulate the peak fuel compact temperature in the desired range during the irradiation test if the ratio of the peak power density to the time-dependent low power density (P/T) at 550 EFPDs is less than 2.5. However, given the near constant neutron flux within the ATR driver core and the depletion of 235U in the graphite fuel compacts versus EFPD, the P/T ratio was calculated to be 5.3, which is unacceptable given the fuel compact temperature control requirement. To flatten the FPD profile versus EFPDs, two proposed options are – (a) add fertile (232Th) particles to the fuel compact and (b) add burnable absorber (B4C) to the graphite holder. The effectiveness of these two proposed options to flatten the FPD profile versus EFPDs were investigated and the results are compared in this study.

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

    SciTech Connect (OSTI)

    Donna Post Guillen

    2012-11-01T23:59:59.000Z

    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.

  10. 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)

    2012-07-01T23:59:59.000Z

    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)

  11. 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

    2015-05-23T23:59:59.000Z

    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.

  12. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. B. Grover

    2007-05-01T23:59:59.000Z

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) 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 new 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 irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new 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.1,2 The experiments, which will each consist of 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 final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation were completed in 2006. The experiment was inserted in the ATR in December 2006, and will serve as a shakedown test of the multi-capsule experiment design that will be used in the subsequent irradiations as well as a test of the early variants of the fuel produced under this program. The experiment test train as well as the monitoring, control, and data collection systems are discussed.

  13. 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.

    1982-03-01T23:59:59.000Z

    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.

  14. First Test of Lorentz Violation with a Reactor-based Antineutrino Experiment

    E-Print Network [OSTI]

    Abe, Y; Anjos, J C dos; Bergevin, M; Bernstein, A; Bezerra, T J C; Bezrukhov, L; Blucher, E; Bowden, N S; Buck, C; Busenitz, J; Cabrera, A; Caden, E; Camilleri, L; Carr, R; Cerrada, M; Chang, P -J; Chimenti, P; Classen, T; Collin, A P; Conover, E; Conrad, J M; Crespo-Anadón, J I; Crum, K; Cucoanes, A; D'Agostino, M V; Damon, E; Dawson, J V; Dazeley, S; Dietrich, D; Djurcic, Z; Dracos, M; Durand, V; Ebert, J; Efremenko, Y; Elnimr, M; Erickson, A; Fallot, M; Fechner, M; von Feilitzsch, F; Felde, J; Fischer, V; Franco, D; Franke, A J; Franke, M; Furuta, H; Gama, R; Gil-Botella, I; Giot, L; Göger-Neff, M; Gonzalez, L F G; Goodman, M C; Goon, J TM; Greiner, D; Haag, N; Habib, S; Hagner, C; Hara, T; Hartmann, F X; Haser, J; Hatzikoutelis, A; Hayakawa, T; Hofmann, M; Horton-Smith, G A; Ishitsuka, M; Jochum, J; Jollet, C; Jones, C L; Kaether, F; Kalousis, L N; Kamyshkov, Y; Kaplan, D M; Katori, T; Kawasaki, T; Keefer, G; Kemp, E; de Kerret, H; Konno, T; Kryn, D; Kuze, M; Lachenmaier, T; Lane, C E; Lasserre, T; Letourneau, A; Lhuillier, D; Lima, H P; Lindner, M; López-Castanő, J M; LoSecco, J M; Lubsandorzhiev, B K; Lucht, S; McKee, D; Maeda, J; Maesano, C N; Mariani, C; Maricic, J; Martino, J; Matsubara, T; Mention, G; Meregaglia, A; Meyer, M; Miletic, T; Milincic, R; Miyata, H; Mueller, Th A; Nagasaka, Y; Nakajima, K; Novella, P; Obolensky, M; Oberauer, L; Onillon, A; Osborn, A; Ostrovskiy, I; Palomares, C; Pepe, I M; Perasso, S; Perrin, P; Pfahler, P; Porta, A; Potzel, W; Pronost, G; Reichenbacher, J; Reinhold, B; Remoto, A; Röhling, M; Roncin, R; Roth, S; Rybolt, B; Sakamoto, Y; Santorelli, R; Sato, F; Schönert, S; Schoppmann, S; Schwetz, T; Shaevitz, M H; Shrestha, D; Sida, J -L; Sinev, V; Skorokhvatov, M; Smith, E; Spitz, J; Stahl, A; Stancu, I; Stokes, L F F; Strait, M; Stüken, A; Suekane, F; Sukhotin, S; Sumiyoshi, T; Sun, Y; Terao, K; Tonazzo, A; Toups, M; Thi, H H Trinh; Valdiviesso, G; Veyssiere, C; Wagner, S; Watanabe, H; White, B; Wiebusch, C; Winslow, L; Worcester, M; Wurm, M; Yanovitch, E; Yermia, F; Zimmer, V

    2012-01-01T23:59:59.000Z

    We present a search for Lorentz violation with 8249 candidate electron antineutrino events taken by the Double Chooz experiment in 227.9 live days of running. This analysis, featuring a search for a sidereal time dependence of the events, is the first test of Lorentz invariance using a reactor-based antineutrino source. No sidereal variation is present in the data and the disappearance results are consistent with sidereal time independent oscillations. Under the Standard-Model Extension (SME), we set the first limits on fourteen Lorentz violating coefficients associated with transitions between electron and tau flavor, and set two competitive limits associated with transitions between electron and muon flavor.

  15. 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

    2013-07-01T23:59:59.000Z

    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.

  16. 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

    2008-02-25T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Yonehara, Katsuya; /Fermilab

    2011-12-02T23:59:59.000Z

    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.

  18. SRF Test Areas Cryogenic System Controls Graphical User Interface

    SciTech Connect (OSTI)

    DeGraff, B.D.; Ganster, G.; Klebaner, A.; Petrov, A.D.; Soyars, W.M.; /Fermilab

    2011-06-09T23:59:59.000Z

    Fermi National Accelerator Laboratory has constructed a superconducting 1.3 GHz cavity test facility at Meson Detector Building (MDB) and a superconducting 1.3 GHz cryomodule test facility located at the New Muon Lab Building (NML). The control of these 2K cryogenic systems is accomplished by using a Synoptic graphical user interface (GUI) to interact with the underlying Fermilab Accelerator Control System. The design, testing and operational experience of employing the Synoptic client-server system for graphical representation will be discussed. Details on the Synoptic deployment to the MDB and NML cryogenic sub-systems will also be discussed. The implementation of the Synoptic as the GUI for both NML and MDB has been a success. Both facilities are currently fulfilling their individual roles in SCRF testing as a result of successful availability of the cryogenic systems. The tools available for creating Synoptic pages will continue to be developed to serve the evolving needs of users.

  19. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    SciTech Connect (OSTI)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01T23:59:59.000Z

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

  20. Corrective Action Investigation Plan for Corrective Action Unit 375: Area 30 Buggy Unit Craters, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    Patrick Matthews

    2010-03-01T23:59:59.000Z

    Corrective Action Unit (CAU) 375 is located in Areas 25 and 30 of the Nevada Test Site, which is approximately 65 miles northwest of Las Vegas, Nevada. Corrective Action Unit 375 comprises the two corrective action sites (CASs) listed below: • 25-23-22, Contaminated Soils Site • 30-45-01, U-30a, b, c, d, e Craters Existing information on the nature and extent of potential contamination present at the CAU 375 CASs is insufficient to evaluate and recommend corrective action alternatives (CAAs). This document details an investigation plan that will provide for the gathering of sufficient information to evaluate and recommend CAAs. Corrective Action Site 25-23-22 is composed of the releases associated with nuclear rocket testing at Test Cell A (TCA). Test Cell A was used to test and develop nuclear rocket motors as part of the Nuclear Rocket Development Station from its construction in 1958 until 1966, when rocket testing began being conducted at Test Cell C. The rocket motors were built with an unshielded nuclear reactor that produced as much as 1,100 kilowatts (at full power) to heat liquid hydrogen to 4,000 degrees Fahrenheit, at which time the expanded gases were focused out a nozzle to produce thrust. The fuel rods in the reactor were not clad and were designed to release fission fragments to the atmosphere, but due to vibrations and loss of cooling during some operational tests, fuel fragments in excess of planned releases became entrained in the exhaust and spread in the immediate surrounding area. Cleanup efforts have been undertaken at times to collect the fuel rod fragments and other contamination. Previous environmental investigations in the TCA area have resulted in the creation of a number of use restrictions. The industrial area of TCA is encompassed by a fence and is currently posted as a radioactive material area. Corrective Action Site 30-45-01 (releases associated with the Buggy Plowshare test) is located in Area 30 on Chukar Mesa. It was a Plowshare test where five nuclear devices were buried 140 feet (ft) deep in a row at 150-ft intervals. These devices were detonated on March 12, 1968, to produce a trench 254 ft wide, 865 ft long, and 70 ft deep. The mesa where the test was conducted is surrounded on three sides by ravines, and the entire end of the mesa is fenced and posted as a contamination area. These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend 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 December 2, 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 375.

  1. Testing and Commissioning of a Multifunctional Tool for the Dismantling of the Activated Internals of the KNK Reactor Shaft - 13524

    SciTech Connect (OSTI)

    Rothschmitt, Stefan; Graf, Anja [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein-Leopoldshafen (Germany)] [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein-Leopoldshafen (Germany); Bauer, Stefan; Klute, Stefan; Koselowski, Eiko [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany)] [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany); Hendrich, Klaus [Ingenieurbuero Hendrich, Moerikeweg 14, 75015 Bretten (Germany)] [Ingenieurbuero Hendrich, Moerikeweg 14, 75015 Bretten (Germany)

    2013-07-01T23:59:59.000Z

    The Compact Sodium Cooled Reactor Facility Karlsruhe (KNK), a prototype reactor to demonstrate the Fast Breeder Reactor Technology in Germany, was in operation from 1971 to 1991. The dismantling activities started in 1991. The project aim is the green field in 2020. Most of the reactor internals as well as the primary and secondary cooling loops are already dismantled. The total contaminated sodium inventory has already been disposed of. Only the high activated reactor vessel shielding structures are remaining. Due to the high dose rates these structures must be dismantled remotely. For the dismantling of the primary shielding of the reactor vessel, 12 stacked cast iron blocks with a total mass of 90 Mg and single masses up to 15.5 Mg, a remote-controlled multifunctional dismantling device (HWZ) was designed, manufactured and tested in a mock-up. After successful approval of the test sequences by the authorities, the HWZ was implemented into the reactor building containment for final assembling of the auxiliary equipment and subsequent hot commissioning in 2012. Dismantling of the primary shielding blocks is scheduled for early 2013. (authors)

  2. Fabrication, Inspection, and Test Plan for the Advanced Test Reactor (ATR) High-Power Mixed-Oxide (MOX) Fuel Irradiation Project

    SciTech Connect (OSTI)

    Wachs, G. W.

    1998-09-01T23:59:59.000Z

    The Department of Energy (DOE) Fissile Disposition Program (FMDP) has announced that reactor irradiation of Mixed-Oxide (MOX) fuel is one of the preferred alternatives for disposal of surplus weapons-usable plutonium (Pu). MOX fuel has been utilized domestically in test reactors and on an experimental basis in a number of Commercial Light Water Reactors (CLWRs). Most of this experience has been with Pu derived from spent low enriched uranium (LEU) fuel, known as reactor grade (RG) Pu. The High-Power MOX fuel test will be irradiated in the Advanced Test Reactor (ATR) to provide preliminary data to demonstrate that the unique properties of surplus weapons-derived or weapons-grade (WG) plutonium (Pu) do not compromise the applicability of this MOX experience base. The purpose of the high-power experiment, in conjunction with the currently ongoing average-power experiment at the ATR, is to contribute new information concerning the response of WG plutonium under more severe irradiation conditions typical of the peak power locations in commercial reactors. In addition, the high-power test will contribute experience with irradiation of gallium-containing fuel to the database required for resolution of generic CLWR fuel design issues. The distinction between "high-power" and "average-power" relates to the position within the nominal CLWR core. The high-power test project is subject to a number of requirements, as discussed in the Fissile Materials Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation High-Power Test Project Plan (ORNL/MD/LTR-125).

  3. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    SciTech Connect (OSTI)

    MH Lane

    2006-02-15T23:59:59.000Z

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  4. 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

    2014-01-01T23:59:59.000Z

    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.

  5. 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

    2011-03-01T23:59:59.000Z

    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.

  6. In-pile testing on thermionic space nuclear power fuel using a TRIGA reactor

    SciTech Connect (OSTI)

    Razvi, J.; Whittemore, W.L. [General Atomics, San Diego, CA (United States)

    1988-07-01T23:59:59.000Z

    The U. S. Department of Energy is sponsoring a thermionic technology development program at General Atomics (GA) to address the feasibility issues associated with the use of thermionic fuel elements (TFE) for space based nuclear power systems. The technology program being carried out at GA during the 1980s is aimed at investigating lifetime issues of the fuelled emitter and insulator materials proposed for use in thermionic reactor based space power systems, both experimentally and analytically, with a goal towards developing a TFE with a lifetime of at least seven years. The experimental investigations of TFE lifetime issues consist of performing in-pile irradiations on prototype units using the 1.5 Mw TRIGA Mark F research reactor to validate TFE emitter deformation models and to measure the amount of emitter deformation as a function of fuel burnup, temperature and emitter thickness. To this end, nine thermionic emitters, fueled with uranium oxide and contained in three irradiation capsules, have been designed and fabricated to date, and are being tested in-core in the Mark F to study emitter materials integrity under irradiation conditions. In addition, a capsule consisting of only insulator test articles has also been irradiated in the Mark F to study insulator materials integrity under irradiation and applied voltage.

  7. Development and Testing of a Groundwater Management Model for the Faultless Underground Nuclear Test, Central Nevada Test Area

    SciTech Connect (OSTI)

    Douglas P. Boyle; Gregg Lamorey; Scott Bassett; Greg Pohll; Jenny Chapman

    2006-01-25T23:59:59.000Z

    This document describes the development and application of a user-friendly and efficient groundwater management model of the Central Nevada Test Area (CNTA) and surrounding areas that will allow the U.S. Department of Energy and state personnel to evaluate the impact of future proposed scenarios. The management model consists of a simple hydrologic model within an interactive groundwater management framework. This framework is based on an object user interface that was developed by the U.S. Geological Survey and has been used by the Desert Research Institute researchers and others to couple disparate environmental resource models, manage the necessary temporal and spatial data, and evaluate model results for management decision making. This framework was modified and applied to the CNTA and surrounding Hot Creek Valley. The utility of the management model was demonstrated through the application of hypothetical future scenarios including mineral mining, regional expansion of agriculture, geothermal energy production, and export of water to large urban areas outside the region. While the results from some of the scenarios indicated potential impacts to the region near CNTA and others did not, together they demonstrate the usefulness of the management tool for managers who need to evaluate the impact proposed changes in groundwater use in or near CNTA may have on radionuclide migration.

  8. The MuCool Test Area and RF Program

    SciTech Connect (OSTI)

    Torun, Y.; Huang, D.; /IIT, Chicago; Norem, J.; /Argonne; Palmer, Robert B.; Stratakis, Diktys; /Brookhaven; Bross, A.; Chung, M.; Jansson, A.; Moretti, A.; Yonehara, K.; /Fermilab; Li, D.; /LBL, Berkeley /Jefferson Lab

    2010-05-01T23:59:59.000Z

    The MuCool RF Program focuses on the study of normal conducting RF structures operating in high magnetic field for applications in muon ionization cooling for Neutrino Factories and Muon Colliders. Here we give an overview of the program, which includes a description of the test facility and its capabilities, the current test program, and the status of a cavity that can be rotated in the magnetic field, which allows for a detailed study of the maximum stable operating gradient vs. magnetic field strength and angle.

  9. The MuCool Test Area and RF Program

    SciTech Connect (OSTI)

    Bross, A D; Jansson, A; Moretti, A; Yonehara, K; Huang, D; Torun, Y; Li, D; Norem, J; Palmer, R B; Stratakis, D

    2010-05-01T23:59:59.000Z

    The MuCool RF Program focuses on the study of normal conducting RF structures operating in high magnetic field for applications in muon ionization cooling for Neutrino Factories and Muon Colliders. This paper will give an overview of the program, which will include a description of the test facility and its capabilities, the current test program, and the status of a cavity that can be rotated in the magnetic field which allows for a more detailed study of the maximum stable operating gradient vs. magnetic field strength and angle.

  10. Tracer Testing 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to:Ezfeedflag JumpID-f <MaintainedInformationThePtyTown Hall MeetingTracer TestingTesting

  11. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Dotson, CW

    1980-08-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  12. Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01T23:59:59.000Z

    The preceding paper described designs and analyses of thermionic reactors employing full-core-length single-cell converters with their heated emitters located on the outside of their internally cooled collectors, and it presented results of detailed parametric analyses which illustrate the benefits of this unconventional design. The present paper describes the fabrication and testing of full-length prototypical converters, both unfueled and fueled, and presents parametric results of electrically heated tests. The unfueled converter tests demonstrated the practicality of operating such long converters without shorting across a 0.3-mm interelectrode gap. They produced a measured peak output of 751 watts(e) from a single diode and a peak efficiency of 15.4%. The fueled converter tests measured the parametric performance of prototypic UO(subscript 2)-fueled converters designed for subsequent in-pile testing. They employed revolver-shaped tungsten elements with a central emitter hole surrounded by six fuel chambers. The full-length converters were heated by a water-cooled RF-induction coil inside an ion-pumped vacuum chamber. This required development of high-vacuum coaxial RF feedthroughs. In-pile test rules required multiple containment of the UO (subscript 2)-fuel, which complicated the fabrication of the test article and required successful development of techniques for welding tungsten and other refractory components. The test measured a peak power output of 530 watts(e) or 7.1 watts/cm (superscript 2) at an efficiency of 11.5%. There are three copies in the file. Cross-Reference a copy FSC-ESD-217-94-529 in the ESD files with a CID #8574.

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

    E-Print Network [OSTI]

    Howel, E. S.

    1990-01-01T23:59:59.000Z

    that is displaced by outside a% in a 1 hour period. The Manual J Load Calculation Handbook 1.1) specifies ACH values n a 2000 square foot home from .2 to 1.8. Load calculations based on Manual J procedures on a home of 1800 square feet in the Houston, Texas... area show that when ACH is .7 (average new home), infiltration becomes a significaht contributor to heatin (20%) and air conditionyng (34% ) loads. Additionally, when the 1800 square foot home has an ACH of 1.5 (average existin home...

  14. Single Channel Testing for Characterization of the Direct Gas Cooled Reactor and the SAFE-100 Heat Exchanger

    SciTech Connect (OSTI)

    Bragg-Sitton, S.M. [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States); Kapernick, R. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Godfroy, T.J. [Propulsion Research Center, NASA Marshall Space Flight Center, Huntsville, AL 35812 (United States)

    2004-02-04T23:59:59.000Z

    Experiments have been designed to characterize the coolant gas flow in two space reactor concepts that are currently under investigation by NASA Marshall Space Flight Center and Los Alamos National Laboratory: the direct-drive gas-cooled reactor (DDG) and the SAFE-100 heatpipe-cooled reactor (HPR). For the DDG concept, initial tests have been completed to measure pressure drop versus flow rate for a prototypic core flow channel, with gas exiting to atmospheric pressure conditions. The experimental results of the completed DDG tests presented in this paper validate the predicted results to within a reasonable margin of error. These tests have resulted in a re-design of the flow annulus to reduce the pressure drop. Subsequent tests will be conducted with the re-designed flow channel and with the outlet pressure held at 150 psi (1 MPa). Design of a similar test for a nominal flow channel in the HPR heat exchanger (HPR-HX) has been completed and hardware is currently being assembled for testing this channel at 150 psi. When completed, these test programs will provide the data necessary to validate calculated flow performance for these reactor concepts (pressure drop and film temperature rise)

  15. 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to: navigation, searchOfRose Bend <Stevens Jump to:source History View NewMarysville Test Well

  16. Tracer Testing At Coso Geothermal Area (1993) | 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are beingZealand Jump to:Ezfeedflag JumpID-f <MaintainedInformationThePtyTown Hall MeetingTracer Testing

  17. 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

    2001-01-01T23:59:59.000Z

    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).

  18. Gas cooled fast reactor control rod drive mechanism deceleration unit. Test program

    SciTech Connect (OSTI)

    Wagner, T.H.

    1981-10-01T23:59:59.000Z

    This report presents the results of the airtesting portion of the proof-of-principle testing of a Control Rod Scram Deceleration Device developed for use in the Gas Cooled Fast Reactor (GCFR). The device utilizes a grooved flywheel to decelerate the translating assembly (T/A). Two cam followers on the translating assembly travel in the flywheel grooves and transfer the energy of the T/A to the flywheel. The grooves in the flywheel are straight for most of the flywheel length. Near the bottom of the T/A stroke the grooves are spiraled in a decreasing slope helix so that the cam followers accelerate the flywheel as they transfer the energy of the falling T/A. To expedite proof-of-principle testing, some of the materials used in the fabrication of certain test article components were not prototypic. With these exceptions the concept appears to be acceptable. The initial test of 300 scrams was completed with only one failure and the failure was that of a non-prototypic cam follower outer sleeve material.

  19. Tensile and impact testing of an HFBR (High Flux Beam Reactor) control rod follower

    SciTech Connect (OSTI)

    Czajkowski, C.J.; Schuster, M.H.; Roberts, T.C.; Milian, L.W.

    1989-08-01T23:59:59.000Z

    The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (K{sub I}) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with K{sub max} values of 24.8 {plus minus} 1.0 Ksi{radical}in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs.

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

    SciTech Connect (OSTI)

    David W. Nigg

    2013-09-01T23:59:59.000Z

    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).

  1. A fission matrix based validation protocol for computed power distributions in the advanced test reactor

    SciTech Connect (OSTI)

    Nielsen, J. W. [Idaho National Laboratory, MS 3840, PO Box 1625, Idaho Falls, ID 83415 (United States); Nigg, D. W. [Idaho National Laboratory, MS 3860, PO Box 1625, Idaho Falls, ID 83415 (United States); LaPorta, A. W. [Idaho National Laboratory, MS 7136, PO Box 1625, Idaho Falls, ID 83415 (United States)

    2013-07-01T23:59:59.000Z

    The Idaho National Laboratory (INL) has been engaged in a significant multi year effort to modernize the computational reactor physics tools and validation procedures used to support operations of the Advanced Test Reactor (ATR) and its companion critical facility (ATRC). Several new protocols for validation of computed neutron flux distributions and spectra as well as for validation of computed fission power distributions, based on new experiments and well-recognized least-squares statistical analysis techniques, have been under development. In the case of power distributions, estimates of the a priori ATR-specific fuel element-to-element fission power correlation and covariance matrices are required for validation analysis. A practical method for generating these matrices using the element-to-element fission matrix is presented, along with a high-order scheme for estimating the underlying fission matrix itself. The proposed methodology is illustrated using the MCNP5 neutron transport code for the required neutronics calculations. The general approach is readily adaptable for implementation using any multidimensional stochastic or deterministic transport code that offers the required level of spatial, angular, and energy resolution in the computed solution for the neutron flux and fission source. (authors)

  2. 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

    2010-09-01T23:59:59.000Z

    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).

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

    E-Print Network [OSTI]

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

    2015-01-01T23:59:59.000Z

    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 na...

  4. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  5. Fire loading calculations for 300 Area N Reactor Fuel Fabrication and Storage Facility

    SciTech Connect (OSTI)

    Myott, C.F.

    1994-01-24T23:59:59.000Z

    Fire loading analyses were provided for the N Reactor Fuel Supply Buildings 3712, 3716, 303A, 303B, 303E, 303G, and 303K. Fire loading calculations, maximum temperatures, and fire durations were provided to support the safety analyses documentation. The ``combustibles`` for this document include: wood, cardboard, cloth, and plastic, and does not include the uranium and fuel assembly loading. The information in this document will also be used to support the fire hazard analysis for the same buildings, therefore, it is assumed that sprinkler systems do not work, or the maximum possible fire loss is assumed.

  6. Short Term Irradiation Test of Fuel Containing Minor Actinides Using the Experimental Fast Reactor Joyo

    SciTech Connect (OSTI)

    Sekine, Takashi; Soga, Tomonori; Koyama, Shin-ichi; Aoyama, Takafumi [Oarai Research and Development Center, Japan Atomic Energy Agency. 4002 Narita, Oarai, Ibaraki 311-1393 (Japan); Wootan, David [Pacific Northwest National Laboratoy, M/S K8-34, P.O. Box 999 Richland, WA 99352 (United States)

    2007-07-01T23:59:59.000Z

    A mixed oxide containing minor actinides (MA-MOX) fuel irradiation program is being conducted using the experimental fast rector Joyo of the Japan Atomic Energy Agency to research early thermal behavior of MA-MOX fuel. Two irradiation experiments were conducted as part of the short-term phase of this program in May and August 2006. Six prepared fuel pins included MOX fuel containing 3% or 5% americium (Am-MOX), and MOX fuel containing 2% americium and 2% neptunium (Np/Am-MOX). The first test was conducted with high linear heat rates of approximately 430 W/cm maintained during only 10 minutes. After 10 minutes irradiation test, the test subassembly was transferred to the hot cell facility and an Am-MOX pin and a Np/Am-MOX pin were replaced with dummy pins with neutron dosimeters. The test subassembly loaded with the remaining four fuel pins was re-irradiated in Joyo for 24-hours in August 2006 at nearly the same linear power to obtain re-distribution data on MA-MOX fuel. The linear heat rate for each MA-MOX test fuel pin was calculated using the Monte Carlo calculation code MCNP. The calculated fission rates were compared with the measured data based on the Nd-148 method. The maximum linear heat rate was approximately 444{+-}19 W/cm at the actual reactor power of 119.6 MWt. Post irradiation examination of these pins to confirm the absence of fuel melting and the local concentration under irradiation of NpO{sub 2-x} or AmO{sub 2-x}, in the (U,Pu)0{sub 2-x}, fuel are underway. The test results are expected to reduce uncertainties on the margin in the thermal design for MA-MOX fuel. (authors)

  7. Human factors engineering evaluation of the Advanced Test Reactor Control Room

    SciTech Connect (OSTI)

    Boone, M.P.; Banks, W.W.

    1980-12-01T23:59:59.000Z

    The information presented here represents preliminary findings related to an ongoing human engineering evaluation of the Advanced Test Reactor (ATR) Control Room. Although many of the problems examined in this report have been previously noted by ATR operations personnel, the systematic approach used in this investigation produced many new insights. While many violations of Human Engineering military standards (MIL-STD) are noted, and numerous recommendations made, the recommendations should be examined cautiously. The reason for our suggested caution lies in the fact that many ATR operators have well over 10-years experience in operating the controls, meters, etc. Hence, it is assumed adaptation to the existing system is quite developed and the introduction of hardware/control changes, even though the changes enhance the system, may cause short-term (or long-term, depending upon the amount of operator experience and training) adjustment problems for operators adapting to the new controls/meters and physical layout.

  8. Advanced Test Reactor Critical Facility safety analysis report five year currency review

    SciTech Connect (OSTI)

    Napper, P.R.; Carpenter, W.R.; Garner, R.W.

    1991-05-01T23:59:59.000Z

    By DOE-ID Order 5481.1A, a five year currency review is required of the Safety Analysis Reports of all ID or ID contractor operations having hazards of a type and magnitude not routinely encountered and/or accepted by the public. In keeping with this order, a currency review has been performed of the Advanced Test Reactor Critical Facility (ADTRC) Safety Analysis Report (SAR), Issue 003, 1990. The objectives of this currency review were to: evaluate the content, completeness, clarity of presentation and compliance with NRC Regulatory Guides and DOE Orders, etc., and evaluate the technical content of the SAR, particularly the Technical Specifications, and to evaluate the safety of continued operation of the ATRC. The reviewers concluded that although improvements may be needed in the overall content, clarity, and demonstration of compliance with current orders and regulations, the safety of the ATRC is in no way compromised and no unreviewed safety questions were identified. 6 figs., 3 tabs.

  9. 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

  10. Atmospheric Pressure Reactor System | EMSL

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

    Atmospheric Pressure Reactor System Atmospheric Pressure Reactor System The atmospheric pressure reactor system is designed for testing the efficiency of various catalysts for the...

  11. Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    S. Blaine Grover

    2009-05-01T23:59:59.000Z

    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 new 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 at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens 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 be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. 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 final design phase for the first experiment was completed in September 2008, and the fabrication and assembly of the experiment test train as well as installation and testing of the control and support systems that will monitor and control the experiment during irradiation are being completed in early calendar 2009. The first experiment is scheduled to be ready for insertion in the ATR by April 30, 2009. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and data collection systems.

  12. Hydrothermal Testing of K Basin Sludge and N Reactor Fuel at Sludge Treatment Project Operating Conditions

    SciTech Connect (OSTI)

    Delegard, Calvin H.; Schmidt, Andrew J.; Thornton, Brenda M.

    2007-03-30T23:59:59.000Z

    The Sludge Treatment Project (STP), managed for the U. S. DOE by Fluor Hanford (FH), was created to design and operate a process to eliminate uranium metal from K Basin sludge prior to packaging for Waste Isolation Pilot Plant (WIPP). The STP process uses high temperature liquid water to accelerate the reaction, produce uranium dioxide from the uranium metal, and safely discharge the hydrogen. Under nominal process conditions, the sludge will be heated in pressurized water at 185°C for as long as 72 hours to assure the complete reaction (corrosion) of up to 0.25-inch diameter uranium metal pieces. Under contract to FH, the Pacific Northwest National Laboratory (PNNL) conducted bench-scale testing of the STP hydrothermal process in November and December 2006. Five tests (~50 ml each) were conducted in sealed, un-agitated reaction vessels under the hydrothermal conditions (e.g., 7 to 72 h at 185°C) of the STP corrosion process using radioactive sludge samples collected from the K East Basin and particles/coupons of N Reactor fuel also taken from the K Basins. The tests were designed to evaluate and understand the chemical changes that may be occurring and the effects that any changes would have on sludge rheological properties. The tests were not designed to evaluate engineering aspects of the process. The hydrothermal treatment affected the chemical and physical properties of the sludge. In each test, significant uranium compound phase changes were identified, resulting from dehydration and chemical reduction reactions. Physical properties of the sludge were significantly altered from their initial, as-settled sludge values, including, shear strength, settled density, weight percent water, and gas retention.

  13. Final Site Specific Decommissioning Inspection Report #2 for the University of Washington Research and Test Reactor, Seattle, Washington

    SciTech Connect (OSTI)

    S.J. Roberts

    2007-03-20T23:59:59.000Z

    During the period of August through November 2006, ORISE performed a comprehensive IV at the University of Washington Research and Test Reactor Facility. The objective of the ORISE IV was to validate the licensee’s final status survey processes and data, and to assure the requirements of the DP and FSSP were met.

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

    SciTech Connect (OSTI)

    C. M. Obi

    2000-12-01T23:59:59.000Z

    The Area 25 Reactor Maintenance, Assembly, and Disassembly Decontamination Facility is identified in the Federal Facility Agreement and Consent Order (FFACO) as Corrective Action Unit (CAU) 254. CAU 254 is located in Area 25 of the Nevada Test Site and consists of a single Corrective Action Site CAS 25-23-06. CAU 254 will be closed, in accordance with the FFACO of 1996. CAU 254 was used primarily to perform radiological decontamination and consists of Building 3126, two outdoor decontamination pads, and surrounding soil within an existing perimeter fence. The site was used to decontaminate nuclear rocket test-car hardware and tooling from the early 1960s through the early 1970s, and to decontaminate a military tank in the early 1980s. The site characterization results indicate that, in places, the surficial soil and building materials exceed clean-up criteria for organic compounds, metals, and radionuclides. Closure activities are expected to generate waste streams consisting of nonhazardous construction waste. petroleum hydrocarbon waste, hazardous waste, low-level radioactive waste, and mixed waste. Some of the wastes exceed land disposal restriction limits and will require off-site treatment before disposal. The recommended corrective action was revised to Alternative 3- ''Unrestricted Release Decontamination, Verification Survey, and Dismantle Building 3126,'' in an addendum to the Correction Action Decision Document.

  15. Deterministic Modeling of the High Temperature Test Reactor with DRAGON-HEXPEDITE

    SciTech Connect (OSTI)

    J. Ortensi; M.A. Pope; R.M. Ferrer; J.J. Cogliati; J.D. Bess; A.M. Ougouag

    2010-10-01T23:59:59.000Z

    The Idaho National Laboratory (INL) is tasked with the development of reactor physics analysis capability of the Next Generation Nuclear Power (NGNP) project. In order to examine the INL’s current prismatic reactor analysis tools, the project is conducting a benchmark exercise based on modeling the High Temperature Test Reactor (HTTR). This exercise entails the development of a model for the initial criticality, a 19 fuel column thin annular core, and the fully loaded core critical condition with 30 fuel columns. Special emphasis is devoted to physical phenomena and artifacts in HTTR that are similar to phenomena and artifacts in the NGNP base design. The DRAGON code is used in this study since it offers significant ease and versatility in modeling prismatic designs. DRAGON can generate transport solutions via Collision Probability (CP), Method of Characteristics (MOC) and Discrete Ordinates (Sn). A fine group cross-section library based on the SHEM 281 energy structure is used in the DRAGON calculations. The results from this study show reasonable agreement in the calculation of the core multiplication factor with the MC methods, but a consistent bias of 2–3% with the experimental values is obtained. This systematic error has also been observed in other HTTR benchmark efforts and is well documented in the literature. The ENDF/B VII graphite and U235 cross sections appear to be the main source of the error. The isothermal temperature coefficients calculated with the fully loaded core configuration agree well with other benchmark participants but are 40% higher than the experimental values. This discrepancy with the measurement partially stems from the fact that during the experiments the control rods were adjusted to maintain criticality, whereas in the model, the rod positions were fixed. In addition, this work includes a brief study of a cross section generation approach that seeks to decouple the domain in order to account for neighbor effects. This spectral interpenetration is a dominant effect in annular HTR physics. This analysis methodology should be further explored in order to reduce the error that is systematically propagated in the traditional generation of cross sections.

  16. Benchmark analysis of high temperature engineering test reactor core using McCARD code

    SciTech Connect (OSTI)

    Jeong, Chang Joon; Jo, Chang Keun; Lee, Hyun Chul; Noh, Jae Man [Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong, Daejeon, 305-353 (Korea, Republic of)

    2013-07-01T23:59:59.000Z

    A benchmark calculation has been performed for a startup core physics test of Japan's High Temperature Engineering Test Reactor (HTTR). The calculation is carried out by the McCARD code, which adopts the Monte Carlo method. The cross section library is ENDF-B/VII.0. The fuel cell is modeled by the reactivity-equivalent physical transform (RPT) method. Effective multiplication factors with different numbers of fuel columns have been analyzed. The calculation shows that the HTTR becomes critical with 19 fuel columns with an excess reactivity of 0.84% ?k/k. The discrepancies between the measurements and Monte Carlo calculations are 2.2 and 1.4 % ?k/k for 24 and 30 columns, respectively. The reasons for the discrepancy are thought to be the current version of cross section library and the impurity in the graphite which is represented by the boron concentration. In the future, the depletion results will be proposed for further benchmark calculations. (authors)

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

    Open Energy Info (EERE)

    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 Survey...

  18. Cotton Fertilizer Tests in the El Paso Area, 1943-51.

    E-Print Network [OSTI]

    Lyerly, Paul J.; Christensen, Paul D.

    1953-01-01T23:59:59.000Z

    Cotton Fertilizer Tests in the El Paso Area, 1943-51 CONTENTS Page Introduction ...................................................................................................... 3 Many Treatments on Field Plots Included... ............................................................... 8 ............................................................. Potassium Fertilizer Not Required 9 No Yield Increases from Trace Elements ............................................... ... 9...

  19. 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 [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein- Leopoldshafen (Germany)] [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein- Leopoldshafen (Germany); Klute, Stefan [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany)] [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany); Stanke, Dieter [Siempelkamp NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)] [Siempelkamp NIS Ingenieurgesellschaft mbH, Industriestrasse 13, 63755 Alzenau (Germany)

    2013-07-01T23:59:59.000Z

    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)

  20. 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

    2012-03-01T23:59:59.000Z

    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.

  1. Closure Report for Corrective Action Unit 425: Area 9 Main Lake Construction Debris Disposal Area, Tonopah Test Range, Nevada

    SciTech Connect (OSTI)

    K. B. Campbell

    2003-03-01T23:59:59.000Z

    Corrective Action Unit (CAU) 425 is located on the Tonopah Test Range, approximately 386 kilometers (240 miles) northwest of Las Vegas, Nevada. CAU 425 is listed in the Federal Facility Agreement and Consent Order (FFACO, 1996) and is comprised of one Corrective Action Site (CAS). CAS 09-08-001-TA09 consisted of a large pile of concrete rubble from the original Hard Target and construction debris associated with the Tornado Rocket Sled Tests. CAU 425 was closed in accordance with the FFACO and the Nevada Division of Environmental Protection-approved Streamlined Approach for Environmental Restoration Plan for CAU 425: Area 9 Main Lake Construction Debris Disposal Area, Tonopah Test Range, Nevada (U.S. Department of Energy, Nevada Operations Office, 2002). CAU 425 was closed by implementing the following corrective actions: The approved corrective action for this unit was clean closure. Closure activities included: (1) Removal of all the debris from the site. (2) Weighing each load of debris leaving the job site. (3) Transporting the debris to the U.S. Air Force Construction Landfill for disposal. (4) Placing the radioactive material in a U.S. Department of Transportation approved container for proper transport and disposal. (5) Transporting the radioactive material to the Nevada Test Site for disposal. (6) Regrading the job site to its approximate original contours/elevation.

  2. Monitoring and Control Research Using a University Reactor and SBWR Test-Loop

    SciTech Connect (OSTI)

    Robert M. Edwards

    2003-09-28T23:59:59.000Z

    The existing hybrid simulation capability of the Penn State Breazeale nuclear reactor was expanded to conduct research for monitoring, operations and control. Hybrid simulation in this context refers to the use of the physical time response of the research reactor as an input signal to a real-time simulation of power-reactor thermal-hydraulics which in-turn provides a feedback signal to the reactor through positioning of an experimental changeable reactivity device. An ECRD is an aluminum tube containing an absorber material that is positioned in the central themble of the reactor kinetics were used to expand the hybrid reactor simulation (HRS) capability to include out-of-phase stability characteristics observed in operating BWRs.

  3. 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

    2006-06-01T23:59:59.000Z

    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.

  4. 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)

    1993-09-01T23:59:59.000Z

    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.

  5. 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

    2011-03-01T23:59:59.000Z

    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.

  6. 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

    2012-02-01T23:59:59.000Z

    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).

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

    SciTech Connect (OSTI)

    Dan Ogden

    2014-10-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Noel Duckwitz

    2011-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Renae Soelberg

    2014-11-01T23:59:59.000Z

    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.

  12. Application of a Virtual Reactivity Feedback Control Loop in Non-Nuclear Testing of a Fast Spectrum Reactor

    SciTech Connect (OSTI)

    Bragg-Sitton, Shannon M. [NASA Marshall Space Flight Center, Huntsville, Al, 35812 (United States); Forsbacka, Matthew [NASA Headquarters, 300 E St. S, Washington, DC 20465 (United States)

    2004-07-01T23:59:59.000Z

    For a compact, fast-spectrum reactor, reactivity feedback is dominated by core deformation at elevated temperature. Given the use of accurate deformation measurement techniques, it is possible to simulate nuclear feedback in non-nuclear electrically heated reactor tests. Implementation of simulated reactivity feedback in response to measured deflection is being tested at the Nasa Marshall Space Flight Center Early Flight Fission Test Facility (EFF-TF). During tests of the SAFE-100 reactor prototype, core deflection was monitored using a high resolution camera. 'Virtual' reactivity feedback was accomplished by applying the results of Monte Carlo calculations (MCNPX) to core deflection measurements; the computational analysis was used to establish the reactivity worth of various core deformations. The power delivered to the SAFE-100 prototype was then adjusted accordingly via kinetics calculations. The work presented in this paper will demonstrate virtual reactivity feedback as core power was increased from 1 kWt to 10 kWt, held approximately constant at 10 kWt, and then allowed to decrease based on the negative thermal reactivity coefficient. (authors)

  13. 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

    2012-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    David W. Nigg; Devin A. Steuhm

    2011-09-01T23:59:59.000Z

    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.

  15. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    SciTech Connect (OSTI)

    Dautel, W.A.

    1996-10-01T23:59:59.000Z

    The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

  16. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    SciTech Connect (OSTI)

    Owen, M.B.

    1997-04-01T23:59:59.000Z

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

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

    SciTech Connect (OSTI)

    None

    1996-12-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    None

    2013-03-01T23:59:59.000Z

    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).

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

    SciTech Connect (OSTI)

    None

    2011-11-01T23:59:59.000Z

    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).

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

    SciTech Connect (OSTI)

    None

    2011-02-01T23:59:59.000Z

    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).

  1. Underground Test Area Subproject Phase I Data Analysis Task. Volume VIII - Risk Assessment Documentation Package

    SciTech Connect (OSTI)

    None

    1996-12-01T23:59:59.000Z

    Volume VIII 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 risk assessment 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. Use of the WECC WAMS in Wide Area Probing Tests for Validation of System Performance & Modeling

    SciTech Connect (OSTI)

    Hauer, John F.; Mittelstadt, William; Martin, Kenneth E.; Burns, J. W.; Lee, Harry; Pierre, John W.; Trudnowski, Daniel

    2009-02-01T23:59:59.000Z

    During 2005 and 2006 the Western Electricity Coordinating Council (WECC) performed three major tests of western system dynamics. These tests used a Wide Area Measurement System (WAMS) based primarily on Phasor Measurement Units (PMUs) to determine response to events including the insertion of the 1400-MW Chief Joseph braking resistor, probing signals, and ambient events. Test security was reinforced through real-time analysis of wide area effects, and high-quality data provided dynamic profiles for interarea modes across the entire western interconnection. The tests established that low-level optimized pseudo-random ±20-MW probing with the Pacific DC Intertie (PDCI) roughly doubles the apparent noise that is natural to the power system, providing sharp dynamic information with negligible interference to system operations. Such probing is an effective alternative to use of the 1400-MW Chief Joseph dynamic brake, and it is under consideration as a standard means for assessing dynamic security.

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

    SciTech Connect (OSTI)

    S. Blaine Grover

    2014-05-01T23:59:59.000Z

    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.

  4. In-Pile Experiment of a New Hafnium Aluminide Composite Material to Enable Fast Neutron Testing in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Donna Post Guillen; Douglas L. Porter; James R. Parry; Heng Ban

    2010-06-01T23:59:59.000Z

    A new hafnium aluminide composite material is being developed as a key component in a Boosted Fast Flux Loop (BFFL) system designed to provide fast neutron flux test capability in the Advanced Test Reactor. An absorber block comprised of hafnium aluminide (Al3Hf) particles (~23% by volume) dispersed in an aluminum matrix can absorb thermal neutrons and transfer heat from the experiment to pressurized water cooling channels. However, the thermophysical properties, such as thermal conductivity, of this material and the effect of irradiation are not known. This paper describes the design of an in-pile experiment to obtain such data to enable design and optimization of the BFFL neutron filter.

  5. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    SciTech Connect (OSTI)

    Not Available

    1992-07-01T23:59:59.000Z

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

  6. Regional groundwater flow and tritium transport modeling and risk assessment of the underground test area, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    None

    1997-10-01T23:59:59.000Z

    The groundwater flow system of the Nevada Test Site and surrounding region was evaluated to estimate the highest potential current and near-term risk to the public and the environment from groundwater contamination downgradient of the underground nuclear testing areas. The highest, or greatest, potential risk is estimated by assuming that several unusually rapid transport pathways as well as public and environmental exposures all occur simultaneously. These conservative assumptions may cause risks to be significantly overestimated. However, such a deliberate, conservative approach ensures that public health and environmental risks are not underestimated and allows prioritization of future work to minimize potential risks. Historical underground nuclear testing activities, particularly detonations near or below the water table, have contaminated groundwater near testing locations with radioactive and nonradioactive constituents. Tritium was selected as the contaminant of primary concern for this phase of the project because it is abundant, highly mobile, and represents the most significant contributor to the potential radiation dose to humans for the short term. It was also assumed that the predicted risk to human health and the environment from tritium exposure would reasonably represent the risk from other, less mobile radionuclides within the same time frame. Other contaminants will be investigated at a later date. Existing and newly collected hydrogeologic data were compiled for a large area of southern Nevada and California, encompassing the Nevada Test Site regional groundwater flow system. These data were used to develop numerical groundwater flow and tritium transport models for use in the prediction of tritium concentrations at hypothetical human and ecological receptor locations for a 200-year time frame. A numerical, steady-state regional groundwater flow model was developed to serve as the basis for the prediction of the movement of tritium from the underground testing areas on a regional scale. The groundwater flow model was used in conjunction with a particle-tracking code to define the pathlines followed by groundwater particles originating from 415 points associated with 253 nuclear test locations. Three of the most rapid pathlines were selected for transport simulations. These pathlines are associated with three nuclear test locations, each representing one of the three largest testing areas. These testing locations are: BOURBON on Yucca Flat, HOUSTON on Central Pahute Mesa, and TYBO on Western Pahute Mesa. One-dimensional stochastic tritium transport simulations were performed for the three pathlines using the Monte Carlo method with Latin hypercube sampling. For the BOURBON and TYBO pathlines, sources of tritium from other tests located along the same pathline were included in the simulations. Sensitivity analyses were also performed on the transport model to evaluate the uncertainties associated with the geologic model, the rates of groundwater flow, the tritium source, and the transport parameters. Tritium concentration predictions were found to be mostly sensitive to the regional geology in controlling the horizontal and vertical position of transport pathways. The simulated concentrations are also sensitive to matrix diffusion, an important mechanism governing the migration of tritium in fractured carbonate and volcanic rocks. Source term concentration uncertainty is most important near the test locations and decreases in importance as the travel distance increases. The uncertainty on groundwater flow rates is as important as that on matrix diffusion at downgradient locations. The risk assessment was performed to provide conservative and bounding estimates of the potential risks to human health and the environment from tritium in groundwater. Risk models were designed by coupling scenario-specific tritium intake with tritium dose models and cancer and genetic risk estimates using the Monte Carlo method. Estimated radiation doses received by individuals from chronic exposure to tritium, and the corre

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

    E-Print Network [OSTI]

    Manning, Sturt

    Private Water Well Testing in Areas Impacted by Marcellus Shale Gas Drilling (Updated November 15th in the absence of shale-gas drilling, well owners are strongly encouraged to evaluate their water on a regular review of shale gas drilling in New York State, as well as the most comprehensive collection of data

  8. 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 Y­sided detector for the readout, its fabrication is similar to a double­sided device, because the backside, which

  9. 105KE Basin Area Radiation Monitor System (ARMS) Acceptance Test Procedure

    SciTech Connect (OSTI)

    KINKEL, C.C.

    1999-12-14T23:59:59.000Z

    This procedure is intended for the Area Radiation Monitoring System, ARMS, that is replacing the existing Programmable Input-Output Processing System, PIOPS, radiation monitoring system in the 105KE basin. The new system will be referred to as the 105KE ARMS, 105KE Area Radiation Monitoring System. This ATP will ensure calibration integrity of the 105KE radiation detector loops. Also, this ATP will test and document the display, printing, alarm output, alarm acknowledgement, upscale check, and security functions. This ATP test is to be performed after completion of the 105KE ARMS installation. The alarm outputs of the 105KE ARMS will be connected to the basin detector alarms, basin annunciator system, and security Alarm Monitoring System, AMS, located in the 200 area Central Alarm Station (CAS).

  10. Validation Analysis of the Groundwater Flow and Transport Model of the Central Nevada Test Area

    SciTech Connect (OSTI)

    A. Hassan; J. Chapman; H. Bekhit; B. Lyles; K. Pohlmann

    2006-09-30T23:59:59.000Z

    The Central Nevada Test Area (CNTA) is a U.S. Department of Energy (DOE) site undergoing environmental restoration. The CNTA is located about 95 km northeast of Tonopah, Nevada, and 175 km southwest of Ely, Nevada (Figure 1.1). It was the site of the Faultless underground nuclear test conducted by the U.S. Atomic Energy Commission (DOE's predecessor agency) in January 1968. The purposes of this test were to gauge the seismic effects of a relatively large, high-yield detonation completed in Hot Creek Valley (outside the Nevada Test Site [NTS]) and to determine the suitability of the site for future large detonations. The yield of the Faultless underground nuclear test was between 200 kilotons and 1 megaton (DOE, 2000). A three-dimensional flow and transport model was created for the CNTA site (Pohlmann et al., 1999) and determined acceptable by DOE and the Nevada Division of Environmental Protection (NDEP) for predicting contaminant boundaries for the site.

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

    SciTech Connect (OSTI)

    Irene Farnham

    2011-05-01T23:59:59.000Z

    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.

  12. 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

    1985-01-01T23:59:59.000Z

    THE DEVELOPMENT AND OPERATIONAL TESTING OF AN EXPERIMENTAL REACTOR FOR GAS-LIQUID-SOLID REACTION SYSTEMS AT HIGH TEMPERATURES AND PRESSURES A Thesis by RICHARD KENNETH HESS Submitted to the Graduate College of Texas A&M University in partial... fulfillment of the requirements for the degree of MASTER OF SCIENCE August 1985 Major Subject: Chemical Engineering THE DEVELOPMENT AND OPERATIONAL TESTING OF AN EXPERIMENTAL REACTOR FOR GAS-LIQUID-SOLID REACTION SYSTEMS AT HIGH TEMPERATURES...

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

    SciTech Connect (OSTI)

    NONE

    1994-02-01T23:59:59.000Z

    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.

  14. 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

    2007-06-30T23:59:59.000Z

    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

  15. Preliminary Results of an On-Line, Multi-Spectrometer Fission Product Monitoring System to Support Advanced Gas Reactor Fuel Testing and Qualification in the Advanced Test Reactor at the Idaho National Laboratory

    SciTech Connect (OSTI)

    Dawn M. Scates; John K. Hartwell; John B. Walter; Mark W. Drigert

    2007-10-01T23:59:59.000Z

    The Advanced Gas Reactor -1 (AGR-1) experiment is the first experiment in a series of eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments scheduled for placement in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The experiment began irradiation in the ATR with a cycle that reached full power on December 26, 2006 and will continue irradiation for about 2.5 years. During this time six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The goals of the irradiation experiment is to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. This paper presents the preliminary test details of the fuel performance, as measured by the control and acquisition software.

  16. 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

    2012-10-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Vinson, D.

    2010-07-11T23:59:59.000Z

    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. PPPL-3164 -Preprint Date: January 1996, UC-420 Alpha Particle Losses from Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    , C. Z. Cheng, H. H. Duong, J. Fang§, N. J. Fisch, R. Fisher, E. D. Fredrickson, G. Y. Fu, R. F Reactor Deuterium-Tritium Plasmas D. S. Darrow, S. J. Zweben, S. Batha*, R. V. Budny, C. E. Bush, Z. Chang

  19. PPPL3164 Preprint Date: January 1996, UC420 Alpha Particle Losses from Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    . Chang, C. Z. Cheng, H. H. Duong # , J. Fang § , N. J. Fisch, R. Fisher # , E. D. Fredrickson, G. Y. Fu Reactor Deuterium­Tritium Plasmas D. S. Darrow, S. J. Zweben, S. Batha*, R. V. Budny, C. E. Bush + , Z

  20. 10 CFR 830 Major Modification Determination for Advanced Test Reactor RDAS and LPCIS Replacement

    SciTech Connect (OSTI)

    David E. Korns

    2012-05-01T23:59:59.000Z

    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.

  1. Results and phenomena observed from the DF-4 BWR (boiling water reactor) control blade-channel box test

    SciTech Connect (OSTI)

    Gauntt, R.O.; Gasser, R.D.; Fryer, C.P.; Walker, J.V.

    1988-01-01T23:59:59.000Z

    The DF-4 in-pile fuel damage experiment, one of a series of tests performed as part of the USNRC's internationally sponsored severe fuel damage (SFD) program, and carried out at Sandia National Laboratories, addressed the behavior of boiling water reactor (BWR) fuel canisters and control blades in the high temperature environment of an unrecovered loss of coolant accident. The DF-4 test is described in some detail herein and results from the experiment are presented. Significant results from prior experiments in the series are also summarized. Important findings from the DF-4 test include the low temperature melting of the stainless steel control blade caused by reaction with the B/sub 4/C, and the subsequent low temperature attack of the Zr-4 channel box by the relocating molten blade components. This early melt and relocation phenomena could significantly influence the progression of meltdown in BWRs and should be considered in SFD accident analyses. 14 refs., 9 figs.

  2. Re-evaluation of Savannah River reactor transient reactivity coefficient tests: The effect of delayed neutron constants and spatial variations

    SciTech Connect (OSTI)

    Burnett, T.W.T. [Westinghouse Electric Corp., Pittsburgh, PA (United States). Nuclear and Advanced Technology Div.; Graves, W.E. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1991-12-31T23:59:59.000Z

    Transient reactivity test conducted in one of the Savannah River production reactors in 1962 have been re-evaluated. A significantly lower (more negative) coolant temperature coefficient is now ascribed to that test; {minus}1 pcm/Deg-C vs the previously obtained +2 pcm/Deg-C. The change from the previous value is because of revisions to delayed neutron constants and accounting for spatial effects. The new value is in reasonable agreement with the currently calculated value of {minus}2 pcm/Deg-C, considering measurements and calculational uncertainties. Therefore, we conclude that the current analytic models for physics and transient analysis are fully consistent with the 1962 test observation, and that there is no basis for assigning a calculational bias or increasing uncertainty allowances. 2 refs.

  3. Re-evaluation of Savannah River reactor transient reactivity coefficient tests: The effect of delayed neutron constants and spatial variations

    SciTech Connect (OSTI)

    Burnett, T.W.T. (Westinghouse Electric Corp., Pittsburgh, PA (United States). Nuclear and Advanced Technology Div.); Graves, W.E. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1991-01-01T23:59:59.000Z

    Transient reactivity test conducted in one of the Savannah River production reactors in 1962 have been re-evaluated. A significantly lower (more negative) coolant temperature coefficient is now ascribed to that test; {minus}1 pcm/Deg-C vs the previously obtained +2 pcm/Deg-C. The change from the previous value is because of revisions to delayed neutron constants and accounting for spatial effects. The new value is in reasonable agreement with the currently calculated value of {minus}2 pcm/Deg-C, considering measurements and calculational uncertainties. Therefore, we conclude that the current analytic models for physics and transient analysis are fully consistent with the 1962 test observation, and that there is no basis for assigning a calculational bias or increasing uncertainty allowances. 2 refs.

  4. Treatability Test Plan for 300 Area Uranium Stabilization through Polyphosphate Injection

    SciTech Connect (OSTI)

    Vermeul, Vincent R.; Williams, Mark D.; Fritz, Brad G.; Mackley, Rob D.; Mendoza, Donaldo P.; Newcomer, Darrell R.; Rockhold, Mark L.; Williams, Bruce A.; Wellman, Dawn M.

    2007-06-01T23:59:59.000Z

    The U.S. Department of Energy has initiated a study into possible options for stabilizing uranium at the 300 Area using polyphosphate injection. As part of this effort, PNNL will perform bench- and field-scale treatability testing designed to evaluate the efficacy of using polyphosphate injections to reduced uranium concentrations in the groundwater to meet drinking water standards (30 ug/L) in situ. This technology works by forming phosphate minerals (autunite and apatite) in the aquifer that directly sequester the existing aqueous uranium in autunite minerals and precipitates apatite minerals for sorption and long term treatment of uranium migrating into the treatment zone, thus reducing current and future aqueous uranium concentrations. Polyphosphate injection was selected for testing based on technology screening as part of the 300-FF-5 Phase III Feasibility Study for treatment of uranium in the 300-Area.

  5. Closure Report for Corrective Action Unit 396: Area 20 Spill Sites, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

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

    2004-06-01T23:59:59.000Z

    Corrective Action Unit (CAU) 396, Area 20 Spill Sites, is located on the Nevada Test Site approximately 105 kilometers (65 miles) northwest of Las Vegas, Nevada. CAU 396 is listed in Appendix II of the Federal Facility Agreement and Consent Order of 1996 and consists of the following four Corrective Action Sites (CASs) located in Area 20 of the Nevada Test Site: CAS 20-25-01, Oil Spills (2); CAS 20-25-02, Oil Spills; CAS 20-25-03, Oil Spill; CAS 20-99-08, Spill. Closure activities for CAU 396 were conducted in accordance with the Federal Facility Agreement and Consent Order and the Nevada Division of Environmental Protection-approved Streamlined Approach for Environmental Restoration Plan for CAU 396.

  6. Estimation of measured control rod worths in Fast Breeder Test Reactor -- Effect of different delayed neutron parameters

    SciTech Connect (OSTI)

    Mohanakrishnan, P.; Reddy, C.P.; Gopalakrishnan, V.; Arul, J. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Reactor Physics Div.

    1996-03-01T23:59:59.000Z

    Control rod worths have been measured by the inverse kinetics method in the small PuC-UC core of the Fast Breeder Test Reactor at Kalpakkam. Delayed neutron fractional yields based on Tuttle`s data, ENDF/B-VI data, and the full summation approach of Brady and England have been used to get measured control rod worths. Unreasonably large reductions in control rod worths are obtained by the ENDF/B-VI data. It is suspected that the procedure, of normalizing fractional yields obtained by the summation approach to earlier evaluated total yields, is inconsistent.

  7. 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

    2012-03-02T23:59:59.000Z

    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.

  8. Corrective Action Decision Document/ Corrective Action Plan for Corrective Action Unit 443: Central Nevada Test Area-Subsurface Central Nevada Test Area, Nevada, Rev. No. 0

    SciTech Connect (OSTI)

    Susan Evans

    2004-11-01T23:59:59.000Z

    This Corrective Action Decision Document/Corrective Action Plan (CADD/CAP) has been prepared for the subsurface at the Central Nevada Test Area (CNTA) Corrective Action Unit (CAU) 443, CNTA - Subsurface, Nevada, in accordance with the ''Federal Facility Agreement and Consent Order'' (FFACO) (1996). CAU 443 is located in Hot Creek Valley in Nye County, Nevada, north of U.S. Highway 6, about 48 kilometers north of Warm Springs, Nevada. The CADD/CAP combines the decision document (CADD) with the corrective action plan (CAP) and provides or references the specific information necessary to recommend corrective actions for the UC-1 Cavity (Corrective Action Site 58-57-001) at CAU 443, as provided in the FFACO. The purpose of the CADD portion of the document (Section 1.0 to Section 4.0) is to identify and provide a rationale for the selection of a recommended corrective action alternative for the subsurface at CNTA. To achieve this, the following tasks were required: (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; and (5) Recommend a preferred corrective action alternative for the subsurface at CNTA. A Corrective Action Investigation (CAI) was performed in several stages from 1999 to 2003, as set forth in the ''Corrective Action Investigation Plan for the Central Nevada Test Area Subsurface Sites (Corrective Action Unit No. 443)'' (DOE/NV, 1999). Groundwater modeling was the primary activity of the CAI. Three phases of modeling were conducted for the Faultless underground nuclear test. The first involved the gathering and interpretation of geologic and hydrogeologic data into a three-dimensional numerical model of groundwater flow, and use of the output of the flow model for a transport model of radionuclide release and migration behavior (Pohlmann et al., 2000). The second modeling phase (known as a Data Decision Analysis [DDA]) occurred after the Nevada Division of Environmental Protection reviewed the first model and was designed to respond to concerns regarding model uncertainty (Pohll and Mihevc, 2000). The third modeling phase updated the original flow and transport model to incorporate the uncertainty identified in the DDA, and focused the model domain on the region of interest to the transport predictions. This third phase culminated in the calculation of contaminant boundaries for the site (Pohll et al., 2003).

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

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Blaine Grover

    2012-10-01T23:59:59.000Z

    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.

  11. 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))

    1990-01-01T23:59:59.000Z

    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.

  12. K Basin Sludge Conditioning Testing Nitric Acid Dissolution Testing of K East Area Sludge Composite, Small- and Large-Scale Testing

    SciTech Connect (OSTI)

    Carlson, C.D.; Delegard, C.H.; Burgeson, I.E.; Schmidt, A.J.; Silvers, K.L.

    1999-04-02T23:59:59.000Z

    This report describes work performed by Pacific Northwest National Laboratory (PNNL) for Numatec Hanford Corporation (NHC) to support the development of the K Basin Sludge Treatment System. For this work, testing was performed to examine the dissolution behavior of a K East Basin floor and Weasel Pit sludge composite, referred to as K East area sludge composite, in nitric acid at the following concentrations: 2 M, 4 M, 6 M and 7.8 M. With the exception of one high solids loading test the nitric acid was added at 4X the stoichiometric requirement (assuming 100% of the sludge was uranium metal). The dissolution tests were conducted at boiling temperatures for 24 hours. Most of the tests were conducted with {approximately}2.5 g of sludge (dry basis). The high solids loading test was conducted with {approximately}7 g of sludge. A large-scale dissolution test was conducted with 26.5 g of sludge and 620 mL of 6 M nitric acid. The objectives of this test were to (1) generate a sufficient quantity of acid-insoluble residual solids for use in leaching studies, and (2) examine the dissolution behavior of the sludge composite at a larger scale.

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

    SciTech Connect (OSTI)

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

    1993-02-01T23:59:59.000Z

    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.

  14. Advanced reactor safety research. Quarterly report, July-September 1981

    SciTech Connect (OSTI)

    Not Available

    1982-10-01T23:59:59.000Z

    Sandia National Laboratories, Albuquerque, New Mexico, is conducting the Advanced Reactor Safety Research Program on behalf of the US Nuclear Regulatory Commission (NRC). Sandia has been given the task to investigate seven major areas of interest which are intimately related to over-all NRC needs. These are: core debris behavior - inherent retention; containment analysis; elevated temperature design assessment; LMFBR accident delineation; advanced reactor core phenomenology; light water reactor (LWR) fuel damage phenomenology; and test and facility technology.

  15. Radiation monitoring around United States nuclear test areas, calendar year 1989

    SciTech Connect (OSTI)

    Not Available

    1990-05-01T23:59:59.000Z

    This report describes the Offsite Radiation Safety Program conducted during 1989 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 the testing is in compliance with existing radiation protection standards, and 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 both animals and humans. To implement protective actions, provide immediate radiation monitoring, and obtain environmental samples rapidly after any release of radioactivity, personnel with mobile monitoring equipment are placed in areas downwind from the test site prior to each test. 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 NTS activities. Trends were evaluated in the Noble Gas and Tritium, Milk Surveillance, TLD, and PIC networks, and the Long-Term Hydrological Monitoring Program. 35 refs., 68 figs., 32 tabs.

  16. 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.

    1997-08-01T23:59:59.000Z

    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.

  17. Flow Components in a NaK Test Loop Designed to Simulate Conditions in a Nuclear Surface Power Reactor

    SciTech Connect (OSTI)

    Polzin, Kurt A.; Godfroy, Thomas J. [NASA Marshall Space Flight Center Propulsion Research and Technology Applications Branch/ER24, MSFC, AL 35812 (United States)

    2008-01-21T23:59:59.000Z

    A test loop using NaK as the working fluid is presently in use to study material compatibility effects on various components that comprise a possible nuclear reactor design for use on the lunar surface. A DC electromagnetic (EM) pump has been designed and implemented as a means of actively controlling the NaK flow rate through the system and an EM flow sensor is employed to monitor the developed flow rate. These components allow for the matching of the flow rate conditions in test loops with those that would be found in a full-scale surface-power reactor. The design and operating characteristics of the EM pump and flow sensor are presented. In the EM pump, current is applied to a set of electrodes to produce a Lorentz body force in the fluid. A measurement of the induced voltage (back-EMF) in the flow sensor provides the means of monitoring flow rate. Both components are compact, employing high magnetic field strength neodymium magnets thermally coupled to a water-cooled housing. A vacuum gap limits the heat transferred from the high temperature NaK tube to the magnets and a magnetically-permeable material completes the magnetic circuit. The pump is designed to produce a pressure rise of 34.5 kPa, and the flow sensor's predicted output is roughly 20 mV at the loop's nominal flow rate of 0.114 m{sup 3}/hr.

  18. Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Tests 1 and 2

    SciTech Connect (OSTI)

    Russcher, G. E.; Wilson, C. L.; Marshall, R, K.; King, L. L.; Parchen, L. J.; Pilger, J. P.; Hesson, G. M.; Mohr, C. L.

    1981-09-01T23:59:59.000Z

    A loss of Coolant Accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects of LOCA conditions on pressurized water reactor test fuel bundles. This experiment operation plan for the second and third experiments of the program will provide peak fuel cladding temperatures of up to 1172K (1650{degree}F) and 1061K (1450{degree}) respectively. for a long enough time to cause test fuel cladding deformation and rupture in both. Reflood coolant delay times and the reflooding rates for the experiments were selected from thermal-hydraulic data measured in the National Research Universal (NRU) reactor facilities and test train assembly during the first experiment.

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

    SciTech Connect (OSTI)

    K. B. Campbell

    2003-04-01T23:59:59.000Z

    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.

  20. Fabrication and Testing of Full-Length Single-Cell Externally Fueled Converters for Thermionic Reactors

    SciTech Connect (OSTI)

    Schock, Alfred

    1995-08-01T23:59:59.000Z

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes the fabrication and testing of full-length prototypcial converters, both unfueled and fueled, and presents parametric results of electrically heated tests.

  1. U.S. Plans for the Next Fast Reactor Transmutation Fuels Irradiation Test

    SciTech Connect (OSTI)

    B. A. Hilton

    2007-09-01T23:59:59.000Z

    The U.S. Advanced Fuel Cycle Initiative (AFCI) 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 dramatically decreasing the volume of material requiring disposal and the long-term radio-toxicity and heat load of high-level waste sent to a geologic repository. One important component of the technology development is actinide-bearing transmutation fuel forms containing plutonium, neptunium, americium (and possibly curium) isotopes. Metallic alloy and oxide fuel forms are being developed as the near term options for fast reactor implementation.

  2. 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.

    1992-01-01T23:59:59.000Z

    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.

  3. 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

    2004-06-24T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2008-09-01T23:59:59.000Z

    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.

  5. Corrective action investigation plan for CAU No. 424: Area 3 Landfill Complex, Tonopah Test Range, Nevada

    SciTech Connect (OSTI)

    NONE

    1997-04-01T23:59:59.000Z

    This Correction Action Investigation Plan contains the environmental sample collection objectives and the criteria for conducting site investigation activities at the Area 3 Landfill Complex, CAU No. 424, which is located at the Tonopah Test Range (TTR). The TTR, included in the Nellis Air Force Range, is approximately 255 kilometers (140 miles) northwest of Las Vegas, nevada. The CAU 424 is comprised of eight individual landfill sites that are located around and within the perimeter of the Area 3 Compound. Due to the unregulated disposal activities commonly associated with early landfill operations, an investigation will be conducted at each CAS to complete the following tasks: identify the presence and nature of possible contaminant migration from the landfills; determine the vertical and lateral extent of possible contaminant migration; ascertain the potential impact to human health and the environment; and provide sufficient information and data to develop and evaluate appropriate corrective action strategies for each CAS.

  6. Closure Report for Corrective Action Unit 261: Area 25 Test Cell A Leachfield System, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    T. M. Fitzmaurice

    2001-04-01T23:59:59.000Z

    The purpose of this Closure Report (CR) is to provide documentation of the completed corrective action at the Test Cell A Leachfield System and to provide data confirming the corrective action. The Test Cell A Leachfield System is identified in the Federal Facility Agreement and Consent Order (FFACO) of 1996 as Corrective Action Unit (CAU) 261. Remediation of CAU 261 is required under the FFACO (1996). CAU 261 is located in Area 25 of the Nevada Test Site (NTS) which is approximately 140 kilometers (87 miles) northwest of Las Vegas, Nevada (Figure 1). CAU 261 consists of two Corrective Action Sites (CASS): CAS 25-05-01, Leachfield; and CAS 25-05-07, Acid Waste Leach Pit (AWLP) (Figures 2 and 3). Test Cell A was operated during the 1960s and 1970s to support the Nuclear Rocket Development Station. Various operations within Building 3124 at Test Cell A resulted in liquid waste releases to the Leachfield and the AWLP. The following existing site conditions were reported in the Corrective Action Decision Document (CADD) (U.S. Department of Energy, Nevada Operations Office [DOE/NV], 1999): Soil in the leachfield was found to exceed the Nevada Division of Environmental Protection (NDEP) Action Level for petroleum hydrocarbons, the U.S. Environmental Protection Agency (EPA) preliminary remediation goals for semi volatile organic compounds, and background concentrations for strontium-90; Soil below the sewer pipe and approximately 4.5 meters (m) (15 feet [ft]) downstream of the initial outfall was found to exceed background concentrations for cesium-137 and strontium-90; Sludge in the leachfield septic tank was found to exceed the NDEP Action Level for petroleum hydrocarbons and to contain americium-241, cesium-137, uranium-234, uranium-238, potassium-40, and strontium-90; No constituents of concern (COC) were identified at the AWLP. The NDEP-approved CADD (DOWNV, 1999) recommended Corrective Action Alternative 2, ''Closure of the Septic Tank and Distribution Box, Partial Excavation, and Administrative Controls.'' The corrective action was performed following the NDEP-approved Corrective Action Plan (CAP) (DOE/NV, 2000).

  7. Neutronics, steady-state, and transient analyses for the Poland MARIA reactor for irradiation testing of LEU lead test fuel assemblies from CERCA : ANL independent verification results.

    SciTech Connect (OSTI)

    Garner, P. L.; Hanan, N. A. (Nuclear Engineering Division)

    2011-06-07T23:59:59.000Z

    The MARIA reactor at the Institute of Atomic Energy (IAE) in Swierk (30 km SE of Warsaw) in the Republic of Poland is considering conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel assemblies (FA). The FA design in MARIA is rather unique; a suitable LEU FA has never been designed or tested. IAE has contracted with CERCA (the fuel supply portion of AREVA in France) to supply 2 lead test assemblies (LTA). The LTAs will be irradiated in MARIA to burnup level of at least 40% for both LTAs and to 60% for one LTA. IAE may decide to purchase additional LEU FAs for a full core conversion after the test irradiation. The Reactor Safety Committee within IAE and the National Atomic Energy Agency in Poland (PAA) must approve the LTA irradiation process. The approval will be based, in part, on IAE submitting revisions to portions of the Safety Analysis Report (SAR) which are affected by the insertion of the LTAs. (A similar process will be required for the full core conversion to LEU fuel.) The analysis required was established during working meetings between Argonne National Laboratory (ANL) and IAE staff during August 2006, subsequent email correspondence, and subsequent staff visits. The analysis needs to consider the current high-enriched uranium (HEU) core and 4 core configurations containing 1 and 2 LEU LTAs in various core positions. Calculations have been performed at ANL in support of the LTA irradiation. These calculations are summarized in this report and include criticality, burn-up, neutronics parameters, steady-state thermal hydraulics, and postulated transients. These calculations have been performed at the request of the IAE staff, who are performing similar calculations to be used in their SAR amendment submittal to the PAA. The ANL analysis has been performed independently from that being performed by IAE and should only be used as one step in the verification process.

  8. Closure report for CAU 93: Area 6 steam cleaning effluent ponds, Nevada Test Site. Volume 1

    SciTech Connect (OSTI)

    NONE

    1997-12-01T23:59:59.000Z

    The Steam Cleaning Effluent Ponds (SCEP) waste unit is located in Area 6 at the Nevada Test Site. The SCEPs are evaporation basins formerly used for the disposal of untreated liquid effluent discharged from steam cleaning activities associated with Buildings 6-623 and 6-800. This closure report documents the strategy and analytical results that support the clean closure or closure in place of each of the components within CAU 93. In addition, the report documents all deviations from the approved closure plan and provides rationale for all deviations.

  9. The large-area hybrid-optics CLAS12 RICH detector: Tests of innovative components

    SciTech Connect (OSTI)

    Contalbrigo, M.; Baltzell, N.; Benmokhtar, F.; Barion, L.; Cisbani, E.; El Alaoui, A.; Hafidi, K.; Hoek, M.; Kubarovsky, V.; Lagamba, L.; Lucherini, V.; Malaguti, R.; Mirazita, M.; Montgomery, R.; Movsisyan, A.; Musico, P.; Orecchini, D.; Orlandi, A.; Pappalardo, L.L.; Pereira, S.; Perrino, R.; Phillips, J.; Pisano, S.; Rossi, P.; Squerzanti, S.; Tomassini, S.; Turisini, M.; Viticchič, A.

    2014-07-01T23:59:59.000Z

    A large area ring-imaging Cherenkov detector has been designed to provide clean hadron identification capability in the momentum range from 3 GeV/c to 8 GeV/c for the CLAS12 experiments at the upgraded 12 GeV continuous electron beam accelerator facility of Jefferson Lab to study the 3D nucleon structure in the yet poorly explored valence region by deep-inelastic scattering, and to perform precision measurements in hadronization and hadron spectroscopy. The adopted solution foresees a novel hybrid optics design based on an aerogel radiator, composite mirrors and densely packed and highly segmented photon detectors. Cherenkov light will either be imaged directly (forward tracks) or after two mirror reflections (large angle tracks). The preliminary results of individual detector component tests and of the prototype performance at test-beams are reported here.

  10. 100-D Area In Situ Redox Treatability Test for Chromate-Contaminated Groundwater

    SciTech Connect (OSTI)

    Williams, Mark D.; Vermeul, Vincent R.; Szecsody, James E.; Fruchter, Jonathan S.

    2000-10-12T23:59:59.000Z

    A treatability test was conducted for the In Situ Redox Manipulation (ISRM) technology at the 100 D Area of the U. S. Department of Energy's Hanford Site. The target contaminant was dissolved chromate in groundwater. The ISRM technology creates a permeable subsurface treatment zone to reduce mobile chromate in groundwater to an insoluble form. The ISRM permeable treatment zone is created by reducing ferric iron to ferrous iron within the aquifer sediments, which is accomplished by injecting aqueous sodium dithionite into the aquifer and then withdrawing the reaction products. The goal of the treatability test was to create a linear ISRM barrier by injecting sodium dithionite into five wells. Well installation and site characterization activities began in spring 1997; the first dithionite injection took place in September 1997. The results of this first injection were monitored through the spring of 1998. The remaining four dithionite injections were carried out in May through July of 1998.These five injections created a reduced zone in the Hanford unconfined aquifer approximately 150 feet in length (perpendicular to groundwater flow) and 50 feet wide. The reduced zone extended over the thickness of the unconfined zone. Analysis of post-emplacement groundwater samples showed concentrations of chromate, in the reduced zone decreased from approximately 1.0 mg/L before the tests to below analytical detection limits (<0.007 mg/L). Chromate concentrations also declined in downgradient monitoring wells to as low as 0.020 mg/L. These data, in addition to results from pre-test reducible iron characterization, indicate the barrier should be effective for 20 to 25 years. The 100-D Area ISRM barrier is being expanded to a length of up to 2,300 ft to capture a larger portion of the chromate plume.

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

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2008-01-01T23:59:59.000Z

    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.

  12. Underground Test Area Project Waste Management Plan (Rev. No. 2, April 2002)

    SciTech Connect (OSTI)

    IT Corporation, Las Vegas

    2002-04-24T23:59:59.000Z

    The U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Operations Office (NNSA/NV) initiated the UGTA Project to characterize the risk posed to human health and the environment as a result of underground nuclear testing activities at the Nevada Test Site (NTS). The UGTA Project investigation sites have been grouped into Corrective Action Units (CAUs) in accordance with the most recent version of the Federal Facility Agreement and Consent Order. The primary UGTA objective is to gather data to characterize the groundwater aquifers beneath the NTS and adjacent lands. The investigations proposed under the UGTA program may involve the drilling and sampling of new wells; recompletion, monitoring, and sampling of existing wells; well development and hydrologic/ aquifer testing; geophysical surveys; and subsidence crater recharge evaluation. Those wastes generated as a result of these activities will be managed in accordance with existing federal and state regulations, DOE Orders, and NNSA/NV waste minimization and pollution prevention objectives. This Waste Management Plan provides a general framework for all Underground Test Area (UGTA) Project participants to follow for the characterization, storage/accumulation, treatment, and disposal of wastes generated by UGTA Project activities. The objective of this waste management plan is to provide guidelines to minimize waste generation and to properly manage wastes that are produced. Attachment 1 to this plan is the Fluid Management Plan and details specific strategies for management of fluids produced under UGTA operations.

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

    SciTech Connect (OSTI)

    Laurie, M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Fourrez, S. [THERMOCOAX SAS, BP 26, Planquivon, F-61438 Flers Cedex (France); Fuetterer, M. A.; Lapetite, J. M. [European Commission, Joint Research Centre, Inst. for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands)

    2011-07-01T23:59:59.000Z

    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)

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

    E-Print Network [OSTI]

    Alexander I. Shlyakhter

    2003-08-06T23:59:59.000Z

    [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 .

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

    SciTech Connect (OSTI)

    D. H. Cox

    2001-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Peterson, Per

    2012-10-30T23:59:59.000Z

    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.

  17. New Sensors for In-Pile Temperature Measurement at the Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

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

    2011-09-01T23:59:59.000Z

    The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) a National Scientific User Facility (NSUF) in April 2007 to support U.S. research in nuclear science and technology. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation’s energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

  18. New Sensors for In-Pile Temperature Detection at the Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    J. L. Rempe; D. L. Knudson; J. E. Daw; K. G. Condie; S. Curtis Wilkins

    2009-09-01T23:59:59.000Z

    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. As a user facility, the ATR is supporting new users from universities, laboratories, and industry, as they conduct basic and applied nuclear research and development to advance the nation’s energy security needs. A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing measurements of key parameters during irradiation. This paper describes the strategy for determining what instrumentation is needed and the program for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available and under development for in-pile detection of temperature at various irradiation locations in the ATR.

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

    SciTech Connect (OSTI)

    Lisa Harvego; Brion Bennett

    2011-11-01T23:59:59.000Z

    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.

  20. 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)

    2014-05-15T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    PETERSEN SW

    2009-09-24T23:59:59.000Z

    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

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

    SciTech Connect (OSTI)

    Robinson, Adam B; Wachs, Daniel M

    2010-03-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    U.S. Department of Energy Nevada Operations Office

    1999-04-02T23:59:59.000Z

    This Corrective Action Decision Document (CADD) identifies and rationalizes the U.S. Department of Energy, Nevada Operations Office's selection of a recommended corrective action alternative (CAA) appropriate to facilitate the closure of Corrective Action Unit (CAU) 417: Central Nevada Test Area Surface, Nevada, under the Federal Facility Agreement and Consent Order. Located in Hot Creek Valley in Nye County, Nevada, and consisting of three separate land withdrawal areas (UC-1, UC-3, and UC-4), CAU 417 is comprised of 34 corrective action sites (CASs) including 2 underground storage tanks, 5 septic systems, 8 shaker pad/cuttings disposal areas, 1 decontamination facility pit, 1 burn area, 1 scrap/trash dump, 1 outlier area, 8 housekeeping sites, and 16 mud pits. Four field events were conducted between September 1996 and June 1998 to complete a corrective action investigation indicating that the only contaminant of concern was total petroleum hydrocarbon (TPH) which was found in 18 of the CASs. A total of 1,028 samples were analyzed. During this investigation, a statistical approach was used to determine which depth intervals or layers inside individual mud pits and shaker pad areas were above the State action levels for the TPH. Other related field sampling activities (i.e., expedited site characterization methods, surface geophysical surveys, direct-push geophysical surveys, direct-push soil sampling, and rotosonic drilling located septic leachfields) were conducted in this four-phase investigation; however, no further contaminants of concern (COCs) were identified. During and after the investigation activities, several of the sites which had surface debris but no COCs were cleaned up as housekeeping sites, two septic tanks were closed in place, and two underground storage tanks were removed. The focus of this CADD was to identify CAAs which would promote the prevention or mitigation of human exposure to surface and subsurface soils with contaminant concentrations above preliminary action levels. Based on the potential exposure pathways, several risk-based CAAs were developed and evaluated against the individual CAS requirements. It was determined that a combination of the CAAs would be recommended to meet all applicable state and federal regulations for closure of these sites and to eliminate potential future exposure pathways to the TPH-contaminated soils.

  4. Hanford Immobilized LAW Product Acceptance: Initial Tanks Focus Area Testing Data Package

    SciTech Connect (OSTI)

    Vienna, John D.; Jiricka, Antonin; McGrail, B. Peter; Jorgensen, Benaiah M.; Smith, Donald E.; Allen, Benjamin R.; Marra, James C.; Peeler, David K.; Brown, Kevin G.; Reamer, I. A.; Ebert, W. L.

    2000-02-08T23:59:59.000Z

    A matrix of 55 glasses was developed and tested with the aim to identify the impact of glass composition on the long-term corrosion behavior and to develop an acceptable low-activity waste glass composition region. Of the 55 glasses, 45 were designed to systematically vary the glass composition and 10 were selected because large and growing databases on their corrosion characteristics had accumulated. The performance of these 55 glasses in the vapor-phase hydration test (VHT) and product consistency test (PCT) were characterized. VHT's were performed at temperatures between 150?C and 300?C for times up to 280 days; preliminary corrosion rates and type of alteration products were identified. PCTs were performed at 90?C with glass surface area's to solution volumes (S/V) of 2000 m-1 for 7 days and S/V of 20 000 m-1 for 10 h, 100 h, and 1000 h. The corrosion extents by PCT were determined as functions of time from solution composition analyses.

  5. 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.

    1999-01-01T23:59:59.000Z

    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.

  6. 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. [and others

    1996-06-01T23:59:59.000Z

    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.

  7. Hydrologic Resources Management Program and Underground Test Area Project FY 2000 Progress Report

    SciTech Connect (OSTI)

    Davisson, M L; Eaton, G F; Hakemi, N L; Hudson, G B; Hutcheon, I D; Lau, C A; Kersting, A B; Kenneally, J M; Moran, J E; Phinney, D L; Rose, T P; Smith, D K; Sylwester, E R; Wang, L; Williams, R; Zavarin, M

    2001-07-01T23:59:59.000Z

    This report highlights the results of FY 2000 technical studies conducted by the Analytical and Nuclear Chemistry Division (ANCD) at Lawrence Livermore National Laboratory (LLNL) in support of the Hydrology and Radionuclide Migration Program (HRMP) and Underground Test Area (UGTA) Project. This is the latest in a series of annual reports published by LLNL-ANCD to document recent investigations of radionuclide migration and transport processes at the Nevada Test Site (NTS). The HRMP is sponsored by Defense Programs (DP) at the U.S. Department of Energy, Nevada Operations Office (DOENV), and supports DP operations at the NTS through studies of radiochemical and hydrologic processes that are relevant to the DP mission. Other organizations that support the HRMP include Los Alamos National Laboratory (LANL), the U.S. Geological Survey (USGS), the Desert Research Institute (DRI) of the University of Nevada, the U.S. Environmental Protection Agency (EPS), and Bechtel Nevada (BN). The UGTA Project is sponsored by the Environmental Management (EM) program at DOENV; its goal is to determine the extent of radionuclide contamination in groundwater resulting from underground nuclear testing at the NTS. The project strategy follows guidelines set forth in a Federal Facilities Agreement and Consent Order between the U.S. Department of Energy, the U.S. Department of Defense, and the State of Nevada. Participating contractors include LLNL (both ANCD and the Energy and Environmental Sciences Directorate), LANL, USGS, DRI, BN, and IT Corporation (with subcontract support from Geotrans Inc.).

  8. 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

    2008-04-08T23:59:59.000Z

    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.

  9. A Validation Process for the Groundwater Flow and Transport Model of the Faultless Nuclear Test at Central Nevada Test Area

    SciTech Connect (OSTI)

    Ahmed Hassan

    2003-01-01T23:59:59.000Z

    Many sites of groundwater contamination rely heavily on complex numerical models of flow and transport to develop closure plans. This has created a need for tools and approaches that can be used to build confidence in model predictions and make it apparent to regulators, policy makers, and the public that these models are sufficient for decision making. This confidence building is a long-term iterative process and it is this process that should be termed ''model validation.'' Model validation is a process not an end result. That is, the process of model validation cannot always assure acceptable prediction or quality of the model. Rather, it provides safeguard against faulty models or inadequately developed and tested models. Therefore, development of a systematic approach for evaluating and validating subsurface predictive models and guiding field activities for data collection and long-term monitoring is strongly needed. This report presents a review of model validation studies that pertain to groundwater flow and transport modeling. Definitions, literature debates, previously proposed validation strategies, and conferences and symposia that focused on subsurface model validation are reviewed and discussed. The review is general in nature, but the focus of the discussion is on site-specific, predictive groundwater models that are used for making decisions regarding remediation activities and site closure. An attempt is made to compile most of the published studies on groundwater model validation and assemble what has been proposed or used for validating subsurface models. The aim is to provide a reasonable starting point to aid the development of the validation plan for the groundwater flow and transport model of the Faultless nuclear test conducted at the Central Nevada Test Area (CNTA). The review of previous studies on model validation shows that there does not exist a set of specific procedures and tests that can be easily adapted and applied to determine the validity of site-specific groundwater models. This is true for both deterministic and stochastic models, with the latter posing a more difficult and challenging problem when it comes to validation. This report then proposes a general validation approach for the CNTA model, which addresses some of the important issues recognized in previous validation studies, conferences, and symposia as crucial to the process. The proposed approach links model building, model calibration, model predictions, data collection, model evaluations, and model validation in an iterative loop. The approach focuses on use of collected validation data to reduce model uncertainty and narrow the range of possible outcomes of stochastic numerical models. It accounts for the stochastic nature of the numerical CNTA model, which used Monte Carlo simulation approach. The proposed methodology relies on the premise that absolute validity is not even a theoretical possibility and is not a regulatory requirement. Rather, it highlights the importance of testing as many aspects of the model as possible and using as many diverse statistical tools as possible for rigorous checking and confidence building in the model and its predictions. It is this confidence that will eventually allow for regulator and public acceptance of decisions based on the model predictions.

  10. 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

    2013-12-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1985-09-01T23:59:59.000Z

    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.

  12. Lessons Learned During Construction and Test of the GLAST Large Area Telescope Tracker

    SciTech Connect (OSTI)

    Latronico, L.; /INFN, Pisa; ,

    2005-08-09T23:59:59.000Z

    The GLAST Large Area Telescope (LAT) is a satellite gamma-ray observatory designed to explore the sky in the energy range 20MeV {approx_equal} 300GeV, a region populated by emissions from the most energetic and mysterious objects in the cosmos, like black holes, AGNs, supernovae, gamma-ray bursters. The silicon-strip tracker is the heart of the photon detection system, and with its 80 m{sup 2} of surface and almost 1M channels is one of the largest silicon tracker ever built. Its construction, to be completed by 2006, and the stringent requirements from operation in space, represent a major technological challenge. Critical design, technology and system engineering issues are addressed in this paper, as well as the approach being followed during construction, test and qualification of the LAT silicon tracker.

  13. 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

    2012-06-01T23:59:59.000Z

    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.

  14. Characteristics of radionuclide-contaminated soils from the Sedan crater area at the Nevada test site

    SciTech Connect (OSTI)

    Lee, S.Y.; Tamura, T.; Larsen, I.L.; Essington, E.H.

    1987-08-01T23:59:59.000Z

    The distribution of radionuclides in selected soil profiles and the characteristics of contaminated ejecta in the Sedan crater area (site of a nuclear excavation test conducted in 19862) were investigated in an attempt to define the physicochemical parameters controlling the redistribution of radionuclides required to assess the environmental and health hazards of the contaminated area. The results showed that concentrations of /sup 239,240/Pu, /sup 241/Am, /sup 155/Eu, /sup 137/Cs, /sup 102m/Rh, /sup 90/Sr, and /sup 60/Co decreased with depth and with increasing distance from ground zero. The contaminated particles occurred as moderately porous calcium aluminosilicate glass, predominantly and size (2-0.1 mm in diameter), in the loamy sand or sandy loam soil. Differences in distribution patterns and concentration ratios among the radionuclides indicted a that a significant fraction of the /sup 137/Cs and /sup 90/Sr present in contaminated overburden materials had moved downward during the past 22 years, under arid environmental conditions, into the buried soil.

  15. 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.

    2010-08-31T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2007-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1994-05-01T23:59:59.000Z

    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.

  18. Experiment Operations Plan for a Loss-of-Coolant Accident Simulation in the National Research Universal Reactor Materials Test 2

    SciTech Connect (OSTI)

    Russcher, G. E.; Barner, J. O.; Hesson, G. M.; Wilson, C. L.; Parchen, L. J.; Cunningham, M. E.; Marshall, R. K.; Mohr, C. L.

    1981-09-01T23:59:59.000Z

    A loss-of-coolant accident (LOCA) simulation program is evaluating the thermal-hydraulic and mechanical effects on pressurized water reactor (PWR) test fuel bundles. This Experiment Operation Plan (EOP) Addendum 2, together with the referenced EOP, describes the desired operating conditions and additional hazards review associated with the four-part MT-2 experiment. The primary portions of the experiment, MT-2.2 and MT-2.3, will evaluate the following: 1) the mechanical deformation of pressurized fuel rods subjected to a slow LOCA, using reflood water for temperature control, that is designed to produce cladding temperatures in the range from 1033 to 1089K (1400 to 1500°F) for an extended time, and 2) the effects of the deformed and possibly failed cladding on the thermal-hydraulic performance of the test assembly during simulated LOCA heating and reflooding. The secondary portions of the experiment, MT-2.1 and MT-2.4, are intended to provide thermal-hydraulic calibration information during two-stage reflood conditions for 1) relatively low cladding temperatures, <839K (1050°F), on nondeformed rods, and 2) moderately high cladding temperatures, <1089K (1500°F), on deformed rods.

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

    SciTech Connect (OSTI)

    Bechtel Nevada Geotechnical Sciences

    2005-06-01T23:59:59.000Z

    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.

  20. Plasma Phys. Control. Fusion 39 (1997) A275A283. Printed in the UK PII: S0741-3335(97)81172-4 Alpha-particle physics in the tokamak fusion test reactor

    E-Print Network [OSTI]

    Plasma Phys. Control. Fusion 39 (1997) A275­A283. Printed in the UK PII: S0741-3335(97)81172-4 Alpha-particle physics in the tokamak fusion test reactor DT experiment S J Zwebena , V Arunasalama fusion test reactor. Alpha particles are generally well confined in MHD-quiescent discharges, and alpha

  1. 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

    2006-09-01T23:59:59.000Z

    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.

  2. TESTING GROUND BASED GEOPHYSICAL TECHNIQUES TO REFINE ELECTROMAGNETIC SURVEYS NORTH OF THE 300 AREA HANFORD WASHINGTON

    SciTech Connect (OSTI)

    PETERSEN SW

    2010-12-02T23:59:59.000Z

    Airborne electromagnetic (AEM) surveys were flown during fiscal year (FY) 2008 within the 600 Area in an attempt to characterize the underlying subsurface and to aid in the closure and remediation design study goals for the 200-PO-1 Groundwater Operable Unit (OU). The rationale for using the AEM surveys was that airborne surveys can cover large areas rapidly at relatively low costs with minimal cultural impact, and observed geo-electrical anomalies could be correlated with important subsurface geologic and hydrogeologic features. Initial interpretation of the AEM surveys indicated a tenuous correlation with the underlying geology, from which several anomalous zones likely associated with channels/erosional features incised into the Ringold units were identified near the River Corridor. Preliminary modeling resulted in a slightly improved correlation but revealed that more information was required to constrain the modeling (SGW-39674, Airborne Electromagnetic Survey Report, 200-PO-1 Groundwater Operable Unit, 600 Area, Hanford Site). Both time-and frequency domain AEM surveys were collected with the densest coverage occurring adjacent to the Columbia River Corridor. Time domain surveys targeted deeper subsurface features (e.g., top-of-basalt) and were acquired using the HeliGEOTEM{reg_sign} system along north-south flight lines with a nominal 400 m (1,312 ft) spacing. The frequency domain RESOLVE system acquired electromagnetic (EM) data along tighter spaced (100 m [328 ft] and 200 m [656 ft]) north-south profiles in the eastern fifth of the 200-PO-1 Groundwater OU (immediately adjacent to the River Corridor). The overall goal of this study is to provide further quantification of the AEM survey results, using ground based geophysical methods, and to link results to the underlying geology and/or hydrogeology. Specific goals of this project are as follows: (1) Test ground based geophysical techniques for the efficacy in delineating underlying geology; (2) Use ground measurements to refine interpretations of AEM data; and (3) Improve the calibration and correlation of AEM information. The potential benefits of this project are as follows: (1) Develop a tool to map subsurface units at the Hanford Site in a rapid and cost effective manner; (2) Map groundwater pathways within the River Corridor; and (3) Aid development of the conceptual site model. If anomalies observed in the AEM data can be correlated with subsurface geology, then the rapid scanning and non-intrusive capabilities provided by the airborne surveys can be used at the Hanford Site to screen for areas that warrant further investigation.

  3. 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

    2000-07-20T23:59:59.000Z

    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.

  4. 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

    2005-12-01T23:59:59.000Z

    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.

  5. New Sensors for the Advanced Test Reactor National Scientific User Facility

    SciTech Connect (OSTI)

    Joy L. Rempe; Darrell L. Knudson; Keith G. Condie; Joshua E. Daw; Heng Ban; Brandon Fox; Gordon Kohse

    2009-06-01T23:59:59.000Z

    A key component of the ATR NSUF effort is to develop and evaluate new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. This paper describes the selection strategy of what instrumentation is needed, and the program generated for developing new or enhanced sensors that can address these needs. Accomplishments from this program are illustrated by describing new sensors now available to users of the ATR NSUF with data from irradiation tests using these sensors. In addition, progress is reported on current research efforts to provide users advanced methods for detecting temperature, fuel thermal conductivity, and changes in sample geometry.

  6. Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 425: Area 9 Main Lake Construction Debris Disposal Area, Tonopah Test Range, Nevada

    SciTech Connect (OSTI)

    K. B. Campbell

    2002-04-01T23:59:59.000Z

    This Streamlined Approach for Environmental Restoration (SAFER) Plan addresses the action necessary for the closure of Corrective Action Unit (CAU) 425, Area 9 Main Lake Construction Debris Disposal Area. This CAU is currently listed in Appendix III of the Federal Facility Agreement and Consent Order (FFACO, 1996). This site will be cleaned up under the SAFER process since the volume of waste exceeds the 23 cubic meters (m{sup 3}) (30 cubic yards [yd{sup 3}]) limit established for housekeeping sites. CAU 425 is located on the Tonopah Test Range (TTR) and consists of one Corrective Action Site (CAS) 09-08-001-TA09, Construction Debris Disposal Area (Figure 1). CAS 09-08-001-TA09 is an area that was used to collect debris from various projects in and around Area 9. The site is located approximately 81 meters (m) (265 feet [ft]) north of Edwards Freeway northeast of Main Lake on the TTR. The site is composed of concrete slabs with metal infrastructure, metal rebar, wooden telephone poles, and concrete rubble from the Hard Target and early Tornado Rocket sled tests. Other items such as wood scraps, plastic pipes, soil, and miscellaneous nonhazardous items have also been identified in the debris pile. It is estimated that this site contains approximately 2280 m{sup 3} (3000 yd{sup 3}) of construction-related debris.

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

    SciTech Connect (OSTI)

    K. Campbell

    2000-04-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2008-04-01T23:59:59.000Z

    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.

  9. 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)

    2002-01-09T23:59:59.000Z

    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.

  10. Development and experimental validation of a calculation scheme for nuclear heating evaluation in the core of the OSIRIS material testing reactor

    SciTech Connect (OSTI)

    Malouch, F. [Saclay Center CEA, DEN/DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette Cedex (France)

    2011-07-01T23:59:59.000Z

    The control of the temperature in material samples irradiated in a material testing reactor requires the knowledge of the nuclear heating caused by the energy deposition by neutrons and photons interacting in the irradiation device structures. Thus, a neutron-photonic three-dimensional calculation scheme has been developed to evaluate the nuclear heating in experimental devices irradiated in the core of the OSIRIS MTR reactor (CEA/Saclay Center). The aim is to obtain a predictive tool for the nuclear heating estimation in irradiation devices. This calculation scheme is mainly based on the TRIPOLI-4 three-dimensional continuous-energy Monte Carlo transport code, developed by CEA (Saclay Center). An experimental validation has been carried out on the basis of nuclear heating measurements performed in the OSIRIS core. After an overview of the experimental devices irradiated in the OSIRIS reactor, we present the calculation scheme and the first results of the experimental validation. (authors)

  11. 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)

    2013-07-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    None

    2009-12-01T23:59:59.000Z

    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).

  13. 2012 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect (OSTI)

    Mike Lewis

    2013-02-01T23:59:59.000Z

    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, 2011 through October 31, 2012. 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 2012 permit year, approximately 183 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.

  14. 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

    2013-09-01T23:59:59.000Z

    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.

  15. 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

    2012-02-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Sam Marutzky

    2010-09-01T23:59:59.000Z

    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

  17. Experiment Safety Assurance Package for Mixed Oxide Fuel Irradiation in an Average Power Position (I-24) in the Advanced Test Reactor

    SciTech Connect (OSTI)

    J. M . Ryskamp; R. C. Howard; R. C. Pedersen; S. T. Khericha

    1998-10-01T23:59:59.000Z

    The Fissile Material Disposition Program Light Water Reactor Mixed Oxide Fuel Irradiation Test Project Plan details a series of test irradiations designed to investigate the use of weapons-grade plutonium in MOX fuel for light water reactors (LWR) (Cowell 1996a, Cowell 1997a, Thoms 1997a). Commercial MOX fuel has been successfully used in overseas reactors for many years; however, weapons-derived test fuel contains small amounts of gallium (about 2 parts per million). A concern exists that the gallium may migrate out of the fuel and into the clad, inducing embrittlement. For preliminary out-of-pile experiments, Wilson (1997) states that intermetallic compound formation is the principal interaction mechanism between zircaloy cladding and gallium. This interaction is very limited by the low mass of gallium, so problems are not expected with the zircaloy cladding, but an in-pile experiment is needed to confirm the out-of-pile experiments. Ryskamp (1998) provides an overview of this experiment and its documentation. The purpose of this Experiment Safety Assurance Package (ESAP) is to demonstrate the safe irradiation and handling of the mixed uranium and plutonium oxide (MOX) Fuel Average Power Test (APT) experiment as required by Advanced Test Reactor (ATR) Technical Safety Requirement (TSR) 3.9.1 (LMITCO 1998). This ESAP addresses the specific operation of the MOX Fuel APT experiment with respect to the operating envelope for irradiation established by the Upgraded Final Safety Analysis Report (UFSAR) Lockheed Martin Idaho Technologies Company (LMITCO 1997a). Experiment handling activities are discussed herein.

  18. Corrective Action Investigation Plan for Corrective Action Unit 266: Area 25 Building 3124 Leachfield, Nevada Test Site, Nevada, Revision 1, February 1999

    SciTech Connect (OSTI)

    U.S. Department Of Energy, Nevada Operations Office

    1999-02-24T23:59:59.000Z

    The Corrective Action Investigation Plan for Corrective Action Unit 266, Area 25 Building 3124 Leachfield, has been developed in accordance with the Federal Facility Agreement and Consent Order that was agreed to by the U.S. Department of Energy, Nevada Operations Office; the State of Nevada Division of Environmental Protection; and the U.S. Department of Defense. Corrective Action Unit 266 consists of the Corrective Action Site 25-05-09 sanitary leachfield and associated collection system. This Corrective Action Investigation Plan is used in combination with the Work Plan for Leachfield Corrective Action Units: Nevada Test Site and Tonopah Test Range, Nevada (DOE/NV, 1998d). This Corrective Action Investigation Plan provides investigative details specific to Corrective Action Unit 266. Corrective Action Unit 266 is located southwest of Building 3124 which is located southwest and adjacent to Test Cell A. Test Cell A was operational during the 1960s to test nuclear rocket reactors in support of the Nuclear Rocket Development Station. Operations within Building 3124 from 1962 through the early 1990s resulted in effluent releases to the leachfield and associated collection system. The subsurface soils in the vicinity of the collection system and leachfield may have been impacted by effluent containing contaminants of potential concern generated by support activities associated with Test Cell A reactor testing operations, various laboratories including a high-level radioactivity environmental sample handling laboratory, and possibly the Treatability Test Facility. Based on site history collected to support the Data Quality Objectives process, contaminants of potential concern for the site include radionuclides, oil/diesel range total petroleum hydrocarbons, and Resource Conservation and Recovery Act characteristic volatile organic compounds, semivolatile organic compounds, and metals. Samples will also be analyzed for radionuclides and polychlorinated biphenyls not considered during the DQO process. Additional samples will be analyzed for geotechnical and hydrological properties and a bioassessment may be performed. The technical approach for investigating this Corrective Action Unit consists of the following activities: (1) Perform a radiological walkover survey. (2) Perform video and radiation surveys of the discharge and outfall lines. (3) Collect samples from within the septic tank. (4) Mark approximate locations of leachfield distribution lines on the ground surface. (5) Collect subsurface soil samples in areas of the collection system including the septic tank and outfall end of the diversion chamber. (6) Collect subsurface soil samples underlying the leachfield distribution pipes. (7) Field screen samples for volatile organic compounds and radiological activity. (8) Drill boreholes and collect subsurface soil samples if required. (9) Analyze soil samples for total volatile organic compounds, total semivolatile organic compounds, total Resource Conservation and Recovery Act metals, total petroleum hydrocarbons (oil/diesel-range organics), and polychlorinated biphenyls. (1) Analyze a minimum of 25 percent of the soil samples for gamma-emitting radionuclides, isotopic uranium, isotopic plutonium, isotopic americium, and strontium-90 if radiological field screening levels are exceeded. (2) Collect samples from native soils beneath the distribution system and analyze for geotechnical/hydrologic parameters. (3) Collect and analyze bioassessment samples at Site Supervisors discretion if volatile organic compounds exceed field-screening levels. Additional sampling and analytical details are presented.

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

    SciTech Connect (OSTI)

    None

    1996-12-01T23:59:59.000Z

    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.

  20. Application specific Tester-On-a-Resident-Chip (TORCH{trademark}) - innovation in the area of semiconductor testing

    SciTech Connect (OSTI)

    Bowles, M. [L& M Technologies, Albuquerque, NM (United States); Peterson, T. [New Mexico Highlands Univ., Las Vegas, NM (United States); Savignon, D.; Campbell, D. [Sandia National Labs., Albuquerque, NM (United States)

    1997-12-01T23:59:59.000Z

    Manufacturers widely recognize testing as a major factor in the cost, producability, and delivery of product in the $100 billion integrated circuit business: {open_quotes}The rapid development of VLSI using sub-micron CMOS technology has suddenly exposed traditional test techniques as a major cost factor that could restrict the development of VLSI devices exceeding 512 pins an operating frequencies above 200 MHz.{close_quotes} -- 1994 Semiconductor Industry Association Roadmap, Design and Test, Summary, pg. 43. This problem increases dramatically for stockpile electronics, where small production quantities make it difficult to amortize the cost of increasingly expensive testers. Application of multiple ICs in Multi-Chip Modules (MCM) greatly multiplies testing problems for commercial and defense users alike. By traditional test methods, each new design requires custom test hardware and software and often dedicated testing equipment costing millions of dollars. Also, physical properties of traditional test systems often dedicated testing equipment costing millions of dollars. Also, physical properties of traditional test systems limit capabilities in testing at-speed (>200 MHz), high-impedance, and high-accuracy analog signals. This project proposed a revolutionary approach to these problems: replace the multi-million dollar external test system with an inexpensive test system integrated onto the product wafer. Such a methodology enables testing functions otherwise unachievable by conventional means, particularly in the areas of high-frequency, at-speed testing, high impedance analog circuits, and known good die assessment. The techniques apply specifically to low volume applications, typical of Defense Programs, where testing costs represent an unusually high proportional of product costs, not easily amortized.

  1. 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

    2013-01-01T23:59:59.000Z

    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 ...

  2. Focused evaluation of selected remedial alternatives for the underground test area

    SciTech Connect (OSTI)

    NONE

    1997-04-01T23:59:59.000Z

    The Nevada Test Site (NTS), located in Nye County in southern Nevada, was the location of 928 nuclear tests conducted between 1951 and 1992. Of the total tests, 824 were nuclear tests performed underground. This report describes the approach taken to determine whether any specific, proven, cost-effective technologies currently exist to aid in the removal of the radioactive contaminants from the groundwater, in the stabilization of these contaminants, and in the removal of the source of the contaminants.

  3. 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]

    experimental thermonuclear reactor on discharge test in July China's new generation experimental Tokamak fusion03/01/2006 09:51 AMLoading "People's Daily Online -- Chinese experimental thermonuclear reactor successful, it would be the world's first experimental nuclear fusion device to come into operation. Li

  4. Corrective action plan for corrective action Unit 342: Area 23 Mercury Fire Training Pit, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    Nacht, S.

    1999-08-01T23:59:59.000Z

    The Mercury Fire Training Pit is a former fire training area located in Area 23 of the Nevada Test Site (NTS). The Mercury Fire Training Pit was used from approximately 1965 to the early 1990s to train fire-fighting personnel at the NTS, and encompasses an area approximately 107 meters (m) (350 feet [ft]) by 137 m (450 ft). The Mercury Fire Training Pit formerly included a bermed burn pit with four small burn tanks, four large above ground storage tanks an overturned bus, a telephone pole storage area, and areas for burning sheds, pallets, and cables. Closure activities will include excavation of the impacted soil in the aboveground storage tank and burn pit areas to a depth of 1.5 m (5 ft), and excavation of the impacted surface soil downgradient of the former ASTs and burnpit areas to a depth of 0.3 m (1 ft). Excavated soil will be disposed in the Area 6 Hydrocarbon Landfill at the NTS.

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

    SciTech Connect (OSTI)

    P. Oberlander; D. McGraw; C. Russell

    2007-10-31T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    none,

    2013-06-27T23:59:59.000Z

    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

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

    SciTech Connect (OSTI)

    Gerber, M.S.

    1993-09-01T23:59:59.000Z

    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.

  8. Smart Parking Linked to Transit: Lessons Learned from the San Francisco Bay Area Field Test

    E-Print Network [OSTI]

    Shaheen, Susan; Kemmerer, Charlene

    2007-01-01T23:59:59.000Z

    LINKED TO TRANSIT: LESSONS LEARNED FROM THE SAN FRANCISCOmonth on average. Key lessons learned include that it wouldof the field test, and lessons learned. Key Words: Smart

  9. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06T23:59:59.000Z

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  10. 2007 Annual Summary Report for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site, Nye County, Nevada

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2008-01-01T23:59:59.000Z

    This report summarizes the results of an annual review of conditions affecting the operation of the Area 3 and Area 5 Radioactive Waste Management Sites (RWMSs) and a determination of the continuing adequacy of the performance assessments (PAs) and composite analyses (CAs). The Area 5 RWMS PA documentation consists of the original PA (Shott et al., 1998), referred to as the 1998 Area 5 RWMS PA and supporting addenda (Bechtel Nevada [BN], 2001b; 2006a). The Area 5 RWMS CA was issued as a single document (BN, 2001a) and has a single addendum (BN, 2001c). The Area 3 PA and CA were issued in a single document (Shott et al., 2000). The Maintenance Plan for the PAs and CAs (National Security Technologies, LLC [NSTec], 2006) and the Disposal Authorization Statements (DASs) for the Area 3 and 5 RWMSs (U.S. Department of Energy [DOE], 2000; 2002) require preparation of an annual summary and a determination of the continuing adequacy of the PAs and CAs. The annual summary report is submitted to DOE Headquarters. Following the annual report format in the DOE PA/CA Maintenance Guide (DOE, 1999), this report presents the annual summary for the PAs in Section 2.0 and the CAs in Section 3.0. The annual summary for the PAs includes the following: Section 2.1 summarizes changes in waste disposal operations; Section 2.1.5 provides an evaluation of the new estimates of the closure inventories derived from the actual disposals through fiscal year (FY) 2007; Section 2.2 summarizes the results of the monitoring conducted under the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office's (NNSA/NSO's) Integrated Closure and Monitoring Plan for the Area 3 and Area 5 Radioactive Waste Management Sites at the Nevada Test Site (BN, 2005), and the research and development (R&D) activities; Section 2.4 is a summary of changes in facility design, operation, or expected future conditions; monitoring and R&D activities; and the maintenance program; and Section 2.5 discusses the recommended changes in disposal facility design and operations, monitoring and R&D activities, and the maintenance program. Similarly, the annual summary for the CAs (presented in Section 3.0) includes the following: Section 3.1 presents the assessment of the adequacy of the CAs, with a summary of the relevant factors reviewed in FY 2007; Section 3.2 presents an assessment of the relevant site activities at the Nevada Test Site (NTS) that would impact the sources of residual radioactive material considered in the CAs; Section 3.3 summarizes the monitoring and R&D results that were reviewed in FY 2007; Section 3.4 presents a summary of changes in relevant site programs (including monitoring, R&D, and the maintenance program) that occurred since the CAs were prepared; and Section 3.5 summarizes the recommended changes to these programs.

  11. Nevada Test Site 2000 Annual Data Report: Groundwater Monitoring Program Area 5 Radioactive Waste Management Site

    SciTech Connect (OSTI)

    Y. E.Townsend

    2001-02-01T23:59:59.000Z

    This report is a compilation of the calendar year 2000 groundwater sampling results from the Area 5 Radioactive Waste Management Site (RWMS). Contamination indicator data are presented in control chart and tabular form with investigation levels (IL) indicated. Gross water chemistry data are presented in graphical and tabular form. Other information in the report includes, the Cumulative Chronology for Area 5 RWMS Groundwater Monitoring Program, a brief description of the site hydrogeology, and the groundwater sampling procedure.

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

    SciTech Connect (OSTI)

    GERBER MS

    2009-04-28T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    NONE

    1994-12-31T23:59:59.000Z

    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.

  14. 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. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

    2012-07-01T23:59:59.000Z

    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)

  15. Summary of aerosol code-comparison results for LWR (Light-Water Reactor) aerosol containment tests LA1, LA2, and LA3: (Final report)

    SciTech Connect (OSTI)

    Wright, A.L.; Wilson, J.H.; Arwood, P.C.

    1986-01-01T23:59:59.000Z

    The Light-Water Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadershiop of an international project board and the Electric Power Research Institute. These tests have two objectives: to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities for the LACE tests are being coordinated at the Oak Ridge National Laboratory. For each of the six experiments, ''pretest'' calculations (for code-to-test data comparisons) are being performed. This paper presents a summary of the pretest aerosol-code results for tests LA1, LA2, and LA3. 7 refs., 11 figs., 2 tabs.

  16. Reactor Safety Research Programs Quarterly Report January - March 1980

    SciTech Connect (OSTI)

    Hagen, C. M

    1980-10-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory from January 1 through March 31, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where serviceinduced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  17. Reactor Safety Research Programs Quarterly Report July- September 1980

    SciTech Connect (OSTI)

    Edler, S. K.

    1980-12-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  18. Reactor Safety Research Programs Quarterly Report April -June 1980

    SciTech Connect (OSTI)

    Edler, S. K.

    1980-11-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  19. 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.

    1987-10-01T23:59:59.000Z

    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.

  20. Flow Test At The Needles 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdf Jump to:ar-80m.pdfFillmoreGabbs Valley Area (DOE GTP)The Needles Area (DOE

  1. Flow Test At Valles Caldera - Sulphur Springs Geothermal Area (Musgrave, Et

    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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdf Jump to:ar-80m.pdfFillmoreGabbs Valley Area (DOE GTP)The Needles Area (DOEAl.,

  2. Tracer Testing At Neal Hot Springs Geothermal Area (U.S. Geothermal...

    Open Energy Info (EERE)

    testing program was initiated in order to map the flow of geothermal fluids through the reservoir and incorporate the data into the numerical reservoir model for a more accurate...

  3. Post-Closure Monitoring Report for Corrective Action Unit 339: Area 12 Fleet Operations Steam Cleaning Discharge Area Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    A. T. Urbon

    2001-08-01T23:59:59.000Z

    The Area 12 Fleet Operations Steam Cleaning site is located in the southeast portion of the Area 12 Camp at the Nevada Test Site (Figure 1). This site is identified in the Federal Facility Agreement and Consent Order (FFACO, 1996) as Corrective Action Site (CAS) 12-19-01 and is the only CAS assigned to Corrective Action Unit (CAU) 339. Post-closure sampling and inspection of the site were completed on March 23, 2001. Because of questionable representativeness and precision of the results, the site was resampled on June 12, 2001. Post-closure monitoring activities were scheduled biennially (every two years) in the Post-Closure Monitoring Plan provided in the December 1997 Closure Report for CAU 339: Area 12 Fleet Operations Steam Cleaning Discharge Area, Nevada Test Site (U.S. Department of Energy, Nevada Operations Office [DOE/NV], 1997). If after six years the rate of degradation appears to be so slow that the greatest concentration of total petroleum hydrocarbons (TPH) present at the site would not decay within 30 years of the site closure, the site will be reevaluated with consideration to enriching the impacted soil at the site to enhance the degradation process. A baseline for the site was established by sampling in 1997. Based on the recommendations from the 1999 post-closure monitoring report, samples were collected in 2000, earlier than originally proposed, because the 1999 sample results did not provide the expected decrease in TPH concentrations at the site. Sampling results from 2000 revealed favorable conditions for natural degradation at the CAU 339 site, but because of differing sample methods and heterogeneity of the soil, the data results from 2000 were not directly correlated with previous results. Post-closure monitoring activities for 2001 consisted of the following: Soil sample collection from three undisturbed plots (Plots A, B, and C, Figure 2); Sample analysis for TPH as oil and bio-characterization parameters (Comparative Enumeration Assay [CEA] and Standard Nutrient Panel [SNP]); Site inspection to evaluate the condition of the fencing and signs; and Preparation and submittal of the Post-Closure Monitoring Report.

  4. Materials Science Division light-water-reactor safety research program. Quarterly progress report, July-September 1982

    SciTech Connect (OSTI)

    Shack, W.J.; Rest, J.; Kassner, T.F.; Neimark, L.A.; Chung, H.M.; Claytor, T.N.; Kupperman, D.S.; Maiya, P.S.; Nichols, F.A.; Park, J.Y.

    1983-08-01T23:59:59.000Z

    This progress report summarizes the Argonne National Laboratory work performed during July, August, and September 1982 on water reactor safety problems. The research and development areas covered are Environmentally Assisted Cracking in Light Water Reactors, Transient Fuel Response and Fission Product Release, Clad Properties for Code Verification, Posttest Fuel Examination of the ORNL Fission Product Release Tests, and Examination of TMI-2 Fuel Specimens.

  5. Testing the improved method for calculating the radiation heat generation at the periphery of the BOR-60 reactor core

    SciTech Connect (OSTI)

    Varivtsev, A. V., E-mail: vav3@niiar.ru; Zhemkov, I. Yu. [JSC “SSC RIAR,” Dimitrovgrad-10 (Russian Federation)

    2014-12-15T23:59:59.000Z

    The application of the improved method for calculating the radiation heat generation in the elements of an experimental device located at the periphery of the BOR-60 reactor core results in a significant reduction in the discrepancies between the calculated and the experimental data. This allows us to conclude that the improved method has an advantage over the one used earlier.

  6. Nuclear reactor control column

    SciTech Connect (OSTI)

    Bachovchin, D.M.

    1982-08-10T23:59:59.000Z

    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 crosssectional 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.

  7. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M. (Plum Borough, PA)

    1982-01-01T23:59:59.000Z

    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.

  8. 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.)

    1982-08-01T23:59:59.000Z

    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.

  9. 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

    2013-10-01T23:59:59.000Z

    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.

  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)

    S. E. Rawlinson

    2001-09-01T23:59:59.000Z

    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.

  11. CORRECTIVE ACTION PLAN FOR CORRECTIVE ACTION UNIT 536: AREA 3 RELEASE SITE, NEVADA TEST SITE, NEVADA

    SciTech Connect (OSTI)

    NONE

    2005-09-01T23:59:59.000Z

    CAU 536 consists of CAS 03-44-02, Steam Jenny Discharge, located in Area 3 of the NTS. The site was characterized in 2004 according to the approved CAIP and the site characterization results are reported in the CAU 536 CADD. The purpose of this Corrective Action Plan (CAP) is to provide the detailed scope of work required to implement the recommended corrective actions as specified in the approved CAU 536 CADD.

  12. Nevada Test Site 2001 Data Report: Groundwater Monitoring Program Area 5 Radioactive Waste Management Site

    SciTech Connect (OSTI)

    Y. E. Townsend

    2002-02-01T23:59:59.000Z

    This report is a compilation of the calendar year 2001 groundwater sampling results from the Area 5 Radioactive Waste Management Site (RWMS). Contamination indicator data are presented in control chart and tabular form with investigation levels (ILs) indicated. Gross water chemistry data are presented in graphical and tabular form. Other information in the report includes, the Cumulative Chronology for Area 5 RWMS Groundwater Monitoring Program, a brief description of the site hydrogeology, and the groundwater sampling procedure. Wells Ue5PW-1, Ue5PW-2, and Ue5PW-3 were sampled semiannually for the required analytes: pH, specific conductance, major cations/anions, metals, tritium, total organic carbon (TOC), and total organic halogen (TOX). Due to detections of TOC and TOX in some samples collected in 2000, a plan, as approved by the Nevada Division of Environmental Protection (NDEP), was executed to collect an increased number and type of samples in 2001. Results from all samples collected in 2001 were below ILs. These data indicate that there has been no measurable impact to the uppermost aquifer from the Resource Conservation and Recovery Act (RCRA) regulated unit within the Area 5 RWMS and confirm that the detections of TOC and TOX in 2000 were false positives. There were no major changes noted in the monitored groundwater elevation. There continues to be an extremely small gradient to the northeast with an average flow velocity of less than one foot per year.

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

    SciTech Connect (OSTI)

    NSTec Environmental Management

    2010-01-19T23:59:59.000Z

    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.

  14. Hanford immobilized LAW product acceptance: Initial Tanks Focus Area testing data package

    SciTech Connect (OSTI)

    JD Vienna; A Jiricka; BP McGrail; BM Jorgensen; DE Smith; BR Allen; JC Marra; DK Peeler; KG Brown; IA Reamer; WL Ebert

    2000-03-08T23:59:59.000Z

    The Hanford Site's mission has been to produce nuclear materials for the US Department of Energy (DOE) and its predecessors. A large inventory of radioactive and mixed waste, largely generated during plutonium production, exists in 177 underground single- and double-shell tanks. These wastes are to be retrieved and separated into low-activity waste (LAW) and high-level waste (HLW) fractions. The total volume of LAW requiring immobilization will include the LAW separated from the tank waste, as well as new wastes generated by the retrieval, pretreatment, and immobilization processes. Per the Tri-Party Agreement (1994), both the LAW and HLW will be vitrified. It has been estimated that vitrification of the LAW waste will result in over 500,000 metric tons or 200,000 m{sup 3} of immobilized LAW (ILAW) glass. The ILAW glass is to be disposed of onsite in a near-surface burial facility. It must be demonstrated that the disposal system will adequately retain the radionuclides and prevent contamination of the surrounding environment. This report describes a study of the impacts of systematic glass-composition variation on the responses from accelerated laboratory corrosion tests of representative LAW glasses. A combination of two tests, the product consistency test and vapor-hydration test, is being used to give indictations of the relative rate at which a glass could be expected to corrode in the burial scenario.

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

    SciTech Connect (OSTI)

    V. Yucel

    2001-09-01T23:59:59.000Z

    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).

  16. Corrective Action Investigation Plan for Corrective Action Unit 551: Area 12 Muckpiles, Nevada Test Site, Nevada, Rev. No.: 0

    SciTech Connect (OSTI)

    David A. Strand

    2004-06-01T23:59:59.000Z

    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) 551, Area 12 muckpiles, 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. Corrective Action Unit 551 is located in Area 12 of the NTS, which is approximately 110 miles (mi) northwest of Las Vegas, Nevada (Figure 1-1). Area 12 is approximately 40 miles beyond the main gate to the NTS. Corrective Action Unit 551 is comprised of the four Corrective Action Sites (CASs) shown on Figure 1-1 and listed below: (1) 12-01-09, Aboveground Storage Tank and Stain; (2) 12-06-05, Muckpile; (3) 12-06-07, Muckpile; and (4) 12-06-08, Muckpile. Corrective Action Site 12-01-09 is located in Area 12 and consists of an above ground storage tank (AST) and associated stain. Corrective Action Site 12-06-05 is located in Area 12 and consists of a muckpile associated with the U12 B-Tunnel. Corrective Action Site 12-06-07 is located in Area 12 and consists of a muckpile associated with the U12 C-, D-, and F-Tunnels. Corrective Action Site 12-06-08 is located in Area 12 and consists of a muckpile associated with the U12 B-Tunnel. In keeping with common convention, the U12B-, C-, D-, and F-Tunnels will be referred to as the B-, C-, D-, and F-Tunnels. The corrective action investigation (CAI) will include field inspections, radiological surveys, and sampling of media, where appropriate. Data will also be obtained to support waste management decisions.

  17. Corrective Action Investigation Plan for Corrective Action Unit 551: Area 12 Muckpiles, Nevada Test Site, Nevada, Rev. No. 0

    SciTech Connect (OSTI)

    Robert F. Boehlecke

    2004-06-01T23:59:59.000Z

    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) 551, Area 12 muckpiles, 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. Corrective Action Unit 551 is located in Area 12 of the NTS, which is approximately 110 miles (mi) northwest of Las Vegas, Nevada (Figure 1-1). Area 12 is approximately 40 miles beyond the main gate to the NTS. Corrective Action Unit 551 is comprised of the four Corrective Action Sites (CASs) shown on Figure 1-1 and listed below: (1) 12-01-09, Aboveground Storage Tank and Stain; (2) 12-06-05, Muckpile; (3) 12-06-07, Muckpile; and (4) 12-06-08, Muckpile. Corrective Action Site 12-01-09 is located in Area 12 and consists of an above ground storage tank (AST) and associated stain. Corrective Action Site 12-06-05 is located in Area 12 and consists of a muckpile associated with the U12 B-Tunnel. Corrective Action Site 12-06-07 is located in Area 12 and consists of a muckpile associated with the U12 C-, D-, and F-Tunnels. Corrective Action Site 12-06-08 is located in Area 12 and consists of a muckpile associated with the U12 B-Tunnel. In keeping with common convention, the U12B-, C-, D-, and F-Tunnels will be referred to as the B-, C-, D-, and F-Tunnels. The corrective action investigation (CAI) will include field inspections, radiological surveys, and sampling of media, where appropriate. Data will also be obtained to support waste management decisions.

  18. Technical Letter Report, An Evaluation of Ultrasonic Phased Array Testing for Reactor Piping System Components Containing Dissimilar Metal Welds, JCN N6398, Task 2A

    SciTech Connect (OSTI)

    Diaz, Aaron A.; Cinson, Anthony D.; Crawford, Susan L.; Anderson, Michael T.

    2009-11-30T23:59:59.000Z

    Research is being conducted for the U.S. Nuclear Regulatory Commission at the Pacific Northwest National Laboratory to assess the effectiveness and reliability of advanced nondestructive examination (NDE) methods for the inspection of light-water reactor components. The scope of this research encom¬passes primary system pressure boundary materials including dissimilar metal welds (DMWs), cast austenitic stainless steels (CASS), piping with corrosion-resistant cladding, weld overlays, inlays and onlays, and far-side examinations of austenitic piping welds. A primary objective of this work is to evaluate various NDE methods to assess their ability to detect, localize, and size cracks in steel components that challenge standard and/or conventional inspection methodologies. This interim technical letter report provides a summary of a technical evaluation aimed at assessing the capabilities of phased-array (PA) ultrasonic testing (UT) methods as applied to the inspection of small-bore DMW components that exist in the reactor coolant systems (RCS) of pressurized water reactors (PWRs). Operating experience and events such as the circumferential cracking in the reactor vessel nozzle-to-RCS hot leg pipe at V.C. Summer nuclear power station, identified in 2000, show that in PWRs where primary coolant water (or steam) are present under normal operation, Alloy 82/182 materials are susceptible to pressurized water stress corrosion cracking. The extent and number of occurrences of DMW cracking in nuclear power plants (domestically and internationally) indicate the necessity for reliable and effective inspection techniques. The work described herein was performed to provide insights for evaluating the utility of advanced NDE approaches for the inspection of DMW components such as a pressurizer surge nozzle DMW, a shutdown cooling pipe DMW, and a ferritic (low-alloy carbon steel)-to-CASS pipe DMW configuration.

  19. Reactor safety research programs. Quarterly report Apr-Jun 81

    SciTech Connect (OSTI)

    Edler, S.K.

    1981-09-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  20. Reactor Safety Research Programs Quarterly Report April- June 1981

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-09-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest laboratory (PNL} from April1 through June 30, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory {INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  1. Flow Test At Fenton Hill HDR Geothermal Area (Dash, Et Al., 1983) | 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnualPropertyd8c-a9ae-f8521cbb8489Information Hydro Inc IosilEnergy Information Area (Dash,

  2. Flow Test At Fenton Hill HDR Geothermal Area (Grigsby, Et Al., 1983) | 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnualPropertyd8c-a9ae-f8521cbb8489Information Hydro Inc IosilEnergy Information Area

  3. Flow Test At Black Warrior 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdf Jump to:ar-80m.pdfFillmore County,andJumpInformationBlack Warrior Area

  4. Flow Test At Gabbs Valley 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdf Jump to:ar-80m.pdfFillmoreGabbs Valley Area (DOE GTP) Exploration Activity

  5. Flow Test At Glass Buttes 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdf Jump to:ar-80m.pdfFillmoreGabbs Valley Area (DOE GTP) Exploration ActivityFlow

  6. Flow Test At Hot Pot 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdf Jump to:ar-80m.pdfFillmoreGabbs Valley Area (DOE GTP) Exploration

  7. Flow Test At Mccoy 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdf Jump to:ar-80m.pdfFillmoreGabbs Valley Area (DOE GTP) ExplorationMccoy

  8. Flow Test At Pilgrim Hot Springs 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are8COaBulkTransmissionSitingProcess.pdf Jump to:ar-80m.pdfFillmoreGabbs Valley Area (DOE GTP) ExplorationMccoyFlow

  9. Nevada Test Site 2002 Data Report: Groundwater Monitoring Program Area 5 Radioactive Waste Management Site

    SciTech Connect (OSTI)

    Y. E. Townsend

    2003-02-01T23:59:59.000Z

    This report is a compilation of the calendar year 2002 groundwater sampling results from the Area 5 Radioactive Waste Management Site (RWMS). Wells Ue5PW-1, Ue5PW-2, and Ue5PW-3 were sampled semiannually for the required analytes: pH, specific conductance, major cations/anions, metals, tritium, total organic carbon (TOC), and total organic halogen (TOX). Results from all samples collected in 2002 were within established criteria. These data indicate that there has been no measurable impact to the uppermost aquifer from the Resource Conservation and Recovery Act(RCRA) regulated unit within the RWMS-5 and confirm that the detections of TOC and TOX in 2000 were false positives. Contamination indicator data are presented in control chart and tabular form with investigation levels (ILs) indicated. Gross water chemistry data are presented in graphical and tabular form. There were no major changes noted in the monitored groundwater elevation. There continues to be an extremely small gradient to the northeast with an average flow velocity of less than one foot per year. Other information in the report includes, the Cumulative Chronology for Area 5 RWMS Groundwater Monitoring Program, a brief description of the site hydrogeology, and the groundwater sampling procedure.

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

    SciTech Connect (OSTI)

    NSTec Environmental Restoration

    2011-09-29T23:59:59.000Z

    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.

  11. Test in a beam of large-area Micromegas chambers for sampling calorimetry

    E-Print Network [OSTI]

    C. Adloff; M. Chefdeville; A. Dalmaz; C. Drancourt; R. Gaglione; N. Geffroy; J. Jacquemier; Y. Karyotakis; I. Koletsou; F. Peltier; J. Samarati; G. Vouters

    2014-05-03T23:59:59.000Z

    Application of Micromegas for sampling calorimetry puts specific constraints on the design and performance of this gaseous detector. In particular, uniform and linear response, low noise and stability against high ionisation density deposits are prerequisites to achieving good energy resolution. A Micromegas-based hadronic calorimeter was proposed for an application at a future linear collider experiment and three technologically advanced prototypes of 1$\\times$1 m$^{2}$ were constructed. Their merits relative to the above-mentioned criteria are discussed on the basis of measurements performed at the CERN SPS test-beam facility.

  12. 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 onYou are now leaving Energy.gov You are now leaving Energy.gov You are being directedAnnualPropertyd8c-a9ae-f8521cbb8489Information Hydro IncEnergy Information RooseveltFlow Test

  13. A review of existing gas-cooled reactor circulators with application of the lessons learned to the new production reactor circulators

    SciTech Connect (OSTI)

    White, L.S.

    1990-07-01T23:59:59.000Z

    This report presents the results of a study of the lessons learned during the design, testing, and operation of gas-cooled reactor coolant circulators. The intent of this study is to identify failure modes and problem areas of the existing circulators so this information can be incorporated into the design of the circulators for the New Production Reactor (NPR)-Modular High-Temperature Gas Cooled Reactor (MHTGR). The information for this study was obtained primarily from open literature and includes data on high-pressure, high-temperature helium test loop circulators as well as the existing gas cooled reactors worldwide. This investigation indicates that trouble free circulator performance can only be expected when the design program includes a comprehensive prototypical test program, with the results of this test program factored into the final circulator design. 43 refs., 7 tabs.

  14. 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.

    2000-06-07T23:59:59.000Z

    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.

  15. 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.

    2000-04-01T23:59:59.000Z

    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.

  16. 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

    2009-05-01T23:59:59.000Z

    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.

  17. Reactor Safety Research Programs Quarterly Report October - December 1980

    SciTech Connect (OSTI)

    Edler, S K

    1981-04-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  18. Design and Testing of a 10B4C Capsule for Spectral-Tailoring in Mixed-Spectrum Reactors

    SciTech Connect (OSTI)

    Greenwood, Lawrence R.; Wittman, Richard S.; Metz, Lori A.; Finn, Erin C.; Friese, Judah I.

    2014-04-11T23:59:59.000Z

    A boron carbide capsule highly enriched in 10B has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. New experiments show that enriching the boron to 96% B-10 results in additional absorption of neutrons in the resonance region thereby producing a neutron spectrum that is much closer to a pure 235U fission spectrum. A cadmium outer cover was used to reduce thermal heating. The neutron spectrum calculated with MCNP was found to be in very good agreement with measured activation rates from neutron fluence monitors.

  19. 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

    2012-09-17T23:59:59.000Z

    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.

  20. Laboratory Evaluation of In Situ Chemical Oxidation for Groundwater Remediation, Test Area North, Operable Unit 1-07B, Idaho National Engineering and Environmental Laboratory, Volume One - Main Text and Appendices A and B

    SciTech Connect (OSTI)

    Cline, S.R.; Denton, D.L.; Giaquinto, J.M.; McCracken, M.K.; Starr, R.C.

    1999-04-01T23:59:59.000Z

    The laboratory investigation was performed to evaluate the feasibility of utilizing in situ chemical oxidation for remediating the secondary source of groundwater contaminants at the Idaho National Engineering and Environmental Laboratory (INEEL) Test Area North (TAN) Site. The study involved trichloroethene (TCE) contaminated media (groundwater, soil, and sludge) from TAN. The effectiveness of the selected oxidant, potassium permanganate (KMn0(sub4)), was evaluated at multiple oxidant and contaminant concentrations. Experiments were performed to determine the oxidant demand of each medium and the rate of TCE oxidation. The experiments were performed under highly controlled conditions (gas-tight reactors, constant 12C temperature). Multiple parameter were monitored over time including MN0(sub 4) and TCE concentrations and pH.

  1. 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

    2009-03-25T23:59:59.000Z

    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.

  2. Final Characterization Report for Corrective Action Unit 109: Area 2 U-2BU Crater, Nevada Test Site, Nevada, Revision 1

    SciTech Connect (OSTI)

    ITLV

    1998-12-01T23:59:59.000Z

    Corrective Action Unit 109, Area 2 U-2bu Crater, is an inactive Resource Conservation and Recovery Act Part A Permit disposal unit located in Area 2 at the Nevada Test Site, Nevada. The Corrective Action Unit has been characterized under the requirements of the Nevada Test Site Resource Conservation and Recovery Act, Part A Permit (NDEP, 1995) and Title 40 Code of Federal Regulations Part 265 (CFR, 1996). The site characterization was performed under the RCRA Part A Permit Characterization Plan for the U-2bu Subsidence Crater (DOE/NV, 1998c), as approved by the Nevada Division of Environmental Protection (Liebendorfer, 1998). The primary objective of the site characterization activities was to evaluate the presence, concentration, and extent of any Resource Conservation and Recovery Act contaminants in the crater. Surface soil samples were collected on April 22, 1998, and subsurface soil samples and geotechnical samples were collected from April 27-29, 1998. Soil samples were collected using a hand auger or a piston-type drive hammer to advance a 5-centimeter (2-inch) diameter steel sampling tool into the ground. The permit for the Nevada Test Site requires that Corrective Action Unit 109 be closed under 40 Code of Federal Regulations 265 Subpart G and 40 Code of Federal Regulations Part 265.310 (CFR, 1996). Analysis of the data collected during the characterization effort indicates that lead was detected in Study Area 1 at 5.7 milligrams per liter, above the regulatory level in 40 Code of Federal Regulations 261.24 of 5.0 milligrams per liter. Except for the lead detection at a single location within the crater, the original Resource Conservation Recovery Act constituents of potential concern determined between the U.S. Department of Energy, Nevada Operations Office and the Nevada Department of Environmental Protection during the Data Quality Objectives process (DOE/NV, 1998b) were not found to be present at Corrective Action Unit 109 above regulatory levels of concern. The single lead detection that exceeded regulatory limits was discovered at a depth of 1.2 meters (4 feet) in Study Area 1. Total petroleum hydrocarbons were detected in Study Area 4 at two locations at concentrations of 130 and 190 milligrams per kilogram. These concentrations exceed the 100 milligram per kilogram action level established in Nevada Administrative Code 445A.2272 (NAC, 1996). The primary conceptual model identified during the Data Quality Objectives process appears to have been substantiated by the analytical results from site characterization; migration of contaminants of potential concern does not appear to be occurring, as none were detected in a significant percentage of the characterization samples. Based on the results of the characterization, clean closure by removal and disposal of impacted soil will be evaluated in the closure plan to address Study Area 1. An (a) through (k) analysis, as specified in Nevada Administrative Code 445A.227, should be used to evaluate total petroleum hydrocarbon contamination within Study Area 4. Study Areas 2, 3, and 5 should be clean closed without further assessment or remediation (see Figures 2-1 and 5-1 in the Characterization Report).

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

    SciTech Connect (OSTI)

    John D. Bess; Nozomu Fujimoto

    2014-11-01T23:59:59.000Z

    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.

  4. 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

    2014-10-09T23:59:59.000Z

    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

  5. 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

    2014-10-09T23:59:59.000Z

    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.

  6. 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

    2010-06-01T23:59:59.000Z

    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.

  7. Reactor operation environmental information document

    SciTech Connect (OSTI)

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

    1989-12-01T23:59:59.000Z

    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.

  8. 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

    2004-04-06T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    BROWN,THERESA J.; WIRTH,SHARON

    1999-09-01T23:59:59.000Z

    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.

  10. Well Installation Report for Corrective Action Unit 443, Central Nevada Test Area, Nye County, Nevada, Rev. No.: 0

    SciTech Connect (OSTI)

    Tim Echelard

    2006-01-01T23:59:59.000Z

    A Corrective Action Investigation (CAI) was performed in several stages from 1999 to 2003, as set forth in the ''Corrective Action Investigation Plan for the Central Nevada Test Area Subsurface Sites, Corrective Action Unit 443'' (DOE/NV, 1999). Groundwater modeling was the primary activity of the CAI. Three phases of modeling were conducted for the Faultless underground nuclear test. The first phase involved the gathering and interpretation of geologic and hydrogeologic data, and inputting the data into a three-dimensional numerical model to depict groundwater flow. The output from the groundwater flow model was used in a transport model to simulate the migration of a radionuclide release (Pohlmann et al., 2000). The second phase of modeling (known as a Data Decision Analysis [DDA]) occurred after NDEP reviewed the first model. This phase was designed to respond to concerns regarding model uncertainty (Pohll and Mihevc, 2000). The third phase of modeling updated the original flow and transport model to incorporate the uncertainty identified in the DDA, and focused the model domain on the region of interest to the transport predictions. This third phase culminated in the calculation of contaminant boundaries for the site (Pohll et al., 2003). Corrective action alternatives were evaluated and an alternative was submitted in the ''Corrective Action Decision Document/Corrective Action Plan for Corrective Action Unit 443: Central Nevada Test Area-Subsurface'' (NNSA/NSO, 2004). Based on the results of this evaluation, the preferred alternative for CAU 443 is Proof-of-Concept and Monitoring with Institutional Controls. This alternative was judged to meet all requirements for the technical components evaluated and will control inadvertent exposure to contaminated groundwater at CAU 443.

  11. Fracture toughness testing of Linde 1092 reactor vessel welds in the transition range using Charpy-sized specimens

    SciTech Connect (OSTI)

    Pavinich, W.A. [Framatome Technologies Inc., Knoxville, TN (United States); Yoon, K.K. [Framatome Technologies Inc., Lynchburg, VA (United States); Hour, K.Y. [Babcock and Wilcox Co., Lynchburg, VA (United States). Research and Development Div.; Hoffman, C.L. [ABB-CE, Windsor, CT (United States)

    1999-10-01T23:59:59.000Z

    The present reference toughness method for predicting the change in fracture toughness can provide over estimates of these values because of uncertainties in initial RT{sub NDT} and shift correlations. It would be preferable to directly measure fracture toughness. However, until recently, no standard method was available to characterize fracture toughness in the transition range. ASTM E08 has developed a draft standard that shows promise for providing lower bound transition range fracture toughness using the master curve approach. This method has been successfully implemented using 1T compact fracture specimens. Combustion Engineering reactor vessel surveillance programs do not have compact fracture specimens. Therefore, the CE Owners Group developed a program to validate the master curve method for Charpy-sized and reconstituted Charpy-sized specimens for future application on irradiated specimens. This method was validated for Linde 1092 welds using unirradiated Charpy-sized and reconstituted Charpy-sized specimens by comparison of results with those from compact fracture specimens.

  12. Thermal Hydraulic Analysis of a Reduced Scale High Temperature Gas-Cooled Reactor Test Facility and its Prototype with MELCOR 

    E-Print Network [OSTI]

    Beeny, Bradley Aaron 1988-

    2012-11-12T23:59:59.000Z

    ................ 86 6.4 Area-averaged outer RPV wall temperature during PCC .............................. 88 6.5 Mass-averaged (by ring) core graphite temperatures during DCC ................ 89 6.6 Mass-averaged (by level) core... graphite temperatures during DCC ............... 91 6.7 Area-averaged outer RPV wall temperature during DCC .............................. 92 6.8 Steady-state core structural temperature distribution...

  13. 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)

    2013-09-15T23:59:59.000Z

    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.

  14. Reactor Safety Research Programs Quarterly Report October - December 1981

    SciTech Connect (OSTI)

    Edler, S. K.

    1982-03-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from October 1 through December 31, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where serviceinduced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and post accident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  15. Reactor Safety Research Programs Quarterly Report July - September 1981

    SciTech Connect (OSTI)

    Edler, S. K.

    1982-01-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  16. Shielding analysis for the 300 area light water reactor spent nuclear fuel within a modified multi-canister overpack canister in a modified multi-canister overpack cask

    SciTech Connect (OSTI)

    Gedeon, S.R.

    1997-04-11T23:59:59.000Z

    Spent light water reactor fuel is to be moved out of the 324 Building. It is anticipated that intact fuel assemblies will be loaded in a modified Multi-Canister Overpack Canister, which in turn will be placed in an Overpack Transportation Cask. An estimate of gamma ray dose rates from a transportation cask is desired.

  17. Record of Technical Change for Corrective Action Plan for Corrective Action Unit 140: Waste Dumps, burn Pits, and Storage Area, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

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

    2005-01-05T23:59:59.000Z

    Record of Technical Change for Corrective Action Plan for Corrective Action Unit 140: Waste Dumps, Burn Pits, and Storage Area, Nevada Test Site, Nevada (DOE/NV--963-Rev 2, dated November 2004).

  18. The Tokamak Fusion Test Reactor decontamination and decommissioning project and the Tokamak Physics Experiment at the Princeton Plasma Physics Laboratory. Environmental Assessment

    SciTech Connect (OSTI)

    NONE

    1994-05-27T23:59:59.000Z

    If the US is to meet the energy needs of the future, it is essential that new technologies emerge to compensate for dwindling supplies of fossil fuels and the eventual depletion of fissionable uranium used in present-day nuclear reactors. Fusion energy has the potential to become a major source of energy for the future. Power from fusion energy would provide a substantially reduced environmental impact as compared with other forms of energy generation. Since fusion utilizes no fossil fuels, there would be no release of chemical combustion products to the atmosphere. Additionally, there are no fission products formed to present handling and disposal problems, and runaway fuel reactions are impossible due to the small amounts of deuterium and tritium present. The purpose of the TPX Project is to support the development of the physics and technology to extend tokamak operation into the continuously operating (steady-state) regime, and to demonstrate advances in fundamental tokamak performance. The purpose of TFTR D&D is to ensure compliance with DOE Order 5820.2A ``Radioactive Waste Management`` and to remove environmental and health hazards posed by the TFTR in a non-operational mode. There are two proposed actions evaluated in this environmental assessment (EA). The actions are related because one must take place before the other can proceed. The proposed actions assessed in this EA are: the decontamination and decommissioning (D&D) of the Tokamak Fusion Test Reactor (TFTR); to be followed by the construction and operation of the Tokamak Physics Experiment (TPX). Both of these proposed actions would take place primarily within the TFTR Test Cell Complex at the Princeton Plasma Physics Laboratory (PPPL). The TFTR is located on ``D-site`` at the James Forrestal Campus of Princeton University in Plainsboro Township, Middlesex County, New Jersey, and is operated by PPPL under contract with the United States Department of Energy (DOE).

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

    SciTech Connect (OSTI)

    Mark Krauss

    2010-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Farnham, Irene; Marutzky, Sam

    2013-01-01T23:59:59.000Z

    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.

  1. Corrective Action Decision Document for Corrective Action Unit 342: Area 23 Mercury Fire Training Pit, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    DOE /NV

    1999-05-26T23:59:59.000Z

    This Corrective Action Decision Document has been prepared for the Nevada Test Site's Area 23 Mercury Fire Training Pit (Corrective Action Unit 342) in accordance with the Federal Facility Agreement and Consent Order (FFACO, 1996). Corrective Action Unit 342 is comprised of Corrective Action Site 23-56-01. 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 Corrective Action Unit 342. The scope of this document consists of the following: Develop corrective action objectives; Identify corrective action alternative screening criteria; Develop corrective action alternatives; Perform detailed and comparative evaluations of corrective action alternatives in relation to corrective action objectives and screening criteria; and Recommend and justify a preferred corrective action alternative for the Corrective Action Unit.

  2. Cultural Resource Assessment of the Test Area North Demolition Landfill at the Idaho National Engineering and Environmental Laboratory

    SciTech Connect (OSTI)

    Brenda R. Pace

    2003-07-01T23:59:59.000Z

    The proposed new demolition landfill at Test Area North on the Idaho National Engineering and Environmental Laboratory (INEEL) will support ongoing demolition and decontamination within the facilities on the north end of the INEEL. In June of 2003, the INEEL Cultural Resource Management Office conducted archival searches, field surveys, and coordination with the Shoshone-Bannock Tribes to identify all cultural resources that might be adversely affected by the project and to provide recommendations to protect those listed or eligible for listing on the National Register of Historic Places. These investigations showed that landfill construction and operation would affect two significant cultural resources. This report outlines protective measures to ensure that these effects are not adverse.

  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

    2014-01-01T23:59:59.000Z

    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 554: Area 23 Release Site, Nevada Test Site, Nevada, Rev. No.: 0

    SciTech Connect (OSTI)

    David A. Strand

    2004-10-01T23:59:59.000Z

    This Corrective Action Investigation Plan (CAIP) contains project-specific information for conducting site investigation activities at Corrective Action Unit (CAU) 554: Area 23 Release Site, Nevada Test Site, Nevada. Information presented in this CAIP includes facility descriptions, environmental sample collection objectives, and criteria for the selection and evaluation of environmental samples. Corrective Action Unit 554 is located in Area 23 of the Nevada Test Site, which is 65 miles northwest of Las Vegas, Nevada. Corrective Action Unit 554 is comprised of one Corrective Action Site (CAS), which is: 23-02-08, USTs 23-115-1, 2, 3/Spill 530-90-002. This site consists of soil contamination resulting from a fuel release from underground storage tanks (USTs). Corrective Action Site 23-02-08 is being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives. Additional information will be obtained by conducting a corrective action investigation prior to evaluating corrective action alternatives and selecting the appropriate corrective action for this CAS. The results of the field investigation will support a defensible evaluation of viable corrective action alternatives that will be presented in the Corrective Action Decision Document for CAU 554. Corrective Action Site 23-02-08 will be investigated based on the data quality objectives (DQOs) developed on July 15, 2004, by representatives of the Nevada Division of Environmental Protection; U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office; and contractor personnel. 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 554.

  5. 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

    2010-02-01T23:59:59.000Z

    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.

  6. Closure Report for Corrective Action Unit 262: Area 25 Septic Systems and Underground Discharge Point, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    D. S. Tobiason

    2003-07-01T23:59:59.000Z

    This Closure Report (CR) documents the activities undertaken to close Corrective Action Unit (CAU) 262: Area 25 Septic Systems and Underground Discharge Point, 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 Corrective Action Plan (CAP) for CAU 262 (U.S. Department of Energy, National Nuclear Security Administration Nevada Operations Office [NNSA/NV, 2002a]). CAU 262 is located at the Nevada Test Site (NTS) approximately 105 kilometers (65 miles) northwest of Las Vegas, Nevada. CAU 262 consists of the following nine Corrective Action Sites (CASs) located in Area 25 of the NTS: 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 CAS 25-51-01, Dry Well.

  7. Corrective Action Decision Document for Corrective Action Unit 135: Area 25 Underground Storage Tanks, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    U.S. Department of Energy, Nevada Operations Office

    1999-12-23T23:59:59.000Z

    This corrective action decision document identifies and rationalizes the US Department of Energy, Nevada Operations Office's selection of a recommended corrective action alternative (CAA) appropriate to facilitate the closure of Corrective Action Unit (CAU) 135, Area 25 Underground Storage Tanks, under the Federal Facility Agreement and Consent Order. Located on the Nevada Test Site (NTS), CAU 135 consists of three Corrective Action Sites (CASs): 25-02-01, Underground Storage Tanks, referred to as the Engine, Maintenance, Assembly, and Disassembly Waste Holdup Tanks and Vault; 25-02-03, Underground Electrical Vault, referred to as the Deluge Valve Pit at the Test Cell A Facility; and 25-02-10, Underground Storage Tank, referred to as the former location of an aboveground storage tank for demineralized water at the Test Cell A Facility. Two of these CASs (25-02-03 and 25-02-10) were originally considered as underground storage tanks, but were found to be misidentified. Further, radio logical surveys conducted by Bechtel Nevada in January 1999 found no radiological contamination detected above background levels for these two sites; therefore, the closure report for CAU 135 will recommend no further action at these two sites. A corrective action investigation for the one remaining CAS (25-02-01) was conducted in June 1999, and analytes detected during this investigation were evaluated against preliminary action levels. It was determined that contaminants of potential concern included polychlorinated biphenyls, Resource Conservation and Recovery Act metals, total petroleum hydrocarbons as diesel-range organics, and radionuclides. Two corrective action objectives were identified for this CAS (i.e., prevention and mitigation of human exposure to sediments and surrounding areas), and subsequently two CAAs developed for consideration based on a review of existing data, future use, and current operations at the NTS. These CAAs were: Alternative 1 - No Further Action, and Alternative 2 - Unrestricted Release Decontamination and Verification Survey. Alternative 2 was chosen as the preferred CAA, after evaluation of technical merit which focused on performance, reliability, feasibility, and safety. This alternative was judged to meet all applicable state and federal regulations for closure of the site and reduces the potential future exposure pathways to the contaminated surfaces at this site.

  8. 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

    2005-06-01T23:59:59.000Z

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

  9. Corrective Action Investigation Plan for Corrective Action Unit 309: Area 12 Muckpiles, Nevada Test Site, Nevada, Rev. No.: 0

    SciTech Connect (OSTI)

    David A. Strand

    2004-12-01T23:59:59.000Z

    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) 309, Area 12 Muckpiles, 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 (DoD). Corrective Action Unit 309 is located in Area 12 of the NTS, which is approximately 65 miles (mi) northwest of Las Vegas, Nevada (Figure 1-1). Area 12 is approximately 40 mi beyond the main gate to the NTS. Corrective Action Unit 309 is comprised of the three Corrective Action Sites (CASs) shown on Figure 1-1 and listed below: CAS 12-06-09, Muckpile; CAS 12-08-02, Contaminated Waste Dump (CWD); and 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. The corrective action investigation (CAI) will include field inspections, radiological surveys, and media sampling, where appropriate. Data will also be obtained to support waste management decisions. The CASs in CAU 309 are being investigated because hazardous and/or radioactive constituents may be present in concentrations that could potentially pose a threat to human health and/or the environment. Existing information on the nature and extent of potential contamination at these sites are insufficient to evaluate and recommend corrective action alternatives for the CASs. Therefore, additional information will be obtained by conducting a CAI prior to evaluating corrective action alternatives and selecting the appropriate corrective action for each CAS.

  10. 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

    2010-09-01T23:59:59.000Z

    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.

  11. Test reports for K Basins vertical fuel handling tools

    SciTech Connect (OSTI)

    Meling, T.A.

    1995-02-01T23:59:59.000Z

    The vertical fuel handling tools, for moving N Reactor fuel elements, were tested in the 305 Building Cold Test Facility (CTF) in the 300 Area. After fabrication was complete, the tools were functionally tested in the CTF using simulated N Reactor fuel rods (inner and outer elements). The tools were successful in picking up the simulated N Reactor fuel rods. These tools were also load tested using a 62 pound dummy to test the structural integrity of each assembly. The tools passed each of these tests, based on the performance objectives. Finally, the tools were subjected to an operations acceptance test where K Basins Operations personnel operated the tool to determine its durability and usefulness. Operations personnel were satisfied with the tools. Identified open items included the absence of a float during testing, and documentation required prior to actual use of the tools in the 100 K fuel storage basin.

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

    SciTech Connect (OSTI)

    Mark Burmeister

    2009-06-01T23:59:59.000Z

    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.

  13. 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

    2005-01-01T23:59:59.000Z

    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.

  14. Corrective Action Investigation Plan for Corrective Action Unit 309: Area 12 Muckpiles, Nevada Test Site, Nevada, Rev. No. 0

    SciTech Connect (OSTI)

    Robert F. Boehlecke

    2004-12-01T23:59:59.000Z

    This Corrective Action Investigation Plan (CAIP) for Corrective Action Unit (CAU) 309, Area 12 Muckpiles, Nevada Test Site (NTS), Nevada, has been developed in accordance with the ''Federal Facility Agreement and Consent Order'' (FFACO) that was agreed to by the State of Nevada, the U.S. Department of Energy, and the U.S. Department of Defense. The general purpose of the investigation is to ensure that adequate data are collected to provide sufficient and reliable information to identify, evaluate, and select technically viable corrective actions. Corrective Action Unit 309 is comprised of the following three corrective action sites (CASs) in Area 12 of the NTS: (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 Site 12-06-09 consists of a muckpile and debris located on the hillside in front of the I-, J-, and K-Tunnels on the eastern slopes of Rainier Mesa in Area 12. The muckpile includes mining debris (muck) and debris generated during the excavation and construction of the I-, J-, and K-Tunnels. Corrective Action Site 12-08-02, CWD, consists of a muckpile and debris and is located on the hillside in front of the re-entry tunnel for K-Tunnel. For the purpose of this investigation CAS 12-28-01 is defined as debris ejected by containment failures during the Des Moines and Platte Tests and the associated contamination that is not covered in the two muckpile CASs. This site consists of debris scattered south of the I-, J-, and K-Tunnel muckpiles and extends down the hillside, across the valley, and onto the adjacent hillside to the south. In addition, the site will cover the potential contamination associated with ''ventings'' along the fault, fractures, and various boreholes on the mesa top and face. One conceptual site model was developed for all three CASs to address possible contamination migration pathways associated with CAU 309. The data quality objective (DQO) process was used to identify and define the type, quantity, and quality of data needed to complete the investigation phase of the corrective action process. The DQO process addresses the primary problem that sufficient information is not available to determine the appropriate corrective action for the CAU. Due to the practical constraints posed by steep slopes on and around the CAU 309 muckpiles, a conservative, simplifying strategy was developed to resolve the presence and nature of contaminants. This strategy includes the use of historical data from similar sites (i.e., previously investigated NTS muckpiles) and the collection of samples from accessible areas of the muckpiles. Based on site history, process knowledge, and previous investigations of similar sites, contaminants of potential concern for CAU 309 collectively include radionuclides, total petroleum hydrocarbons (diesel range only), polychlorinated biphenyls, ''Resource Conservation and Recovery Act'' metals, volatile organic compounds, and semivolatile organic compounds.

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

    SciTech Connect (OSTI)

    DOE /NV

    1998-03-03T23:59:59.000Z

    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.

  16. Effect of improved target designs on the sup 238 Pu production at the Fast Flux Test Reactor

    SciTech Connect (OSTI)

    Karnesky, R.A.; Wootan, D.W.; Jordheim, D.P. (Westinghouse Hanford Co., Richland, WA (United States))

    1991-11-01T23:59:59.000Z

    This paper present the results of a series of calculations made to determine the {sup 238}Pu production potential of several advanced target assembly designs in the Fast Flux Test Facility (FFTF). These calculations show that by using advanced target designs the intimately mix the {sup 237}Np target material with an yttrium hydride moderator, the FFTF has the potential of producing up to 30 kg of high-quality {sup 238}Pu per year.

  17. Corrective Action Investigation Plan for Corrective Action Unit 335: Area 6 Injection Well and Drain Pit, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    DOE/NV

    2000-12-01T23:59:59.000Z

    This Corrective Action Investigation Plan 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) 335, Area 6 Injection Well and Drain Pit, under the Federal Facility Agreement and Consent Order. Corrective Action Unit 335 consists of three Corrective Action Sites (CASs). The CAU is located in the Well 3 Yard in Area 6 at the Nevada Test Site. Historical records indicate that the Drain Pit (CAS 06-23-03) received effluent from truck-washing; the Drums/Oil Waste/Spill (CAS 06-20-01) consisted of four 55-gallon drums containing material removed from the Cased Hole; and the Cased Hole (CAS 06-20-02) was used for disposal of used motor oil, wastewater, and debris. These drums were transported to the Area 5 Hazardous Waste Accumulation Site in July 1991; therefore, they are no longer on site and further investigation or remediation efforts are not required. Consequently, CAS 06-20-01 will be closed with no further action and details of this decision will be described in the Closure Report for this CAU. Any spills that may have been associated with this CAS will be investigated and addressed under CAS 06-20-02. Field investigation efforts will be focused on the two remaining CASs. The scope of the investigation will center around identifying any contaminants of potential concern (COPCs) and, if present, determining the vertical and lateral extent of contamination. The COPCs for the Drain Pit include: total volatile/ semivolatile organic compounds, total petroleum hydrocarbons (gasoline-and diesel-range organics), ethylene glycol monobutyl ether, polychlorinated biphenyls, total Resource Conservation and Recovery Act metals, and radionuclides. The COPCs for the Cased Hole include: total volatile/ semivolatile organic compounds, total petroleum hydrocarbons (diesel-range organics only), and total Resource Conservation an d Recovery Act metals. Both biased surface and subsurface soil sampling will be conducted, augmented by visual inspection, video surveys, and electromagnetic surveys. The results of this field investigation will support a defensible evaluation of corrective action alternatives in the corrective action decision document.

  18. 100-D Area In Situ Redox Treatability Test for Chromate-Contaminated Groundwater: FY 1998 Year-End Report

    SciTech Connect (OSTI)

    Williams, M.D.; Vermeul, V.R.; Szecsody, J.E.; Fruchter, J.S.; Cole, C.R.

    1999-04-15T23:59:59.000Z

    A treatability test was conducted for the In Situ Redox Manipulation (ISRM) technology at the US Department of Energy's Hanford, Washington 100D Area. The target contaminant was dissolved chromate [Cr(VI)] in groundwater. The ISRM technology involves creating a permeable subsurface treatment zone to reduce mobile chromate in groundwater to an insoluble form. The ISRM permeable treatment zone is created by reducing ferric iron [Fe(III)] to ferrous iron [Fe(II)] within the aquifer sediments. This is accomplished by injecting aqueous sodium dithionite into the aquifer and withdrawing the reaction products. The goal of the treatability test was to create a linear ISRM barrier by injecting sodium dithionite into five wells. Well installation and site characterization activities began in the spring of 1997. The first dithionite injection took place in September 1997. The results of this first injection were monitored through the spring of 1998; the remaining four dithionite injections were carried out in May through July of 1998. These five injections created a reduced zone in the Hanford unconfined aquifer 150 feet in length (perpendicular to groundwater flow) by 50 feet wide. The reduced zone extended over the thickness of the unconfined zone, which is approximately 15 feet. Analysis of recent groundwater sampling events shows that the concentrations of chromate [Cr(VI)] in groundwater in the reduced zone have been decreased from starting concentrations of approximately 900 ppb to below analytical detection limits (<7 ppb). Chromate concentrations are also declining in some downgradient monitoring wells. Laboratory analysis of iron in the soil indicates that the barrier should remain in place for approximately 20 to 25 years. These measurements will be confirmed by analysis of sediment cores in FY 1999.

  19. Designing Reactors to Facilitate Decommissioning

    SciTech Connect (OSTI)

    Richard H. Meservey

    2006-06-01T23:59:59.000Z

    Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

  20. Preliminary data from an instantaneous profile test conducted near the Mixed Waste Landfill, Technical Area 3, Sandia National Laboratories/New Mexico

    SciTech Connect (OSTI)

    Bayliss, S.C. [DanShar, Inc., Bosque Farms, NM (United States); Goering, T.J.; McVey, M.D. [GRAM, Inc., Albuquerque, NM (United States); Strong, W.R.; Peace, J.L. [Sandia National Labs., Albuquerque, NM (United States). Environmental Restoration Project

    1996-04-01T23:59:59.000Z

    This paper presents data from an instantaneous profile test conducted near the Sandia National Laboratories/New Mexico Mixed Waste Landfill in Technical Area 3. The test was performed from December 1993 through 1995 as part of the environmental Restoration Project`s Phase 2 RCRA Facility Investigation of the Mixed Waste Landfill. The purpose of the test was to measure the unsaturated hydraulic properties of soils near the Mixed Waste Landfill. The instantaneous profile test and instrumentation are described, and the pressure and moisture content data from the