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Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
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1

Massive Hanford Test Reactor Removed- Plutonium Recycle Test Reactor removed from Hanford’s 300 Area  

Energy.gov (U.S. Department of Energy (DOE))

RICHLAND, WA – Hanford’s River Corridor contractor, Washington Closure Hanford, has met a significant cleanup challenge on the U.S. Department of Energy’s (DOE) Hanford Site by removing a 1,082-ton nuclear test reactor from the 300 Area.

2

An Engineering Test Reactor  

SciTech Connect

A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.

Fahrner, T.; Stoker, R.L.; Thomson, A.S.

1951-03-16T23:59:59.000Z

3

Idaho National Laboratory Advanced Test Reactor Probabilistic...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012...

4

Microstructural Characterization of Test Reactor Irradiated RPV ...  

Science Conference Proceedings (OSTI)

Presentation Title, Microstructural Characterization of Test Reactor Irradiated RPV ... Evolution in High Purity Reference V-4Cr-4Ti Alloy for Fusion Reactor.

5

A Transient Numerical Simulation of Perched Ground-Water Flow at the Test Reactor Area, Idaho National Engineering and Environmental Laboratory, Idaho, 1952-94  

SciTech Connect

Studies of flow through the unsaturated zone and perched ground-water zones above the Snake River Plain aquifer are part of the overall assessment of ground-water flow and determination of the fate and transport of contaminants in the subsurface at the Idaho National Engineering and Environmental Laboratory (INEEL). These studies include definition of the hydrologic controls on the formation of perched ground-water zones and description of the transport and fate of wastewater constituents as they moved through the unsaturated zone. The definition of hydrologic controls requires stratigraphic correlation of basalt flows and sedimentary interbeds within the saturated zone, analysis of hydraulic properties of unsaturated-zone rocks, numerical modeling of the formation of perched ground-water zones, and batch and column experiments to determine rock-water geochemical processes. This report describes the development of a transient numerical simulation that was used to evaluate a conceptual model of flow through perched ground-water zones beneath wastewater infiltration ponds at the Test Reactor Area (TRA).

B. R. Orr (USGS)

1999-11-01T23:59:59.000Z

6

Production test IP-412-AI: B and C reactors export system test  

SciTech Connect

Purpose of this test was to determine the adequacy of the export system for supplying flow to a dual reactor area under simulated emergency conditions.

Benson, J.L.; Jones, S.S.

1961-08-02T23:59:59.000Z

7

Material Science Advances Using Test Reactor Facilities  

Science Conference Proceedings (OSTI)

Aug 2, 2010 ... About this Symposium. Meeting, 2011 TMS Annual Meeting & Exhibition. Symposium, Material Science Advances Using Test Reactor Facilities.

8

Idaho Cleanup Project completes work at Test Area North complex...  

NLE Websites -- All DOE Office Websites (Extended Search)

Idaho Cleanup Project completes work at Test Area North complex at DOEs Idaho site Loss-Of-Fluid Test Reactor Facility (before) Idaho Cleanup Project workers have completed all...

9

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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...

10

Engineering Test Reactor (ETR) Vessel Relocated after 50 years.  

NLE Websites -- All DOE Office Websites (Extended Search)

Printer Friendly Printer Friendly Engineering Test Reactor (ETR) Vessel Relocated Engineering Test Reactor Vessel Pre-startup 1957 Click on image to enlarge. Image 1 of 5 Gantry jacks attached to ETR vessel. Initial lift starts. - Click on image to enlarge. Image 2 of 5 ETR vessel removed from substructure. Vessel lifted approximately 40 ft. - Click on image to enlarge. On Monday, September 24, 2007 the Engineering Test Reactor (ETR) vessel was removed from its location and delivered to the Idaho CERCLA Disposal Facility (ICDF). The long history of the ETR began for this water-cooled reactor with its start up in 1957, after taking only 2 years to build. According to "Proving the Principles," by Susan M. Stacy: When the Engineering Test Reactor started up at the Test Reactor Area in

11

F Reactor Area Cleanup Complete | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

F Reactor Area Cleanup Complete F Reactor Area Cleanup Complete F Reactor Area Cleanup Complete September 19, 2012 - 12:00pm Addthis Media Contact Cameron Hardy, DOE Cameron.Hardy@rl.doe.gov 509-376-5365 RICHLAND, Wash. - U.S. Department of Energy (DOE) contractors have cleaned up the F Reactor Area, the first reactor area at the Hanford Site in southeastern Washington state to be fully remediated. While six of Hanford's nine plutonium production reactors have been sealed up, or cocooned, the F Reactor Area is the first to have all of its associated buildings and waste sites cleaned up in addition to having its reactor sealed up. "The cleanup of the F Reactor Area shows the tremendous progress workers are making along Hanford's River Corridor," said Dave Huizenga, Senior Advisor for the DOE Office of Environmental Management. "The River

12

Ground test facility for nuclear testing of space reactor subsystems  

SciTech Connect

Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs.

Quapp, W.J.; Watts, K.D.

1985-01-01T23:59:59.000Z

13

REACTOR FUEL ELEMENTS TESTING CONTAINER  

DOE Patents (OSTI)

This patent shows a method for detecting leaks in jacketed fuel elements. The element is placed in a sealed tank within a nuclear reactor, and, while the reactor operates, the element is sparged with gas. The gas is then led outside the reactor and monitored for radioactive Xe or Kr. (AEC)

Whitham, G.K.; Smith, R.R.

1963-01-15T23:59:59.000Z

14

TEST REACTORS MEETING FOR INDUSTRY, IDAHO FALLS, IDAHO, MAY 13-15, 1959. PART I. CONSTRUCTION AND OPERATION OF TEST REACTORS. PART II. UTILIZATION OF TEST REACTORS  

SciTech Connect

Twelve papers on construction and operation of test reactors and nine papers on the utilization of test reactors are presented.(W.D.M.)

1959-10-31T23:59:59.000Z

15

CENTRAL NEVPJJA SUPPLEMENTAL TEST AREA  

NLE Websites -- All DOE Office Websites (Extended Search)

r r r r r t r r t r r r * r r r r r r CENTRAL NEVPJJA SUPPLEMENTAL TEST AREA ,FACILITY RECORDS 1970 UNITED STATES ATOMIC ENERGY COMMlSSION NEVADA OPERATIONS OFFICE LAS VEGAS, NEVADA September 1970 Prepared By Holmes & Narver. Inc. On-Continent Test Division P.O. Box 14340 Las Vegas, Nevada 338592 ...._- _._--_ .. -- - - - - - - .. .. - .. - - .. - - - CENTRAL NEVPJJA SUPPLEMENTAL TEST AREA FACILITY RECORDS 1970 This page intentionally left blank - - .. - - - PURPOSE This facility study has been prepared in response to a request of the AEC/NVOO Property Management Division and confirmed by letter, W. D. Smith to L. E. Rickey, dated April 14, 1970, STS Program Administrative Matters. The purpose is to identify each facility, including a brief description, the acquisition cost either purchase and/or construction, and the AE costs if identi- fiable. A narrative review of the history of the subcontracts

16

Advanced Test Reactor National Scientific User Facility  

Science Conference Proceedings (OSTI)

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.

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

17

Advanced Burner Test Reactor - Preconceptual Design Report  

NLE Websites -- All DOE Office Websites (Extended Search)

Burner Test Reactor Preconceptual Design Report ANL-ABR-1 (ANL-AFCI-173) Nuclear Engineering Division Disclaimer This report was prepared as an account of work sponsored by an...

18

INITIAL TESTING AND OPERATION OF THE ARGONNE LOW POWER REACTOR (ALPR)  

SciTech Connect

The major events of a program designed to test and operate the completed reactor power plant and associated equipment are described. The design and construction phases of the project, component installation, preliminary systems testing, zero-power experiments, areas affected by the design parameters, reactor operation, plant safety, and reactor operator training are covered. (W.D.M.)

Hamer, E.E. ed.

1959-12-01T23:59:59.000Z

19

Mechanical Testing of Core Fast Reactor Materials for the Advanced ...  

Science Conference Proceedings (OSTI)

To achieve this goal, the core fast reactor materials (cladding and duct) must be ... in situ Mechanical Test Methods in the US Fusion Reactor Materials Program.

20

SRS Commemorates P & R Reactor Area Completions - SRS Closes...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

SRS Commemorates P & R Reactor Area Completions - SRS Closes the Door on Past Cold War Operations and Opens the Door for Future Missions through Enterprise SRS SRS Commemorates P &...

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

HEAVY WATER COMPONENTS TEST REACTOR DECOMMISSIONING  

Science Conference Proceedings (OSTI)

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.

Austin, W.; Brinkley, D.

2011-10-13T23:59:59.000Z

22

Final safeguards analysis, High Temperature Lattice Test Reactor  

SciTech Connect

Information on the HTLTR Reactor is presented concerning: reactor site; reactor buildings; reactor kinetics and design characteristics; experimental and test facilitles; instrumentation and control; maintenance and modification; initial tests and operations; administration and procedural safeguards; accident analysis; seifterminated excursions; main heat exchanger leak; training program outline; and reliability analysis of safety systems. (7 references) (DCC)

Hanthorn, H.E.; Brown, W.W.; Clark, R.G.; Heineman, R.E.; Humes, R.M.

1966-01-01T23:59:59.000Z

23

Instrumentation to Enhance Advanced Test Reactor Irradiations  

SciTech Connect

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.

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

2009-09-01T23:59:59.000Z

24

Advanced burner test reactor preconceptual design report.  

Science Conference Proceedings (OSTI)

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.

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

25

Reduced enrichment for research and test reactors: Proceedings  

SciTech Connect

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.

1988-05-01T23:59:59.000Z

26

Decommissioning of the Tokamak Fusion Test Reactor  

SciTech Connect

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.

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

2003-10-28T23:59:59.000Z

27

FUNDAMENTALS IN THE OPERATION OF NUCLEAR TEST REACTORS. VOLUME 1. REACTOR SCIENCE AND TECHNOLOGY  

SciTech Connect

A resume of nuclear physics basic to reactor operation precedes discussion of aspects of reactor physics, engineering, chemistry, metallurgy, instrumentation, control, kinetics, and safety. The object is to provide an approach to and understanding of problems in irradiation test programs in the Materials Testing and Engineering Test Reactors. (D.C.W.)

1963-06-01T23:59:59.000Z

28

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Recovery Act Funds Test Reactor Dome Removal in Historic D&D Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project Recovery Act Funds Test Reactor Dome Removal in Historic D&D Project February 1, 2011 - 12:00pm Addthis Media Contacts Jim Giusti, DOE (803) 952-7697 james-r.giusti@srs.gov Paivi Nettamo, SRNS (803) 646-6075 paivi.nettamo@srs.gov AIKEN, S.C. - The landscape of the Savannah River Site (SRS) is a little flatter and a little less colorful with the removal today of the 75-foot-tall rusty-orange dome from the Cold War-era test reactor. This $25-million reactor decommissioning and deactivation project is funded By the American Recovery and Reinvestment Act. Affectionately known by SRS employees as "Hector," the iconic Heavy Water Components Test Reactor (HWCTR) has stood in the Site's B Area since 1959

29

FAST FUEL TEST REACTOR-FFTR CONCEPTUAL DESIGN STUDY  

SciTech Connect

The Fast Fuel Test Reactor (FFTR) is a nuclear facility for the purpose of irradiating samples of fuels and structural components for use in fast reactors. The core consisis of a plate type element in a square configuration. Beryllium metal between the fuel elements is used to obtain a neutron energy spectrum in the hard intermediate region. Cooling of the core and test specimens is accomplished by means of liquid sodium. The design concept was carried through in sufficient degree in the following areas of preliminary concern: number and size of irradiation facilities, sample power requirements, plant layout to evaluate site requirements, plant and nuclear design parameters to evaluate essential equipment requirements. plant-capital-cost estimate, annual- operating-cost estimate, and estimate of construction time schedule. (W.D.M.)

Brubaker, R.; Hummel, H.H.; McArthy, A.; Smaardyk, A.; Kittel, J.H.

1960-08-01T23:59:59.000Z

30

Nuclear reactor containment spray testing system. [PWR  

SciTech Connect

Disclosed is a method for periodic testing of a spray system in a nuclear reactor containment. The method includes injecting a gas into the spray system such that a temperature differential exists between the gas and the containment atmosphere. Scanning the gas jet discharged from the spray nozzles with infrared apparatus then provides a real-time thermal image on a monitor, such as a cathode ray tube, and detects any partially or completely blocked nozzles in the spray system. The scanning may be performed from the containment operating deck. 1 claim, 4 figures.

Rubin, K.

1978-01-10T23:59:59.000Z

31

Overview of Component Testing Requirements for a Small Fluoride Salt-Cooled High Tempreature Reactor  

Science Conference Proceedings (OSTI)

This article summarizes the information necessary to provide reasonable assurance that components for a small fluoride salt-cooled high temperature reactor will meet their functional requirements. In support of the analysis of testing requirements, a simplified, conceptual description of the systems, structures, and components specific to this reactor class was developed. These reactor system elements were divided into major categories based on their functions: (1) reactor core systems, (2) heat transport system, (3) reactor auxiliary cooling system, and (4) instrumentation and controls system. An assessment of technical maturity for each element was made, and a gap analysis was performed to identify specific areas that require further testing. A prioritized list of the testing requirements was then developed. The prioritization was based on both the relative importance of the system to reactor viability, and performance and time requirements to perform the testing.

Cetiner, Mustafa Sacit [ORNL; Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2010-01-01T23:59:59.000Z

32

STATEMENT OF CONSIDERATIONS Advance Test Reactor Class Waiver  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Advance Test Reactor Class Waiver 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 pressure and temperature. The ATR was originally designed to study the effects of intense radiation on reactor material and fuels . It has a "Four Leaf Clover" design that allows a diverse array of testing locations. The unique design allows for different flux in various locations and specialized systems also allow for certain experiments to be run at their own temperature and pressure. The U.S. Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007. This designation will allow the ATR to

33

The Advanced Test Reactor National Scientific User Facility  

Science Conference Proceedings (OSTI)

Symposium, Materials Solutions for the Nuclear Renaissance ... U.S. Department of Energy designated the Advanced Test Reactor (ATR) as a National Scientific ...

34

The Nevada Test Site as a Lunar Analog Test Area  

Science Conference Proceedings (OSTI)

The Nevada Test Site (NTS) is a large (1,350 square miles) secure site currently operated by National Security Technologies, LLC (NSTec), for the U.S. Department of Energy and was established in 1951 to provide a venue for testing nuclear weapons. Three areas with a variety of elevation and geological parameters were used for testing, but the largest number of tests was in Yucca Flat. The Yucca Flat area is approximately 5 miles wide and 20 miles long and approximately 460 subsidence craters resulted from testing in this area. The Sedan crater displaced approximately 12 million tons of earth and is the largest of these craters at 1,280 feet across and 320 feet deep. The profiles of Sedan and the other craters offer a wide variety of shapes and depths that are ideally suited for lunar analog testing.

Sheldon Freid

2007-02-13T23:59:59.000Z

35

Transpiring wall supercritical water oxidation test reactor design report  

Science Conference Proceedings (OSTI)

Sandia National Laboratories is working with GenCorp, Aerojet and Foster Wheeler Development Corporation to develop a transpiring wall supercritical water oxidation reactor. The transpiring wall reactor promises to mitigate problems of salt deposition and corrosion by forming a protective boundary layer of pure supercritical water. A laboratory scale test reactor has been assembled to demonstrate the concept. A 1/4 scale transpiring wall reactor was designed and fabricated by Aerojet using their platelet technology. Sandia`s Engineering Evaluation Reactor serves as a test bed to supply, pressurize and heat the waste; collect, measure and analyze the effluent; and control operation of the system. This report describes the design, test capabilities, and operation of this versatile and unique test system with the transpiring wall reactor.

Haroldsen, B.L.; Ariizumi, D.Y.; Mills, B.E.; Brown, B.G. [Sandia National Labs., Livermore, CA (United States). Engineering for Transportation and Environment Dept.; Rousar, D.C. [GenCorp Aerojet, Sacramento, CA (United States)

1996-02-01T23:59:59.000Z

36

NEUTRONIC REACTOR HAVING LOCALIZED AREAS OF HIGH THERMAL NEUTRON DENSITIES  

DOE Patents (OSTI)

A nuclear reactor for the irradiation of materials designed to provide a localized area of high thermal neutron flux density in which the materials to be irradiated are inserted is described. The active portion of the reactor is comprised of a cubicle graphite moderator of about 25 feet in length along each axis which has a plurality of cylindrical channels for accommodatirg elongated tubular-shaped fuel elements. The fuel elements have radial fins for spacing the fuel elements from the channel walls, thereby providing spaces through which a coolant may be passed, and also to serve as a heatconductirg means. Ducts for accommnodating the sample material to be irradiated extend through the moderator material perpendicular to and between parallel rows of fuel channels. The improvement is in the provision of additional fuel element channels spaced midway between 2 rows of the regular fuel channels in the localized area surrounding the duct where the high thermal neutron flux density is desired. The fuel elements normally disposed in the channels directly adjacent the duct are placed in the additional channels, and the channels directly adjacent the duct are plugged with moderator material. This design provides localized areas of high thermal neutron flux density without the necessity of providing additional fuel material.

Newson, H.W.

1958-06-01T23:59:59.000Z

37

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

Science Conference Proceedings (OSTI)

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.

Vinson, Dennis

2010-06-01T23:59:59.000Z

38

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012 Presenter: Bentley Harwood, Advanced Test Reactor Nuclear Safety Engineer Battelle Energy Alliance Idaho National Laboratory Topics covered: PRA studies began in the late 1980s 1989, ATR PRA published as a summary report 1991, ATR PRA full report 1994 and 2004 various model changes 2011, Consolidation, update and improvement of previous PRA work 2012/2013, PRA risk monitor implementation Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment More Documents & Publications DOE's Approach to Nuclear Facility Safety Analysis and Management Nuclear Regulatory Commission Handling of Beyond Design Basis Events for

39

Space reactor fuel element testing in upgraded TREAT  

DOE Green Energy (OSTI)

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-05-01T23:59:59.000Z

40

Space reactor fuel element testing in upgraded TREAT  

DOE Green Energy (OSTI)

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. initial results suggest that full-scale PBR, elements could be tested at an average energy deposition of {approximately}60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperature limit, average energy deposition of {approximately}100 MW/L may be achievable.

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-01-14T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Testing of Biomass in a Transport Reactor Gasifier  

Science Conference Proceedings (OSTI)

A 200-hour gasification test was undertaken on biomass fuels from sources that include wood waste and a potential energy crop such as switchgrass. The test involved the design and construction of a feed system to allow 100% biomass to be continuously fed to the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center. Biomass performance was also assessed in a high-efficiency transport reactor gasifier, the centerpiece of an advanced biomass integrated ...

2012-11-28T23:59:59.000Z

42

Flow Test At Colrado Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Flow Test At Colrado Area (DOE GTP) Exploration Activity Details Location Colado Geothermal Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding...

43

100 area excavation treatability test plan  

SciTech Connect

This test plan documents the requirements for a treatability study on field radionuclide analysis and dust control techniques. These systems will be used during remedial actions involving excavation. The data from this treatability study will be used to support the feasibility study (FS) process. Development and screening of remedial alternatives for the 100 Area, using existing data, have been completed and are documented in the 100 Area Feasibility Study, Phases 1 and 2 (DOE-RL 1992a). Based on the results of the FS, the Treatability Study Program Plan (DOE-RL 1992b) identifies and prioritizes treatability studies for the 100 Area. The data from the treatability study program support future focused FS, interim remedial measures (IRM) selection, operable unit final remedy selection, remedial design, and remedial actions. Excavation is one of the high-priority, near-term, treatability study needs identified in the program plan (DOE-RL 1992b). Excavation of contaminated soils and buried solid wastes is included in several of the alternatives identified in the 100 Area FS. Although a common activity, excavation has only been used occasionally at the Hanford Site for waste removal applications.

Not Available

1993-05-01T23:59:59.000Z

44

New strategy for accelerating cleanup of Hanford reactor areas  

SciTech Connect

The initial work plans for cleanup of the 100 areas at Hanford followed the traditional Superfund path with a somewhat linear and phased process of investigation and decision making. Due to the complexity of the waste sites, the need to characterize existing mixed waste and hazardous-waste contamination and the need to obtain high-quality data for decision making, the proposed investigation schedules were typically 7 to 9 yr long. In addition, a large amount of resources and funding was committed to this investigative phase without achieving any remediation or reduction in risk. To correct these deficiencies, a new strategy was developed for use at the Hanford site, the Hanford Past Practice Investigation Strategy (HPPIS). In late 1991, work plans were revised to reflect this strategy, and field work is under way. These changes will result in savings in excess of $100,000,000 in remedial investigation/feasibility study (RI/FS) costs. The plutonium production reactor areas at the U.S. Department of Energy (DOE) Hanford site, near Richland, Washington, were included on the U.S. Environmental Protection Agency's (EPA's) National Priorities List (NPL) under the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) in 1989.

Krug, A.D.; Day, R.E.; Lauterbach, M.J. (Westinghouse Hanford Co., Richland, WA (United States))

1992-01-01T23:59:59.000Z

45

EBR-2 (Experimental Breeder Reactor-2) test programs  

SciTech Connect

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

Sackett, J.I.; Lehto, W.K.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA)); Planchon, H.P.; Lambert, J.D.B.; Hill, D.J. (Argonne National Lab., IL (USA))

1990-01-01T23:59:59.000Z

46

Preliminary Advanced Test Reactor LEU Fuel Conversion Feasibility Study  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density, high neutron flux research reactor operating in the United States. The ATR has large irradiation test volumes located in high flux areas. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. As a result, the ATR is a representative candidate for assessing the necessary modifications and evaluating the subsequent operating effects associated with low-enriched uranium (LEU) fuel conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed for the fuel cycle burnup comparison analysis. Using the current HEU 235U enrichment of 93.0 % as a baseline, an analysis can be performed to determine the LEU uranium density and 235U enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the 235U loading in the LEU core, such that the differences in Keff between the HEU and LEU core can be minimized for operation at 150 EFPD with a total core power of 115 MW. The Monte-Carlo with ORIGEN-2 (MCWO) method was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the LEU core conversion designer should be able to optimize the 235U content of each fuel plate, so that the Keff and relative radial fission heat flux profile are similar to the reference ATR HEU case. However, to demonstrate that the LEU core fuel cycle performance can meet the Upgraded Final Safety Analysis Report (UFSAR) safety requirements, a further study will be required in order to investigate the detailed radial, axial, and azimuthal heat flux profile variations versus EFPDs.

G. S. Chang; R. G. Ambrosek

2005-11-01T23:59:59.000Z

47

EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs  

SciTech Connect

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))

1990-01-01T23:59:59.000Z

48

200 Area treated effluent disposal facility operational test report  

Science Conference Proceedings (OSTI)

This document reports the results of the 200 Area Treated Effluent Disposal Facility (200 Area TEDF) operational testing activities. These completed operational testing activities demonstrated the functional, operational and design requirements of the 200 Area TEDF have been met.

Crane, A.F.

1995-03-01T23:59:59.000Z

49

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

Science Conference Proceedings (OSTI)

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.

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

1984-06-01T23:59:59.000Z

50

THE ADVANCED TEST REACTOR-ATR FINAL CONCEPTUAL DESIGN  

SciTech Connect

The results of a study are presented which provided additional experimental-loop irradiation space for the AECDRD testing program. It was a premise that the experiments allocated to this reactor were those which could not be accommodated in the MTR, ETR, or in existing commercial test reactors. To accomplish the design objectives called for a reactor producing perturbed neutron fluxes exceeding 1O/sup 15/ thermal n/cm/sup 2/-sec and 1.5 x 1O/sup 15/ epithermal n/cm/sup 2/-sec. To accommodate the experimental samples, the reactor fuel core is four feet long in the direction of experimental loops. This is twice the length of the MTR core and a third longer than the ETR core. The vertical arrangement of reactor and experiments permits the use of loops penetrating the top cap of the reactor vessel running straight and vertically through the reactor core. The design offers a high degree of accessibility of the exterior portions of the experiments and offers very convenient handling and discharge of experiments. Since the loops are to be integrated into the reactor design and the in-pile portions installed before reactor start-up, it is felt that many of the problems encountered in MTR and ETR experience will cease to exist. Installation of the loops prior to startup will have an added advantage in that the flux variations experienced in experiments in ETR every time a new loop is installed will be absent. The Advanced Test Reactor has a core configuration that provides essentially nine flux-trap regions in a geometry that is almost optimum for cylindrical experiments. The geometry is similar to that of a fourleaf clover with one flux trap in each leaf, one at the intersection of the leaves, and one between each pair of leaves. The nominal power level is 250 Mw. The study was carried out in enough detail to permit the establishment of the design parameters and to develop the power requirement which, conservatively rated, will definitely reach the flux specifications. A critical mockup of an arrangement similar to ATR was loaded into the Engineering Test Reactor Critical Facility. (auth)

deBoisblanc, D.R. et al

1960-11-01T23:59:59.000Z

51

THERMAL PERFORMANCE OF A FAST NEUTRON TEST CONCEPT FOR THE ADVANCED TEST REACTOR  

Science Conference Proceedings (OSTI)

Since 1967, the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL) has provided state-of-the-art experimental irradiation testing capability. A unique design is investigated herein for the purpose of providing a fast neutron flux test capability in the ATR. This new test capability could be brought on line in approximately 5 or 6 years, much sooner than a new test reactor could be built, to provide an interim fast-flux test capability in the timeframe before a fast-flux research reactor could be built. The proposed cost for this system is approximately $63M, much less than the cost of a new fast-flux test reactor. A concept has been developed to filter out a large portion of the thermal flux component by using a thermally conductive neutron absorber block. The objective of this study is to determine the feasibility of this experiment cooling concept.

Donna Post Guillen

2008-06-01T23:59:59.000Z

52

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study 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 light water reactor 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 Risk Informed Safety Margin Characterization (RISMC)

53

Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

(RISMC) Advanced Test (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study 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 light water reactor 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 Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for

54

In-Situ Creep Testing Capability for the Advanced Test Reactor  

Science Conference Proceedings (OSTI)

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.

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

2012-09-01T23:59:59.000Z

55

TEST-HOLE CONSTRUCTION FOR A NEUTRONIC REACTOR  

DOE Patents (OSTI)

Test-hole construction is described for a reactor which provides safe and ready access to the neutron flux region for specimen materials which are to be irradiated therein. An elongated tubular thimble adapted to be inserted in the access hole through the wall of the reactor is constructed of aluminum and is provided with a plurality of holes parallel to the axis of the thimble for conveying the test specimens into position for irradiation, and a conduit for the circulation of coolant. A laminated shield formed of alternate layers of steel and pressed wood fiber is disposed lengthwise of the thimble near the outer end thereof.

Ohlinger, L.A.; Seitz, F.; Young, G.J.

1959-02-17T23:59:59.000Z

56

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

NLE Websites -- All DOE Office Websites (Extended Search)

Light Water Reactor Fuel Cladding Research Light Water Reactor Fuel Cladding Research June 01, 2013 Severe Accident Test Station ORNL is the focus point for Light Water Reactor (LWR) fuel cladding research and testing. The purpose of this research is to furnish U.S. industry (EPRI, Areva, Westinghouse), and regulators (NRC) with much-needed data supporting safe and economical nuclear power generation and used fuel management. LWR fuel cladding work is tightly integrated with ORNL accident tolerant fuel development and used fuel disposition programs thereby providing a powerful capability that couples basic materials science research with the nuclear applications research and development. The ORNL LWR fuel cladding program consists of five complementary areas of research: Accident tolerant fuel and cladding material testing under design

57

200 Area treated effluent disposal facility operational test specification  

Science Conference Proceedings (OSTI)

This document identifies the test specification and test requirements for the 200 Area Treated Effluent Disposal Facility (200 Area TEDF) operational testing activities. These operational testing activities, when completed, demonstrate the functional, operational and design requirements of the 200 Area TEDF have been met.

Crane, A.F.

1995-01-12T23:59:59.000Z

58

200 Area treated effluent disposal facility operational test specification  

Science Conference Proceedings (OSTI)

This document identifies the test specification and test requirements for the 200 Area Treated Effluent Disposal Facility (200 Area TEDF) operational testing activities. These operational testing activities, when completed, demonstrate the functional, operational and design requirements of the 200 Area TEDF have been met.

Crane, A.F.

1995-02-02T23:59:59.000Z

59

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

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Wister Area (DOE GTP) Exploration Activity Details Location Wister Area Exploration...

60

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

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Maui Area (DOE GTP) Exploration Activity Details Location Maui Area Exploration...

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

FUNDAMENTALS IN THE OPERATION OF NUCLEAR TEST REACTORS. VOLUME 2. MATERIALS TESTING REACTOR DESIGN AND OPERATION  

SciTech Connect

The reactor components, building, control system and circuitry, and experimental and handling facilities are described and discussed, together with operation, shutdown, tank work and supplemental facilities. Training questions and answers are included. (D.C.W.)

1963-10-01T23:59:59.000Z

62

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)  

SciTech Connect

The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Test Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.

Totev, T.; Stevens, J.; Kim, Y. S.; Hofman, G.; Matos, J.; Hanan, N.; Garner, P.; Dionne, B.; Olson, A.; Feldman, E.; Dunn, F.; Nuclear Engineering Division; Atomic Research Center; Inst. of Nuclear Physics; LLNL; INL; Korea Atomic Energy Research Inst.; Comisi?n Nacional de Energ?a At?mica; Nuclear Reactor Lab.; Inst. of Atomic Energy-Poland; AECL-Canada; Hungarian Academy of Sciences KFKI Atomic Energy Research Inst.; Japan Atomic Energy Agency; Nuclear Power Inst. of China; Kyoto Univ. Research Reactor Inst.

2010-03-01T23:59:59.000Z

63

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

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).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

64

Heavy Water Components Test Reactor Decommissioning - Major Component Removal  

SciTech Connect

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).

Austin, W.; Brinkley, D.

2010-05-05T23:59:59.000Z

65

Aerial Photography 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: Aerial Photography At Nevada Test And...

66

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...

67

Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor  

Science Conference Proceedings (OSTI)

Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

2006-10-01T23:59:59.000Z

68

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

SciTech Connect

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.

1993-07-01T23:59:59.000Z

69

Enhanced In-Pile Instrumentation at the Advanced Test Reactor  

SciTech Connect

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.

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

2012-08-01T23:59:59.000Z

70

Enhanced In-Pile Instrumentation at the Advanced Test Reactor  

Science Conference Proceedings (OSTI)

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.

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

2011-06-01T23:59:59.000Z

71

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

SciTech Connect

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.

Whitehead, Donnie Wayne; Varnado, G. Bruce

2008-09-01T23:59:59.000Z

72

The Advanced Test Reactor National Scientific User Facility  

Science Conference Proceedings (OSTI)

In 2007, the Advanced Test Reactor (ATR), located at Idaho National Laboratory (INL), was designated by the Department of Energy (DOE) 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 approved researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide those 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, obtained access to additional PIE equipment, taken steps to enable the most advanced post-irradiation analysis possible, and initiated an educational program and digital learning library to help potential users better understand the critical issues in reactor technology and how a test reactor facility could be used to address this critical research. 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 invited universities to nominate their capability to become part of a broader user facility. Any university is eligible to self-nominate. Any nomination is then peer reviewed to ensure that the addition of the university facilities adds useful capability to the NSUF. Once added to the NSUF team, the university capability is then integral to the NSUF operations and is available to all users via the proposal process. So far, six universities 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 university capabilities, 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. The current NSUF partners are shown in Figure 1. This article describes the ATR as well as the expanded capabilities, partnerships, and services that allow researchers to take full advantage of this national resource.

Todd R. Allen; Collin J. Knight; Jeff B. Benson; Frances M. Marshall; Mitchell K. Meyer; Mary Catherine Thelen

2011-08-01T23:59:59.000Z

73

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

SciTech Connect

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.

Susan Stacy; Hollie K. Gilbert

2005-02-01T23:59:59.000Z

74

SRS Commemorates P & R Reactor Area Completions - SRS Closes the Door on  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Commemorates P & R Reactor Area Completions - SRS Closes the Commemorates P & R Reactor Area Completions - SRS Closes the Door on Past Cold War Operations and Opens the Door for Future Missions through Enterprise SRS SRS Commemorates P & R Reactor Area Completions - SRS Closes the Door on Past Cold War Operations and Opens the Door for Future Missions through Enterprise SRS September 29, 2011 - 12:00pm Addthis Media Contacts Jim Giusti, DOE james-r.giusti@srs.gov 803-952-7697 Paivi Nettamo, SRNS paivi.nettamo@srs.gov 803-952-6938 AIKEN, S.C. - Today, the U.S. Department of Energy (DOE) Savannah River celebrated the cleanup and closure of the P and R Reactor Areas under the American Recovery and Reinvestment Act program at the Savannah River Site and announced the vision for future missions at the site. Acting Assistant Secretary for the DOE's Office of Environmental

75

Advanced Test Reactor National Scientific User Facility Partnerships  

SciTech Connect

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.

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

2012-03-01T23:59:59.000Z

76

Development of a Fissile Materials Irradiation Capability for Advanced Fuel Testing at the MIT Research Reactor  

SciTech Connect

A fissile materials irradiation capability has been developed at the Massachusetts Institute of Technology (MIT) Research Reactor (MITR) to support nuclear engineering studies in the area of advanced fuels. The focus of the expected research is to investigate the basic properties of advanced nuclear fuels using small aggregates of fissile material. As such, this program is intended to complement the ongoing fuel evaluation programs at test reactors. Candidates for study at the MITR include vibration-packed annular fuel for light water reactors and microparticle fuels for high-temperature gas reactors. Technical considerations that pertain to the design of the MITR facility are enumerated including those specified by 10 CFR 50 concerning the definition of a research reactor and those contained in a separate license amendment that was issued by the U.S. Nuclear Regulatory Commission to MIT for these types of experiments. The former includes limits on the cross-sectional area of the experiment, the physical form of the irradiated material, and the removal of heat. The latter addresses experiment reactivity worth, thermal-hydraulic considerations, avoidance of fission product release, and experiment specific temperature scrams.

Hu Linwen; Bernard, John A.; Hejzlar, Pavel; Kohse, Gordon [Massachusetts Institute of Technology (United States)

2005-05-15T23:59:59.000Z

77

Nevada Test And Training Range Geothermal Area | Open Energy Information  

Open Energy Info (EERE)

Nevada Test And Training Range Geothermal Area Nevada Test And Training Range Geothermal Area (Redirected from Nevada Test And Training Range Area) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Geothermal Resource Area: Nevada Test And Training Range Geothermal Area Contents 1 Area Overview 2 History and Infrastructure 3 Regulatory and Environmental Issues 4 Exploration History 5 Well Field Description 6 Geology of the Area 7 Geofluid Geochemistry 8 NEPA-Related Analyses (0) 9 Exploration Activities (5) 10 References Area Overview Geothermal Area Profile Location: Nevada Exploration Region: Northern Basin and Range Geothermal Region GEA Development Phase: 2008 USGS Resource Estimate Mean Reservoir Temp: Estimated Reservoir Volume: Mean Capacity: Click "Edit With Form" above to add content

78

200 area effluent treatment facility opertaional test report  

Science Conference Proceedings (OSTI)

This document reports the results of the 200 Area Effluent Treatment Facility (200 Area ETF) operational testing activities. These Operational testing activities demonstrated that the functional, operational and design requirements of the 200 Area ETF have been met and identified open items which require retesting.

Crane, A.F.

1995-10-26T23:59:59.000Z

79

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

SciTech Connect

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).

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

2011-09-01T23:59:59.000Z

80

Nevada Test And Training Range Geothermal Area | Open Energy Information  

Open Energy Info (EERE)

Page Page Edit with form History Facebook icon Twitter icon » Nevada Test And Training Range Geothermal Area Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Geothermal Resource Area: Nevada Test And Training Range Geothermal Area Contents 1 Area Overview 2 History and Infrastructure 3 Regulatory and Environmental Issues 4 Exploration History 5 Well Field Description 6 Geology of the Area 7 Geofluid Geochemistry 8 NEPA-Related Analyses (0) 9 Exploration Activities (5) 10 References Area Overview Geothermal Area Profile Location: Nevada Exploration Region: Northern Basin and Range Geothermal Region GEA Development Phase: 2008 USGS Resource Estimate Mean Reservoir Temp: Estimated Reservoir Volume: Mean Capacity: Click "Edit With Form" above to add content

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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81

DIAMOND WIRE CUTTING OF THE TOKAMAK FUSION TEST REACTOR  

Science Conference Proceedings (OSTI)

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 Mev neutrons. The total tritium content within the vessel is in excess of 7,000 Curies while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the Tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the techno logy was improved and redesigned for the actual cutting of the vacuum vessel. 10 complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D activity.

Rule, Keith; Perry, Erik; Parsells, Robert

2003-02-27T23:59:59.000Z

82

Diamond Wire Cutting of the Tokamak Fusion Test Reactor  

Science Conference Proceedings (OSTI)

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a-kind, tritium-fueled fusion research reactor that ceased operation in April 1997. As a result, decommissioning commenced in October 1999. The 100 cubic meter volume of the donut-shaped reactor makes it the second largest fusion reactor in the world. The deuterium-tritium experiments resulted in contaminating the vacuum vessel with tritium and activating the materials with 14 MeV neutrons. The total tritium content within the vessel is in excess of 7,000 Curies, while dose rates approach 50 mRem/hr. These radiological hazards along with the size of the tokamak present a unique and challenging task for dismantling. Engineers at the Princeton Plasma Physics Laboratory (PPPL) decided to investigate an alternate, innovative approach for dismantlement of the TFTR vacuum vessel: diamond wire cutting technology. In August 1999, this technology was successfully demonstrated and evaluated on vacuum vessel surrogates. Subsequently, the technology was improved and redesigned for the actual cutting of the vacuum vessel. Ten complete cuts were performed in a 6-month period to complete the removal of this unprecedented type of D&D (Decontamination and Decommissioning) activity.

Keith Rule; Erik Perry; Robert Parsells

2003-01-31T23:59:59.000Z

83

Nevada Test Site Area 25. Radiological survey and cleanup project, 1974-1983. Final report  

SciTech Connect

This report describes radiological survey, decontamination and decommissioning of the Nevada Test Site (NTS) Area 25 facilities and land areas incorporated in the Nuclear Rocket Development Station (NRDS). Buildings, facilities and support systems used after 1959 for nuclear reactor and engine testing were surveyed for the presence of radioactive contamination. The cleanup was part of the Surplus Facilities Management Program funded by the Department of Energy's Richland Operations Office. The radiological survey portion of the project encompassed portable instrument surveys and removable contamination surveys (swipe) for alpha and beta plus gamma radiation contamination of facilities, equipment and land areas. Soil sampling was also accomplished. The majority of Area 25 facilities and land areas have been returned to unrestricted use. Remaining radiologically contaminated areas are posted with warning signs and barricades. 12 figures.

McKnight, R.K.; Rosenberry, C.E.; Orcutt, J.A.

1984-01-01T23:59:59.000Z

84

L-Area Reactor - 1993 annual - groundwater monitoring report  

Science Conference Proceedings (OSTI)

Groundwater was sampled and analyzed during 1993 from wells monitoring the water table at the following locations in L Area: the L-Area Acid/Caustic Basin (four LAC wells), L-Area Research Wells in the southern portion of the area (outside the fence; three LAW wells), the L-Area Oil and Chemical Basin (four LCO wells), the L-Area Disassembly Basin (two LDB wells), the L-Area Burning/Rubble Pit (four LRP wells), and the L-Area Seepage Basin (four LSB wells). During 1993, tetrachloroethylene was detected above its drinking water standard (DWS) in the LAC, LAW, LCO, and LDB well series. Lead exceeded its 50 {mu}g/L standard in the LAW, LDB, and LRP series, and tritium was above its DWS in the LAW, LCO, and LSB series. Apparently anomalous elevated levels of the common laboratory contaminant bis(2-ethylhexyl)phthalate were reported during first quarter in one well each in the LAC series and LCO series, and during third quarter in a different LCO well. Extensive radionuclide analyses were performed during 1993 in the LAC, LAW, and LCO well series. No radionuclides other than tritium were reported above DWS or Flag 2 criteria.

Chase, J.A.

1994-09-01T23:59:59.000Z

85

Advanced Test Reactor National Scientific User Facility Progress  

SciTech Connect

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.

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

2012-10-01T23:59:59.000Z

86

Injectivity Test At Fenton Hill Hdr Geothermal Area (Grigsby...  

Open Energy Info (EERE)

navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Injectivity Test At Fenton Hill Hdr Geothermal Area (Grigsby, Et Al., 1983) Exploration Activity Details...

87

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

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Alum Geothermal Area (DOE GTP) Exploration Activity Details Location Alum Geothermal...

88

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

Open Energy Info (EERE)

Page Edit History Facebook icon Twitter icon Geodetic Survey At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Jump to: navigation, search GEOTHERMAL...

89

Test storage of spent reactor fuel in the Climax granite at the Nevada Test Site  

SciTech Connect

A test of retrievable dry geologic storage of spent fuel assemblies from an operating commercial nuclear reactor is underway at the Nevada Test Site. This generic test is located 420 m below the surface in the Climax granitic stock. Eleven canisters of spent fuel approximately 2.3 years out of reactor core (about 2 kW/canister thermal output) will be emplaced in a storage drift along with 6 electrical simulator canisters and their effects will be compared. Two adjacent drifts will contain electrical heaters, which will be operated to simulate within the test array the thermal field of a large repository. The test objectives, technical concepts and rationale, and details of the test are stated and discussed.

Ramspott, L.D.; Ballou, L.B.

1980-02-13T23:59:59.000Z

90

Injectivity Test At Raft River Geothermal Area (1979) | Open Energy  

Open Energy Info (EERE)

Injectivity Test At Raft River Geothermal Area (1979) Injectivity Test At Raft River Geothermal Area (1979) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Injectivity Test At Raft River Geothermal Area (1979) Exploration Activity Details Location Raft River Geothermal Area Exploration Technique Injectivity Test Activity Date 1979 Usefulness useful DOE-funding Unknown Notes Quantification of the pressure response prior to 600 minutes is not always possible. Short-duration (< 24-hour) injection or pump tests are conducted with the drilling rig equipment, and long-duration (21-day) injection and pump tests are then conducted with the permanent pumping facilities. References Allman, D. W.; Goldman, D.; Niemi, W. L. (1 January 1979) Evaluation of testing and reservoir parameters in geothermal wells at Raft

91

Geographic Information System At Nevada Test And Training Range Area  

Open Energy Info (EERE)

Geographic Information System At Nevada Test And Training Range Area Geographic Information System At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Geographic Information System At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Exploration Activity Details Location Nevada Test And Training Range Area Exploration Technique Geographic Information System Activity Date Usefulness not indicated DOE-funding Unknown Notes Nellis Air Force Range (NAFR) occupies over 3 million acres in southern Nevada (Figure 1). We recently assessed potential utility-grade geothermal resources and possible target areas for exploration by constructing a GIS of this area and applying the occurrence model ideas outlined above (ITSI, 2003; Sabin et al., 2004). We list below many of the factors considered.

92

Technology issues for decommissioning the Tokamak Fusion Test Reactor  

SciTech Connect

The approach for decommissioning the Tokamak Fusion Test Reactor has evolved from a conservative plan based on cutting up and burying all of the systems, to one that considers the impact tritium contamination will have on waste disposal, how large size components may be used as their own shipping containers, and even the possibility of recycling the materials of components such as the toroidal field coils and the tokamak structure. In addition, the project is more carefully assessing the requirements for using remotely operated equipment. Finally, valuable cost database is being developed for future use by the fusion community.

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

1994-07-01T23:59:59.000Z

93

Tracer Testing At East Mesa Geothermal Area (1983) | Open Energy  

Open Energy Info (EERE)

Tracer Testing At East Mesa Geothermal Area (1983) Tracer Testing At East Mesa Geothermal Area (1983) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Tracer Testing At East Mesa Geothermal Area (1983) Exploration Activity Details Location East Mesa Geothermal Area Exploration Technique Tracer Testing Activity Date 1983 Usefulness not indicated DOE-funding Unknown Notes Two field experiments were conducted to develop chemical tracer procedures for use with injection-backflow testing, one on the fracture-permeability Raft River reservoir and the other on the matrix-permeability East Mesa reservoir. Results from tests conducted with incremental increases in the injection volume at both East Mesa and Raft River suggests that, for both reservoirs, permeability remained uniform with increasing distance from the

94

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

SciTech Connect

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.

A. B. Culp

2007-01-26T23:59:59.000Z

95

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

Science Conference Proceedings (OSTI)

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)

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

96

Run - Beyond - Cladding - Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program  

Science Conference Proceedings (OSTI)

In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab.

Batte, G.L. (Argonne National Lab., Idaho Falls, ID (USA)); Hoffman, G.L. (Argonne National Lab., IL (USA))

1990-01-01T23:59:59.000Z

97

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

Open Energy Info (EERE)

Stress Test At Coso Geothermal Area (2004) Stress Test At Coso Geothermal Area (2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Stress Test At Coso Geothermal Area (2004) Exploration Activity Details Location Coso Geothermal Area Exploration Technique Stress Test Activity Date 2004 Usefulness not indicated DOE-funding Unknown Exploration Basis EGS potential of Coso Geothermal Region Notes A hydraulic fracturing stress test at 3,703 feet TVD was used to constrain a normal faulting and strike-slip faulting stress tensor for this reservoir. The shear and normal stresses resolved on the fracture and fault planes were calculated and used to identify the subset of critically stressed planes that act to maintain permeability within the Coso Geothermal Field. References

98

Safety Assurance for Irradiating Experiments in the Advanced Test Reactor  

SciTech Connect

The Advanced Test Reactor (ATR), located at the Idaho National Engineering and Environmental Laboratory (INEEL), was specifically designed to provide a high neutron flux test environment for conducting a variety of experiments. This paper addresses the safety assurance process for two general types of experiments conducted in the ATR facility and how the safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore, this type of experiment is addressed in more detail in the ATR safety basis. This allows the individual safety analysis for this type of experiment to be more standardized. The second type of experiment is defined in more general terms in the ATR safety basis and is permitted under more general controls. Therefore, the individual safety analysis for the second type of experiment tends to be more unique and is tailored to each experiment.

T. A. Tomberlin; S. B. Grover

2004-11-01T23:59:59.000Z

99

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

E-Print Network (OSTI)

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 ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

100

Advanced LWR Fuel Testing Capabilities in the ORNL High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

A new test capability for the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) is being developed that will allow testing of advanced nuclear fuels and cladding materials under prototypic light-water reactor (LWR) operating conditions in less time than it takes in other research reactors. This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiments currently planned to start in late 2008.

Ott, Larry J [ORNL; McDuffee, Joel Lee [ORNL; Spellman, Donald J [ORNL

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Flow Test At Coso Geothermal Area (1978) | Open Energy Information  

Open Energy Info (EERE)

Flow Test At Coso Geothermal Area (1978) Flow Test At Coso Geothermal Area (1978) Exploration Activity Details Location Coso Geothermal Area Exploration Technique Flow Test Activity Date 1978 Usefulness not indicated DOE-funding Unknown Notes Flow tests of well CGEH No. 1 were conducted. LBL performed eight temperature surveys after completion of the well to estimate equilibrium reservoir temperatures. Downhole fluid samples were obtained by the U.S. Geological Survey (USGS) and Lawrence Berkeley Laboratory (LBL), and a static pressure profile was obtained. The first test began September 5, 1978 using nitrogen stimulation to initiate flow; this procedure resulted in small flow and subsequent filling of the bottom hole with drill cuttings. The second test, on November 2, 1978, utilized a nitrogen-foam-water mixture to clean residual particles from bottom hole,

102

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

SciTech Connect

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

P.C. Weaver

2010-12-15T23:59:59.000Z

103

Tracer Testing At Coso Geothermal Area (1993) | Open Energy Information  

Open Energy Info (EERE)

Tracer Testing At Coso Geothermal Area (1993) Tracer Testing At Coso Geothermal Area (1993) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Tracer Testing At Coso Geothermal Area (1993) Exploration Activity Details Location Coso Geothermal Area Exploration Technique Tracer Testing Activity Date 1993 Usefulness useful DOE-funding Unknown Exploration Basis To determine the steam and water mass flow rate Notes The method involves precisely metered injection of liquid and vapor phase tracers into the two-phase production pipeline and concurrent sampling of each phase downstream of the injection point. Subsequent chemical analysis of the steam and water samples for tracer content enables the calculation of mass flowrate for each phase given the known mass injection rates of

104

Tracer Testing At Coso Geothermal Area (2006) | Open Energy Information  

Open Energy Info (EERE)

Tracer Testing At Coso Geothermal Area (2006) Tracer Testing At Coso Geothermal Area (2006) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Tracer Testing At Coso Geothermal Area (2006) Exploration Activity Details Location Coso Geothermal Area Exploration Technique Tracer Testing Activity Date 2006 Usefulness useful DOE-funding Unknown Exploration Basis To characterize the flow patterns of fluid injected into well 68-20RD. Notes A conservative liquid phase tracer, 2-naphthalene sulfonate, and a two-phase tracer, ethanol, were injected into well 68-20RD. Surrounding production wells were sampled over the subsequent 125 days and analyzed for the two tracers. The results demonstrate the efficacy of the simultaneous use of liquid-phase and two-phase tracers in fluid-depleted geothermal

105

PRA insights applicable to the design of the Broad Applications Test Reactor  

SciTech Connect

Design insights applicable to the design of a new Broad Applications Test Reactor (BATR), being studied at Idaho National Engineering Laboratory, are summarized. Sources of design insights include past probabilistic risk assessments and related studies for department of Energy-owned Class A reactors and for commercial reactors. The report includes a preliminary risk allocation scheme for the BATR.

Khericha, S.T.; Reilly, H.J.

1993-01-01T23:59:59.000Z

106

Implementation of a testing and diagnostic concept for an NPP reactor protection system  

Science Conference Proceedings (OSTI)

This paper presents the concept and practical realization of the testing and diagnostic methodology for a reactor protection system in a nuclear power plant. The test concept utilizes the highly redundant nature of these systems to conduct tests during ...

Tamás Bartha; István Varga; Alexandros Soumelidis; Géza Szabé

2005-04-01T23:59:59.000Z

107

Testing of an advanced thermochemical conversion reactor system  

DOE Green Energy (OSTI)

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.

Not Available

1990-01-01T23:59:59.000Z

108

Enhanced In-pile Instrumentation for Material Testing Reactors  

Science Conference Proceedings (OSTI)

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.

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

2012-07-01T23:59:59.000Z

109

Tracer Testing At Raft River Geothermal Area (1983) | Open Energy  

Open Energy Info (EERE)

3) 3) Exploration Activity Details Location Raft River Geothermal Area Exploration Technique Tracer Testing Activity Date 1983 Usefulness not indicated DOE-funding Unknown Exploration Basis To develop chemical tracing procedures for geothermal areas. Notes Two field experiments were conducted to develop chemical tracer procedures for use with injection-backflow testing, one on the fracture-permeability Raft River reservoir and the other on the matrix-permeability East Mesa reservoir. Results from tests conducted with incremental increases in the injection volume at both East Mesa and Raft River suggests that, for both reservoirs, permeability remained uniform with increasing distance from the well bore. Increased mixing during quiescent periods, between injection and

110

High uranium density dispersion fuel for the reduced enrichment of research and test reactors program.  

E-Print Network (OSTI)

??This work describes the fabrication of a high uranium density fuel for the Reduced Enrichment of Research and Test Reactors Program. In an effort to… (more)

[No author

2006-01-01T23:59:59.000Z

111

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

SciTech Connect

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

Sarah Roberts

2006-10-18T23:59:59.000Z

112

Fast Flux Test Reactor: Re-evaluation of the Department's Approach...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Sites Power Marketing Administration Other Agencies You are here Home Fast Flux Test Reactor: Re-evaluation of the Department's Approach to Deactivation, Decontamination,...

113

Deterministic Modeling of the High Temperature Test Reactor  

SciTech Connect

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.

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

2010-06-01T23:59:59.000Z

114

Home Area Network Test, Evaluation, and Diagnostics Tools  

Science Conference Proceedings (OSTI)

This report provides a detailed description of a set of tools that may be used for evaluation of the communication quality of Home Area Networks (HANs). The development of these tools is a continuation of work begun in 2008 and documented in EPRI report 1020106, "Home Area Network Performance Metrics and Monitoring." This report presents advancements and modifications made during 2011 to extend the capabilities of these test tools for a range of applications. This report also introduces hardware and soft...

2011-12-20T23:59:59.000Z

115

Marysville Test Well Geothermal Area | Open Energy Information  

Open Energy Info (EERE)

Test Well Geothermal Area Test Well Geothermal Area Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Geothermal Resource Area: Marysville Test Well Geothermal Area Contents 1 Area Overview 2 History and Infrastructure 3 Regulatory and Environmental Issues 4 Exploration History 5 Well Field Description 6 Geology of the Area 7 Geofluid Geochemistry 8 NEPA-Related Analyses (0) 9 Exploration Activities (0) 10 References Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"TERRAIN","zoom":6,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"500px","height":"300px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":46.75333333,"lon":-112.3766667,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

116

Hot-Gas Filter Testing with a Transport Reactor Gasifier  

Science Conference Proceedings (OSTI)

Today, coal supplies over 55% of the electricity consumed in the United States and will continue to do so well into the next century. One of the technologies being developed for advanced electric power generation is an integrated gasification combined cycle (IGCC) system that converts coal to a combustible gas, cleans the gas of pollutants, and combusts the gas in a gas turbine to generate electricity. The hot exhaust from the gas turbine is used to produce steam to generate more electricity from a steam turbine cycle. The utilization of advanced hot-gas particulate and sulfur control technologies together with the combined power generation cycles make IGCC one of the cleanest and most efficient ways available to generate electric power from coal. One of the strategic objectives for U.S. Department of Energy (DOE) IGCC research and development program is to develop and demonstrate advanced gasifiers and second-generation IGCC systems. Another objective is to develop advanced hot-gas cleanup and trace contaminant control technologies. One of the more recent gasification concepts to be investigated is that of the transport reactor gasifier, which functions as a circulating fluid-bed gasifier while operating in the pneumatic transport regime of solid particle flow. This gasifier concept provides excellent solid-gas contacting of relatively small particles to promote high gasification rates and also provides the highest coal throughput per unit cross-sectional area of any other gasifier, thereby reducing capital cost of the gasification island.

Swanson, M.L.; Hajicek, D.R.

2002-09-18T23:59:59.000Z

117

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

E-Print Network (OSTI)

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.

Norman, Eric B; Chodash, Perry A

2011-01-01T23:59:59.000Z

118

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

E-Print Network (OSTI)

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 – 131,132 I, 132 Te, and 134,137 Cs. The activity levels we have measured for these isotopes are very low and pose no health risk to the public.

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

2011-01-01T23:59:59.000Z

119

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

E-Print Network (OSTI)

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.

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

2011-03-30T23:59:59.000Z

120

Radiation Portal Monitoring Test Area and Large Detector  

E-Print Network (OSTI)

a transformation. In the summer of 2007, work began on the nearly 200,000-square-foot Physical Sciences Facility, equipment and staff displaced from accelerated cleanup of Hanford's 300 Area. This federally financed detection and testing · Borderandinterdiction technology · Materialsdevelopmentand engineering

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

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

SciTech Connect

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.

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

122

MATERIALS TESTING REACTOR-ENGINEERING TEST REACTOR TECHNICAL BRANCHES. Quarterly Report No. 3, July 1-September 30, 1963  

SciTech Connect

8 6 < platelets containing U/sub 3/O/sub 8/, UO/sub 2/, or UAl/sub 3/ in aluminum matrices were irradiated in the ETR at inltial surface temperatures of 180 deg C to burnups of 1 x 10/sup 21/ fiss/ cm/sup 3/. The high fuel loadings (approximately 35 wt% U/sup 235/) in UO/sub 2/ and U/sub 3/O/sub 8/ blistered under these conditions; the UAl/sub samples were still in good condition at the end of the test. Electrolyzed coatings on aluminum deteriorated badly under exposures of 3 to 5 x 10/sup 20/ n/cm/sup 2/ (>1Mev) in the ETR process water. The ARMF-1 regulating rod was repaired and digital regulating rod position readout instrumentation installed during an extended shutdown after more than two years of operation. Fission product transient curves extrapolated to about the same zero time reactivity value with initial data varying from 30 minutes to 6 hours. This limited considerably the probability that short-lived high cross section fission products exist. Under present ETR operating conditions the maximum decrease in effectiveness of a nickel absorber section as a result of burnup would be less than 20% in 20 years. Thus, burnup appears not to be a factor which limits its useful life. The preliminary analysis and flow charts for the Phillips General Purpose Monte Carlo Program for the IBM 7040 are nearing completion. Two reactor simulation devices were put into service in the Analog Computer Facility, a reactor kinetics simulator and seven transport lag simulation channels. Preliminary design of a Xe-135 simulator was completed. The fission cross section of Pu/sup 241/ was measured from 2 to 100 ev. Resolution of the linear electron accelerator used was sufficient to permit multilevel analysis of the neutron levels below 36 ev. Transmission measurements were obtained on a separated Pa/sup 233/ sample containing approximately 10 mg of Pa/sup 2/O/sub 5/. The energies of these resonances observed with the unseparated sample with their relative sizes are presented. Several experiments were conducted to determine the useful lifetime of solid state detectors under in- pile conditions of fission fragment bombardment. A single detector, using an external U/sup 235/ fission source was irradiated to approximately 3 x 10/sup 9/ total fission, at which point the fission fragment peaks were still well resolved and the signal pulses were sufficientiy large compared to noise level so that the latter could be effectively biased out. Averaged reduced partial differential scatiering cross sections for a powder Be sample were obtained. A Data Processing System for transient data was developed for use in the SPERT reactor complex. Data are recorded on an FM tape system and applied to a magnetic memory for temporary storage and from there to one or more of several readout devices. An Eight-Input Adapter and an Initial Delay Counter were developed to increase the utility of an existing time-of-flight analyzer. A Personnel Monitor ( Frisker'') is described, which approaches closely an ideal monitor for use with widely varying radiation backgrounds. Current feedback around an operational amplifier is used to provide a current source used to drive oscillograph galvanometers thereby extending the range of linear operation of the galvanometers. The work of placing a large telemetered radiological survey system in operation is described along with the description of a remote station simulator. Dynamic pressure tests of several commercial transducers are described together with the criteria established for suitability for their use in reactor transient studies. Rod drop deceleration times were measured on an ETR control rod; the test instrumentation is described. The 7090 version of PDQ-4 (20,000 mesh points) was converted and modified for operation on the 7040. The following reactor codes are also now in operation on the 7040: TEMPEST-II, GAM, FOG, ZUT, MIST, ULCER, and TOPIC. Being de-bugged are HEAT-I and IREKIN. In addition, the following programs for the 7040 were written and placed in operation: matrix inversion, ordinary differ

1964-02-15T23:59:59.000Z

123

Coupled thermohydromechanical analysis of a heater test in unsaturated clay and fractured rock at Kamaishi Mine  

E-Print Network (OSTI)

Instrumentation. Power reactor and nuclear fiel developmentexperiment area. Power reactor and nuclear fbel developmentHydraulic tests. Power reactor and nuclear fiel development

Rutqvist, J.

2011-01-01T23:59:59.000Z

124

IRRADIATION TESTING OF THE RERTR FUEL MINIPLATES WITH BURNABLE ABSORBERS IN THE ADVANCED TEST REACTOR  

SciTech Connect

Based on the results of the reactor physics assessment, conversion of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) can be potentially accomplished in two ways, by either using U-10Mo monolithic or U-7Mo dispersion type plates in the ATR fuel element. Both designs, however, would require incorporation of the burnable absorber in several plates of the fuel element to compensate for the excess reactivity and to flatten the radial power profile. Several different types of burnable absorbers were considered initially, but only borated compounds, such as B4C, ZrB2 and Al-B alloys, were selected for testing primarily due to the length of the ATR fuel cycle and fuel manufacturing constraints. To assess and compare irradiation performance of the U-Mo fuels with different burnable absorbers we have designed and manufactured 28 RERTR miniplates (20 fueled and 8 non-fueled) containing fore-mentioned borated compounds. These miniplates will be tested in the ATR as part of the RERTR-13 experiment, which is described in this paper. Detailed plate design, compositions and irradiations conditions are discussed.

I. Glagolenko; D. Wachs; N. Woolstenhulme; G. Chang; B. Rabin; C. Clark; T. Wiencek

2010-10-01T23:59:59.000Z

125

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

DOE Green Energy (OSTI)

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).

D. S. Tobiason

2002-03-01T23:59:59.000Z

126

Tracer Testing At Coso Geothermal Area (2004) | Open Energy Information  

Open Energy Info (EERE)

Coso Geothermal Area (2004) Coso Geothermal Area (2004) Exploration Activity Details Location Coso Geothermal Area Exploration Technique Tracer Testing Activity Date 2004 Usefulness not indicated DOE-funding Unknown Exploration Basis To determine the EGS potential of the Coso Geothermal Field Notes A dramatic decrease in the ratio of chloride to boron was observed in the liquid discharge of a well proposed for EGS development. The decrease appears to be related to the transformation of some feed zones in the well from liquid-dominated to vapor-dominated. High concentrations of boron are transported to the wellbore in the steam, where it fractionates to the liquid phase flowing in from liquid-dominated feed zones. The high-boron steam is created when the reservoir liquid in some of the feed zones boils

127

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

Science Conference Proceedings (OSTI)

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

Ryskamp, J.M.; Sterbentz, J.W.; Chang, G.S. [and others

1995-09-01T23:59:59.000Z

128

An Aerial Radiological Survey of Selected Areas of Area 18 - Nevada Test Site  

Science Conference Proceedings (OSTI)

As part of the proficiency training for the Radiological Mapping mission of the Aerial Measuring System (AMS), a survey team from the Remote Sensing Laboratory-Nellis (RSL-Nellis) conducted an aerial radiological survey of selected areas of Area 18 of the Nevada Test Site (NTS) for the purpose of mapping man-made radiation deposited as a result of the Johnnie Boy and Little Feller I tests. The survey area centered over the Johnnie Boy ground zero but also included the ground zero and deposition area of the Little Feller I test, approximately 7,000 feet (2133 meters) southeast of the Johnnie Boy site. The survey was conducted in one flight. The completed survey covered a total of 4.0 square miles. The flight lines (with the turns) over the surveyed areas are presented in Figure 1. One 2.5-hour-long flight was performed at an altitude of 100 ft above ground level (AGL) with 200 foot flight-line spacing. A test-line flight was conducted near the Desert Rock Airstrip to ensure quality control of the data. The test line is not shown in Figure 1. However, Figure 1 does include the flight lines for a ''perimeter'' flight. The path traced by the helicopter flying over distinct roads within the survey area can be used to overlay the survey data on a base map or image. The flight survey lines were flown in an east-west orientation perpendicular to the deposition patterns for both sites. This technique provides better spatial resolution when contouring the data. The data were collected by the AMS data acquisition system (REDAR V) using an array of twelve 2-inch x 4-inch x 16-inch sodium iodide (NaI) detectors flown on-board a twin-engine Bell 412 helicopter. Data, in the form of gamma energy spectra, were collected every second over the course of the survey and were geo-referenced using a differential Global Positioning System. Spectral data allows the system to distinguish between ordinary fluctuations in natural background radiation levels and the signature produced by man-made radioisotopes. Spectral data can also identify specific radioactive isotopes. Based on the results of the RSL NTS 1994 surveys, this area was chosen for a resurvey to improve the spatial resolution of the reported depositions for the Johnnie Boy and Little Feller I events. In addition, the survey was expected to confirm the absence of detectable concentrations of Americium-241 (Am-241) at the Johnnie Boy site and attempt to confirm the presence of Uranium-235 (U-235).

Craig Lyons

2009-07-31T23:59:59.000Z

129

An X-Band Gun Test Area at SLAC  

SciTech Connect

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

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

2012-09-07T23:59:59.000Z

130

Testing mass-varying neutrinos with reactor experiments  

E-Print Network (OSTI)

We propose that reactor experiments could be used to constrain the environment dependence of neutrino mass and mixing parameters, which could be induced due to an acceleron coupling to matter fields. There are several short-baseline reactor experiment projects with different fractions of air and earth matter along the neutrino path. Moreover, the short baselines, in principle, allow the physical change of the material between source and detector. Hence, such experiments offer the possibility for a direct comparison of oscillations in air and matter. We demonstrate that for sin 2 (2?13) ? 0.04, two reactor experiments (one air, one matter) with baselines of at least 1.5 km can constrain any oscillation effect which is different in air and matter at the level of a few per cent. Furthermore, we find that using the same experiment while physically moving the material between source and detector improves systematics. PACS: 14.60.Pq

unknown authors

2005-01-01T23:59:59.000Z

131

MATERIALS TESTING REACTOR PROJECT. QUARTERLY REPORT FOR PERIOD ENDING MARCH 1, 1950  

SciTech Connect

Progress is reported in finaiizing basic design data for the Materials Testing Reactor. The major emphasis at ANL was on issurance of design reports on practically all phases of the MTR project outside the reactor face and low the first fioor level. Operation of the mock-up reacr at ORNL at 10 watts resulted in no major design changes. Topics discussed include the reactor building, wing, and reactor service building; canal and canal facilities; water systems; air exhaust systems; electrical power systems; effluent control; and shielding requirements. 11 drawings. (C.H.)

Huffman, J.R.

1958-10-31T23:59:59.000Z

132

An update for the MuCool test area  

DOE Green Energy (OSTI)

Construction of a new facility known as the MuCool Test Area (MTA) has been completed at Fermi National Accelerator Laboratory. This facility supports research in new accelerator technologies for future endeavors such as a Neutrino Factory or Muon Collider. During the summer of 2004, an initial set of tests was completed for the filling of a convection-style liquid hydrogen absorber designed by KEK. The absorber contained 6.2 liquid liters of hydrogen and was tested for a range of heating conditions to quantify the absorber's heat exchanger performance. Future work at Fermilab includes the design, construction, and installation of a forced-flow absorber to be used with other components built to investigate the properties of a muon ionization cooling channel. A Tevatron-style refrigerator/compressor building is to be operational by spring of 2006 in support of the absorber tests and also to provide 5-K helium and liquid nitrogen to a 5-T solenoid magnet, an active element of the future test apparatus. The refrigerator will be configured in such a manner as to meet the 5 K and 14-20-K helium needs of the MTA. This paper reviews the challenges and successes of the past KEK absorber tests as well as looks into the future cryogenic capabilities and intentions of the site.

Bross, A.; Cummings, M.A.; Darve, C.; Ishimoto, S.; Klebaner, A.; Martinez, A.; Norris, B.; Pei, L.; /Fermilab /KEK, Tsukuba /Northern Illinois U.

2006-01-01T23:59:59.000Z

133

100 Area excavation treatability test plan. Revision 1  

SciTech Connect

This test plan documents the requirements for a treatability study on field radionuclide analysis and dust control techniques. These systems will be used during remedial actions involving excavation. The data from this treatability study will be used to support the feasibility study (FS) process. Excavation is one of the high-priority, near-term, treatability study needs identified in the program plan (DOE-RL 1992f). Excavation of contaminated soils and buried solid wastes is included in several of the alternatives identified in the 100 Area FS. Although a common activity, excavation has only been used occasionally at the Hanford Site for waste removal applications. The most recent applications are excavation of the 618-9 burial ground and partial remediation of the 316-5 process trenches (DOE-RL 1992a, 1992b). Both projects included excavation of soil and dust control (using water sprays). Excavation is a well-developed technology and equipment is readily available; however, certain aspects of the excavation process require testing before use in full-scale operations. These include the following: Measurement and control of excavation-generated dust and airborne contamination; verification of field analytical system capabilities; demonstration of soil removal techniques specific to the 100 Area waste site types and configurations. The execution of this treatability test may produce up to 500 yd{sub 3} of contaminated soil, which will be used for future treatability tests. These tests may include soil washing with vitrification of the soil washing residuals. Other tests will be conducted if soil washing is not a viable alternative.

Not Available

1993-08-01T23:59:59.000Z

134

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

Science Conference Proceedings (OSTI)

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).

D. L. Knudson; J. L. Rempe

2012-02-01T23:59:59.000Z

135

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

Science Conference Proceedings (OSTI)

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.

D. L. Knudson; J. L. Rempe

2012-02-01T23:59:59.000Z

136

Hanford 100-D Area Biostimulation Treatability Test Results  

SciTech Connect

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.

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

137

Flow Test At Fort Bliss Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Flow Test At Fort Bliss Area (DOE GTP) Exploration Activity Details Location Fort Bliss Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding...

138

Flow Test At Glass Buttes Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Flow Test At Glass Buttes Area (DOE GTP) Exploration Activity Details Location Glass Buttes Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding...

139

Flow Test At The Needles Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Flow Test At The Needles Area (DOE GTP) Exploration Activity Details Location The Needles Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding...

140

Flow Test At Mccoy Geothermal Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Flow Test At Mccoy Geothermal Area (DOE GTP) Exploration Activity Details Location Mccoy Geothermal Area Exploration Technique Flow Test Activity Date Usefulness not indicated...

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Flow Test At Gabbs Valley Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Flow Test At Gabbs Valley Area (DOE GTP) Exploration Activity Details Location Gabbs Valley Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding...

142

Tracer Testing At Jemez Pueblo Area (DOE GTP) | Open Energy Informatio...  

Open Energy Info (EERE)

Tracer Testing At Jemez Pueblo Area (DOE GTP) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Tracer Testing At Jemez Pueblo Area (DOE GTP)...

143

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

Science Conference Proceedings (OSTI)

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

Terry, W.K.; Ryskamp, J.M.; Sterbentz, J.W. [and others

1995-08-01T23:59:59.000Z

144

Evaluation and Test of Improved Fire Resistant Fluid Lubricants for Water Reactor Coolant Pump Motors, Volume 1: Fluid Evaluation, Bearing Model Tests, Motor Tests, and Fire Tests  

Science Conference Proceedings (OSTI)

Commercially available fire-resistant fluid lubricants were evaluated to determine their suitability for use in primary-system pump motors in nuclear reactors. Volume 1 describes the procedures and results of tests of lubrication properties; fire and radiation resistance; and thermal, oxidative, and hydrolytic stability.

1980-07-01T23:59:59.000Z

145

An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components  

SciTech Connect

This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

Holcomb, David Eugene [ORNL; Cetiner, Mustafa Sacit [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

2009-11-01T23:59:59.000Z

146

PEROXIDE DESTRUCTION TESTING FOR THE 200 AREA EFFLUENT TREATMENT FACILITY  

Science Conference Proceedings (OSTI)

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.

HALGREN DL

2010-03-12T23:59:59.000Z

147

Field investigation at the Faultless Site Central Nevada Test Area  

DOE Green Energy (OSTI)

An evaluation of groundwater monitoring at non-Nevada Test Site underground nuclear test sites raised questions about the potential for radionuclide migration from the Faultless event and how to best monitor for such migration. With its long standing interest in the Faultless area and background in Nevada hydrogeology, the Desert Research Institute conducted a field investigation in FY92 to address the following issues: The status of chimney infilling (which determines the potential for migration); the best level(s) from which to collect samples from the nearby monitoring wells, HTH-1 and HTH-2; the status of hydraulic heads in the monitoring well area following records of sustained elevated post-shot heads. The field investigation was conducted from July 27 to 31 and August 4 to 7, 1992. Temperature and electrical conductivity logging were performed in HTH-1, HTH-2, and UC-1-P-2SR. Water samples were collected from HTH-1 and HTH-2. Lawrence Livermore National Laboratory (LLNL) also collected samples during the July trip, including samples from UC-1-P-2SR. This report presents the data gathered during these field excursions and some preliminary conclusions. Full interpretation of the data in light of the issues listed above is planned for FY93.

Chapman, J.B.; Mihevc, T.M.; Lyles, B.

1992-11-01T23:59:59.000Z

148

Production test IP-338-A, Supp. A, DR-Reactor heat decay test at high outlet water temperatures  

SciTech Connect

This test is identical to the original except that it authorizes the performance of a trial reduction in reactor flow during a prior reactor shutdown. This trial flow reduction will be performed in the same manner as proposed for the actual test, with one exception. This is, that based upon the results of this preliminary test some changes in the timing of the different steps may be indicated. Such changes can readily be handled by making each step dependent upon the observed reactor outlet temperature during the test performance. The other significant change in the production test is the increase in the allowable bulk outlet temperature from Ti + 40 {plus_minus} 3{degrees}C{sup *}. This change is needed to obtain a reasonable extrapolation of the results of tests No. 1 and No.2 to 90{degrees}C, and is justified from a hazards standpoint by the excellent flow control achieved during test No. 1 and by the trial test that will be run prior to the performance of the actual test No. 2. Other aspects of the test basis and justification are presented in the original production test.

Jones, S.S.

1962-05-18T23:59:59.000Z

149

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

SciTech Connect

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.

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

2012-05-01T23:59:59.000Z

150

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

Science Conference Proceedings (OSTI)

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.

Frank Darmann; Robert Lombaerde; Franco Moriconi; Albert Nelson

2011-10-31T23:59:59.000Z

151

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

SciTech Connect

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.

S. Blaine Grover

2009-09-01T23:59:59.000Z

152

Fuel subassembly leak test chamber for a nuclear reactor  

DOE Patents (OSTI)

A container with a valve at one end is inserted into a nuclear reactor coolant pool. Once in the pool, the valve is opened by a mechanical linkage. An individual fuel subassembly is lifted into the container by a gripper; the valve is then closed providing an isolated chamber for the subassembly. A vacuum is drawn on the chamber to encourage gaseous fission product leakage through any defects in the cladding of the fuel rods comprising the subassembly; this leakage may be detected by instrumentation, and the need for replacement of the assembly ascertained.

Divona, Charles J. (Santa Ana, CA)

1978-04-04T23:59:59.000Z

153

MODELING ASSUMPTIONS FOR THE ADVANCED TEST REACTOR FRESH FUEL SHIPPING CONTAINER  

SciTech Connect

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

Rick J. Migliore

2009-09-01T23:59:59.000Z

154

Solar test of an integrated sodium reflux heat pipe receiver/reactor for thermochemical energy transport  

DOE Green Energy (OSTI)

A chemical reactor for carbon dioxide reforming of methane was integrated into a sodium reflux heat pipe receiver and tested in the solar furnace of the Weizmann Institute of Science, Rehovot, Israel. The receiver/reactor was a heat pipe with seven tubes inside an evacuated metal box containing sodium. The catalyst, 0.5 wt% Rh on alumina, filled two of the tubes with the front surface of the box serving as the solar absorber. In operation, concentrated sunlight heated the front plate and vaporized sodium from a wire mesh wick attached to other side. Sodium vapor condensed on the reactor tubes, releasing latent heat and returning to the wick by gravity. The receiver system performed satisfactorily in many tests under varying flow conditions. The maximum power absorbed was 7.5 kW at temperatures above 800C. The feasibility of operating a heat pipe receiver/reactor under solar conditions was proven, and the advantages of reflux devices confirmed.

Diver, R.B.; Fish, J.D. (Sandia National Labs., Albuquerque, NM (United States)); Levitan, R.; Levy, M.; Meirovitch, E.; Rosin, H. (Weizmann Inst. of Science, Rehovot (Israel)); Paripatyadar, S.A.; Richardson, J.T. (Univ. of Houston, TX (United States))

1992-01-01T23:59:59.000Z

155

Accident safety analysis for 300 Area N Reactor Fuel Fabrication and Storage Facility  

SciTech Connect

The purpose of the accident safety analysis is to identify and analyze a range of credible events, their cause and consequences, and to provide technical justification for the conclusion that uranium billets, fuel assemblies, uranium scrap, and chips and fines drums can be safely stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility, the contaminated equipment, High-Efficiency Air Particulate filters, ductwork, stacks, sewers and sumps can be cleaned (decontaminated) and/or removed, the new concretion process in the 304 Building will be able to operate, without undue risk to the public, employees, or the environment, and limited fuel handling and packaging associated with removal of stored uranium is acceptable.

Johnson, D.J.; Brehm, J.R.

1994-01-01T23:59:59.000Z

156

Cryogenics for the MuCool Test Area (MTA)  

DOE Green Energy (OSTI)

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.

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

2005-09-01T23:59:59.000Z

157

Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test  

Science Conference Proceedings (OSTI)

This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy.

Cowell, B.S.

1997-06-01T23:59:59.000Z

158

Advanced Test Reactor LEU Fuel Conversion Feasibility Study -- 2006 Annual Report  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the U.S. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of these operating parameters, and the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U-235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U 235 enrichment required in the fuel meat to yield an equivalent Keff between the HEU core and a LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U 235 loading in the LEU core, such that the differences in Keff and heat profile between the HEU and LEU core can be minimized for operation at 125 EFPD with a total core power of 115 MW. The Monte-Carlo coupled with ORIGEN2 (MCWO) depletion methodology was used to calculate Keff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the Keff versus EFPDs plot is similar in shape to the reference ATR HEU case. The LEU core conversion feasibility study can also be used to optimize the U-235 content of each fuel plate, so that the relative radial fission heat flux profile is bounded by the reference ATR HEU case. The detailed radial, axial, and azimuthal heat flux profiles of the HEU and optimized LEU cases have been investigated. However, to demonstrate that the LEU core fuel cycle performance can meet the UFSAR safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (OSCC, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.

G. S. Chang; R. G. Ambrosek

2006-10-01T23:59:59.000Z

159

Advanced Test Reactor LEU Fuel Conversion Feasibility Study (2006 Annual Report)  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power density and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth with a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of these operating parameters, and the large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. The present work investigates the necessary modifications and evaluates the subsequent operating effects of this conversion. A detailed plate-by-plate MCNP ATR 1/8th core model was developed and validated for a fuel cycle burnup comparison analysis. Using the current HEU U 235 enrichment of 93.0 % as a baseline, an analysis can be performed to determine the low-enriched uranium (LEU) density and U-235 enrichment required in the fuel meat to yield an equivalent K-eff between the HEU core and the LEU core versus effective full power days (EFPD). The MCNP ATR 1/8th core model will be used to optimize the U-235 loading in the LEU core, such that the differences in K-eff and heat profile between the HEU and LEU core can be minimized for operation at 125 EFPD with a total core power of 115 MW. The depletion methodology, Monte-Carlo coupled with ORIGEN2 (MCWO), was used to calculate K-eff versus EFPDs. The MCWO-calculated results for the LEU case demonstrated adequate excess reactivity such that the K-eff versus EFPDs plot is similar in shape to the reference ATR HEU case. The LEU core conversion feasibility study can also be used to optimize the U-235 content of each fuel plate, so that the relative radial fission heat flux profile is bounded by the reference ATR HEU case. The detailed radial, axial, and azimuthal heat flux profiles of the HEU and optimized LEU cases have been investigated. However, to demonstrate that the LEU core fuel cycle performance can meet the UFSAR safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders (OSCCs), safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.

Gray S. Chang; Richard G. Ambrosek; Misti A. Lillo

2006-12-01T23:59:59.000Z

160

Isotope correlation studies relative to high enrichment test reactor fuels  

SciTech Connect

Several correlations of fission product isotopic ratios with atom percent fission and neutron flux, for highly enriched /sup 235/U fuel irradiated in two different water moderated thermal reactors, have been evaluated. In general, excellent correlations were indicated for samples irradiated in the same neutron spectrum; however, significant differences in the correlations were noted with the change in neutron spectrum. For highly enriched /sup 235/U fuel, the correlation of the isotopic ratio /sup 143/Nd//sup 145 +146/Nd with atom percent fission has wider applicability than the other fission product isotopic ratio evaluated. The /sup 137/Cs//sup 135/Cs atom ratio shows promise for correlation with neutron flux. Correlations involving heavy element ratios are very sensitive to the neutron spectrum.

Maeck, W.J.; Tromp, R.L.; Duce, F.A.; Emel, W.A.

1978-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


161

The Palo Verde Reactor Neutrino Experiment A Test for Long Baseline Neutrino Oscillations  

E-Print Network (OSTI)

:1. Our range of sensitivity is tuned to test the š¯ $ še solution of the atmospheric neutrino anomaly. 11 The Palo Verde Reactor Neutrino Experiment A Test for Long Baseline Neutrino Oscillations 94305 e Palo Verde Nuclear Generating Station,Tonopah AZ 85354 Our collaboration has installed a long

Piepke, Andreas G.

162

The Palo Verde Reactor Neutrino Experiment A Test for Long Baseline Neutrino Oscillations  

E-Print Network (OSTI)

\\Gamma3 eV 2 and sin 2 2\\Theta ! 0:1. Our range of sensitivity is tuned to test the š ¯ $ š e solutionThe Palo Verde Reactor Neutrino Experiment A Test for Long Baseline Neutrino Oscillations Presented 85287 S. Pittalwala, R. Wilferd, S. Young Palo Verde Nuclear Generating Station, Tonopah AZ 85354 Our

Piepke, Andreas G.

163

Closure Report for Corrective Action Unit 240: Area 25 Vehicle Washdown Nevada Test Site, Nevada  

SciTech Connect

The Area 25 Vehicle Washdown, Corrective Action Unit (CAU) 240, was clean-closed following the approved Corrective Action Decision Document closure alternative and in accordance with the Federal Facility Agreement and Consent Order (FFACO, 1996). The CAU consists of thee Corrective Action Sites (CASs): 25-07-01 - Vehicle Washdown Area (Propellant Pad); 25-07-02 - Vehicle Washdown Area (F and J Roads Pad); and 25-07-03 - Vehicle Washdown Station (RADSAFE Pad). Characterization activities indicated that only CAS 25-07-02 (F and J Roads Pad) contained constituents of concern (COCs) above action levels and required remediation. The COCs detected were Total Petroleum Hydrocarbons (TPH) as diesel, cesium-137, and strontium-90. The F and J Roads Pad may have been used for the decontamination of vehicles and possibly disassembled engine and reactor parts from Test Cell C. Activities occurred there during the 1960s through early 1970s. The F and J Roads Pad consisted of a 9- by 5-meter (m) (30- by 15-foot [ft]) concrete pad and a 14- by 13-m (46-by 43-ft) gravel sump. The clean-closure corrective action consisted of excavation, disposal, verification sampling, backfilling, and regrading. Closure activities began on August 21, 2000, and ended on September 19, 2000. Waste disposal activities were completed on December 12, 2000. A total of 172 cubic meters (223 cubic yards) of impacted soil was excavated and disposed. The concrete pad was also removed and disposed. Verification samples were collected from the bottom and sidewalls of the excavation and analyzed for TPH diesel and 20-minute gamma spectroscopy. The sample results indicated that all impacted soil above remediation standards was removed. The closure was completed following the approved Corrective Action Plan. All impacted waste was disposed in the Area 6 Hydrocarbon Landfill. All non-impacted debris was disposed in the Area 9 Construction Landfill and the Area 23 Sanitary Landfill.

D. L. Gustafason

2001-03-01T23:59:59.000Z

164

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

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.

M. Chen; CM Regan; D. Noe

2006-01-09T23:59:59.000Z

165

The ORNL High Flux Isotope Reactor and New Advanced Fuel Testing Capabilities  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy s High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), was originally designed (in the 1960s) primarily as a part of the overall program to produce transuranic isotopes for use in the heavy-element research program of the United States. Today, the reactor is a highly versatile machine, producing medical and transuranic isotopes and performing materials test experimental irradiations and neutron-scattering experiments. The ability to test advanced fuels and cladding materials in a thermal neutron spectrum in the United States is limited, and a fast-spectrum irradiation facility does not currently exist in this country. The HFIR has a distinct advantage for consideration as a fuel/cladding irradiation facility because of the extremely high neutron fluxes that this reactor provides over the full thermal- to fast-neutron energy range. New test capabilities have been developed that will allow testing of advanced nuclear fuels and cladding materials in the HFIR under prototypic light-water reactor (LWR) and fast-reactor (FR) operating conditions.

Ott, Larry J [ORNL; McDuffee, Joel Lee [ORNL

2011-01-01T23:59:59.000Z

166

In-situ Creep Testing Capability Development for Advanced Test Reactor  

SciTech Connect

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.

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

2010-08-01T23:59:59.000Z

167

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

DOE Green Energy (OSTI)

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.

G. N. Doyle

2002-02-01T23:59:59.000Z

168

Flow Test At Fish Lake Valley Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Fish Lake Valley Area (DOE GTP) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Fish Lake Valley Area (DOE GTP) Exploration Activity...

169

Flow Test At Rye Patch Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Rye Patch Area (DOE GTP) Exploration Activity Details Location Rye Patch Area...

170

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

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Jemez Pueblo Area (DOE GTP) Exploration Activity Details Location Jemez Pueblo Area...

171

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

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Silver Peak Area (DOE GTP) Exploration Activity Details Location Silver Peak Area...

172

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

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At New River Area (DOE GTP) Exploration Activity Details Location New River Area...

173

Flow Test At Hot Pot Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Pot Area (DOE GTP) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Hot Pot Area (DOE GTP) Exploration Activity Details Location Hot...

174

Warm Water Oxidation Verification - Scoping and Stirred Reactor Tests  

Science Conference Proceedings (OSTI)

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).

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

2011-06-15T23:59:59.000Z

175

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Ott, Larry J [ORNL; Ellis, Ronald James [ORNL; McDuffee, Joel Lee [ORNL; Spellman, Donald J [ORNL; Bevard, Bruce Balkcom [ORNL

2009-01-01T23:59:59.000Z

176

Gas-cooled fast breeder reactor steady-state irradiation testing program  

Science Conference Proceedings (OSTI)

The requirements for the gas-cooled fast breeder reactor irradiation program are specified, and an irradiation program plan which satisfies these requirements is presented. The irradiation program plan consists of three parts and includes a schedule and a preliminary cost estimate: (1) a steady-state irradiation program, (2) irradiations in support of the design basis transient test program, and (3) irradiations in support of the GRIST-2 safety test program. Data from the liquid metal fast breeder reactor program are considered, and available irradiation facilities are examined.

Acharya, R.T.; Campana, R.J.; Langer, S.

1980-08-01T23:59:59.000Z

177

Tests of candidate materials for particle bed reactors  

DOE Green Energy (OSTI)

Rhenium metal hot frits and zirconium carbide-coated fuel particles appear suitable for use in flowing hydrogen to at least 2000 K, based on previous tests. Recent tests on alternate candidate cooled particle and frit materials are described. Silicon carbide-coated particles began to react with rhenium frit material at 1600 K, forming a molten silicide at 2000 K. Silicon carbide was extensively attacked by hydrogen at 2066 K for 30 minutes, losing 3.25% of its weight. Vitrous carbon was also rapidly attacked by hydrogen at 2123 K, losing 10% of its weight in two minutes. Long term material tests on candidate materials for closed cycle helium cooled particle bed fuel elements are also described. Surface imperfections were found on the surface of pyrocarbon-coated fuel particles after ninety days exposure to flowing (approx.500 ppM) impure helium at 1143 K. The imperfections were superficial and did not affect particle strength.

Horn, F.L.; Powell, J.R.; Wales, D.

1987-01-01T23:59:59.000Z

178

Solar test of an integrated sodium reflux heat-pipe receiver/reactor for thermochemical energy transport  

DOE Green Energy (OSTI)

In October 1987, a chemical reactor integrated into a sodium reflux heat-pipe receiver was tested in the solar furnace at the Weizmann Institute of Science, Rehovot, Israel. The reaction carried out was the carbon dioxide reforming of methane. This reaction is one of the leading candidates for thermochemical energy transport either within a distributed solar receiver system or over long distances. The Schaeffer Solar Furnace consists of a 96 square meter heliostat and a 7.3 meter diameter dish concentrator with a 65-degree rim angle and a 3.5 meter focal length. Measurements have shown a peak concentration ratio of over 10,000 and a total power of 15 kW at an insolation of 800 w/square meter. The receiver/reactor contains seven catalyst-filled tubes inside an evacuated metal box containing sodium. The front surface of this box serves as the solar absorber of the receiver. In operation, concentrated sunlight heats the 1/8-inch Inconel plate and vaporizes sodium from the wire-mesh wick attached to the back of it. The sodium vapor condenses on the reactor tubes, releases its latent heat, and returns by gravity to the wick. Test results and areas for future development are discussed.

Diver, R.B.; Fish, J.D.; Levitan, R.; Levy, M.; Rosin, H.; Richardson, J.T.

1988-01-01T23:59:59.000Z

179

Results and Analyses of Irradiation/Anneal Experiments Conducted on Yankee Rowe Reactor Pressure Vessel Surrogate Materials: Yankee Atomic Electric Company Test Reactor Program  

Science Conference Proceedings (OSTI)

Many variables influence the response of reactor vessel steels to neutron irradiation. This study looks at the influence of irradiation temperature, steel heat treatment and microstructure, and nickel and phosphorus content on the irradiation response of high-copper reactor vessel steel. Also addressed are several studies evaluating the potential of thermal annealing to restore the mechanical properties of the steels tested.

1996-03-22T23:59:59.000Z

180

D0 Experimental Area Emergency Backup Power and Generator Test  

SciTech Connect

The DO experimental area has a generator designated as emergency power. This generator provides power for critical loads and starts automatically upon loss of commercial power. This note concerns the testing of this generator. A list of loads is attached to this note. One of the loads on the emergency power grid is a 10KVA Uninterruptable Power Supply(UPS). The UPS powers the cryogenic controls and Oxygen deficiency hazard equipment(ODH) and has a minimum rating of 20 minutes while on its batteries(to cover the transfer time to/from the emergency generator). Jan 23,1991 at 1640 hrs this system was tested under the supervision of the Terry Ross, Marv Johnson, Dan Markley, Kelly Dixon, and John Urbin. The power feeder to the emergency power grid at DO was disconnected. The generator responded immediately and was supplying power to the emergency power grid in less than 10 seconds. During the 10 seconds that there was no power on the emergency grid the UPS switched on its inverter and provided uninterrupted power to the cryogenic control system and the ODH system. All of the motorized equipment shut off instrument air compressor, vacuum pumps 1 and 2, insulating vacuum blower, glycol cooling pumps, cooling tower fan, and Exhaust Fan 7(EF7). Upon reengagement of power to the grid from the emergency generator, all of the motorized loads started back up with the exception of vacuum pumps 1 and 2, and the UPS inverter turned off. Vacuum pumps 1 and 2 were delay started 20 seconds by the cryogenic control system as not to cause too large of a surge in power by all of the inductive loads starting at once. The DO building elevator which is also on emergency power was test run while the emergency generator was on line with all other emergency loads. The emergency generator current was 140 amps with all loads on line and running except the building elevator. This load of 140 amps is 27% of the generator's capacity. The cryogenic control and ODH system continued to function properly throughout the entire test due to the UPS responding correctly to each power situation. The cryogenic control system isolated both the Utility(UV) and insulating(IV) vacuum systems as to preserve their vacua while the pumps were off. Once the vacuum pumps were reestablished the IV and UV vacua were put back on line to their respective pumps by the cryogenic control system. The instrument air is backed up by a high pressure trailer, regulated down to instrument air pressure and switches automatically on line through a check valve. During the time that the instrument air compressor was off, instrument air never went below 80 psig (high pressure regulator setting).

Markley, D.; /Fermilab

1991-01-24T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

D0 Experimental Area Emergency Backup Power and Generator Test  

SciTech Connect

The DO experimental area has a generator designated as emergency power. This generator provides power for critical loads and starts automatically upon loss of commercial power. This note concerns the testing of this generator. A list of loads is attached to this note. One of the loads on the emergency power grid is a 10KVA Uninterruptable Power Supply(UPS). The UPS powers the cryogenic controls and Oxygen deficiency hazard equipment(ODH) and has a minimum rating of 20 minutes while on its batteries(to cover the transfer time to/from the emergency generator). Jan 23,1991 at 1640 hrs this system was tested under the supervision of the Terry Ross, Marv Johnson, Dan Markley, Kelly Dixon, and John Urbin. The power feeder to the emergency power grid at DO was disconnected. The generator responded immediately and was supplying power to the emergency power grid in less than 10 seconds. During the 10 seconds that there was no power on the emergency grid the UPS switched on its inverter and provided uninterrupted power to the cryogenic control system and the ODH system. All of the motorized equipment shut off instrument air compressor, vacuum pumps 1 and 2, insulating vacuum blower, glycol cooling pumps, cooling tower fan, and Exhaust Fan 7(EF7). Upon reengagement of power to the grid from the emergency generator, all of the motorized loads started back up with the exception of vacuum pumps 1 and 2, and the UPS inverter turned off. Vacuum pumps 1 and 2 were delay started 20 seconds by the cryogenic control system as not to cause too large of a surge in power by all of the inductive loads starting at once. The DO building elevator which is also on emergency power was test run while the emergency generator was on line with all other emergency loads. The emergency generator current was 140 amps with all loads on line and running except the building elevator. This load of 140 amps is 27% of the generator's capacity. The cryogenic control and ODH system continued to function properly throughout the entire test due to the UPS responding correctly to each power situation. The cryogenic control system isolated both the Utility(UV) and insulating(IV) vacuum systems as to preserve their vacua while the pumps were off. Once the vacuum pumps were reestablished the IV and UV vacua were put back on line to their respective pumps by the cryogenic control system. The instrument air is backed up by a high pressure trailer, regulated down to instrument air pressure and switches automatically on line through a check valve. During the time that the instrument air compressor was off, instrument air never went below 80 psig (high pressure regulator setting).

Markley, D.; /Fermilab

1991-01-24T23:59:59.000Z

182

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

SciTech Connect

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.

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

2000-02-01T23:59:59.000Z

183

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

Science Conference Proceedings (OSTI)

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.

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

1997-11-01T23:59:59.000Z

184

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

SciTech Connect

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.

NA

2006-03-01T23:59:59.000Z

185

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

SciTech Connect

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.

Mike Lewis

2013-02-01T23:59:59.000Z

186

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

SciTech Connect

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.

Mike Lewis

2012-02-01T23:59:59.000Z

187

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

SciTech Connect

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.

mike lewis

2011-02-01T23:59:59.000Z

188

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

SciTech Connect

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.

Lowry, N.J.

1998-10-21T23:59:59.000Z

189

DESIGN CRITERIA FOR HIGH TEMPERATURE LATTICE TEST REACTOR PROJECT CAH-100  

SciTech Connect

Design and construction specifications to be followed in the development of the reactor, its associated systems and experimental facilities, and the housing and required services for the facility are presented. The testing procedures to be used are outlined. (D.C.W.)

Ballard, D.L.; Brown, W.W.; Harrison, C.W.; Heineman, R.E.; Henry, H.L.; Jeffs, T.W.; Morrow, G.W.; Russell, J.T.; Waite, J.K.

1963-05-24T23:59:59.000Z

190

STREAMLINED APPROACH FOR ENVIRONMENTAL RESTORATION PLAN FOR CORRECTIVE ACTION UNIT 116: AREA 25 TEST CELL C FACILITYNEVADA TEST SITE, NEVADA  

SciTech Connect

This Streamlined Approach for Environmental Restoration Plan identifies the activities required for the closure of Corrective Action Unit 116, Area 25 Test Cell C Facility. The Test Cell C Facility is located in Area 25 of the Nevada Test Site approximately 25 miles northwest of Mercury, Nevada.

NONE

2006-07-01T23:59:59.000Z

191

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

Science Conference Proceedings (OSTI)

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

Erika N. Bailey

2011-10-10T23:59:59.000Z

192

REACTOR  

DOE Patents (OSTI)

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Roman, W.G.

1961-06-27T23:59:59.000Z

193

Mapping of a reactor coolant effluent ground disposal test using an infrared imaging system and ground water potential and temperature measurements  

SciTech Connect

The concept of reactor effluent disposal to ground in infiltration trenches was proposed by Nelson and Alkire in 1963. At that time the available data indicated that radionuclide infiltration rates were probably adequate for trench disposal and that decontamination factors of 10 to 100 should be obtainable. Field tests at 100-F Area 1965 and 100-D Area 1967 have indicated that the infiltration rates are adequate and DF`s of from 2.5 for {sup 51}Cr to 7276 for {sup 65}Zn were obtained during the 100-D test. The purpose of this report is to present the results and interpretations of data from studies conducted over a reactor coolant effluent disposal test site. Data presented in this report were collected over the 100-C Area test in which a significant percentage of the reactor coolant effluent was disposed to an existing trench for a five-month period. Results of infrared thermal surveys and ground water temperature and potential measurements collected during this test are presented.

Eliason, J.R.

1969-04-10T23:59:59.000Z

194

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

Science Conference Proceedings (OSTI)

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

Ishii, M.; Xu, Y.; Revankar, S.T. [Purdue University, West Lafayette, IN 47907 (United States)

2002-07-01T23:59:59.000Z

195

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

SciTech Connect

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.

Reilly, Raymond W.

2012-07-30T23:59:59.000Z

196

Microsoft Word - Sludge Test Area CX Determination Form12172012  

NLE Websites -- All DOE Office Websites (Extended Search)

Sludge Test Facility at the Transuranic (TRU) Waste Processing Center (TWPC) [CX-TWPC-13-0001] Sludge Test Facility at the Transuranic (TRU) Waste Processing Center (TWPC) [CX-TWPC-13-0001] Program or Field Office: Environmental Management - Oak Ridge Location(s) (City/County/State): Oak Ridge, Tennessee Proposed Action Description: The proposed action is to construct and operate a sludge test facility at the Transuranic (TRU) Waste Processing Center (TWPC) to conduct testing activities for sludge mobilization, mixing, and removal from the Melton Valley Storage Tanks (MVST). The testing is needed to develop appropriate, compliant treatment to a final waste form that will meet the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC). This testing is needed for the mobilization, removal, and treatment of the sludge regardless of

197

Preliminary Benchmark Evaluation of Japan’s High Temperature Engineering Test Reactor  

SciTech Connect

A benchmark model of the initial fully-loaded start-up core critical of Japan’s High Temperature Engineering Test Reactor (HTTR) was developed to provide data in support of ongoing validation efforts of the Very High Temperature Reactor Program using publicly available resources. The HTTR is a 30 MWt test reactor utilizing graphite moderation, helium coolant, and prismatic TRISO fuel. The benchmark was modeled using MCNP5 with various neutron cross-section libraries. An uncertainty evaluation was performed by perturbing the benchmark model and comparing the resultant eigenvalues. The calculated eigenvalues are approximately 2-3% greater than expected with an uncertainty of ±0.70%. The primary sources of uncertainty are the impurities in the core and reflector graphite. The release of additional HTTR data could effectively reduce the benchmark model uncertainties and bias. Sensitivity of the results to the graphite impurity content might imply that further evaluation of the graphite content could significantly improve calculated results. Proper characterization of graphite for future Next Generation Nuclear Power reactor designs will improve computational modeling capabilities. Current benchmarking activities include evaluation of the annular HTTR cores and assessment of the remaining start-up core physics experiments, including reactivity effects, reactivity coefficient, and reaction-rate distribution measurements. Long term benchmarking goals might include analyses of the hot zero-power critical, rise-to-power tests, and other irradiation, safety, and technical evaluations performed with the HTTR.

John Darrell Bess

2009-05-01T23:59:59.000Z

198

The results of systems tests of the 500 kV busbar controllable shunting reactor in the Tavricheskaya substation  

Science Conference Proceedings (OSTI)

The results of systems tests of the 500 kV busbar magnetization-controllable shunting reactor (CSR), set up in the Tavricheskaya substation, including measurements of the quality of the electric power, the harmonic composition of the network currents of the reactor for different values of the reactive power consumed, the determination of the regulating characteristics of the reactor, the speed of response of the shunting reactor in the current and voltage stabilization modes, and also the operation of the reactor under dynamic conditions for different perturbations, are presented. The results obtained are analyzed.

Gusev, S. I. [JSC 'FSK EES' (Russian Federation); Karpov, V. N.; Kiselev, A. N.; Kochkin, V. I. [Scientific-Research Institute of Electric Power Engineering (VNIIE) - Branch of the JSC 'NTTs Elektroenergetiki', (Russian Federation)

2009-09-15T23:59:59.000Z

199

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

Science Conference Proceedings (OSTI)

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

NSTec Environmental Management

2009-06-23T23:59:59.000Z

200

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

DOE Green Energy (OSTI)

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.

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

1981-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Initial data testing of ENDF/B-VI for thermal reactor benchmark analysis  

SciTech Connect

This paper summarizes some early data testing of ENDF/B-VI by members of the Cross Section Evaluation Working Group (CSEWG) Thermal Reactor Data Testing Subcommittee. Projections of ENDF/B-VI performance in thermal benchmark calculations are beginning to be available; and in some cases the calculations were performed with only a portion of the cross sections taken from version VI, the remainder taken from earlier data files. A factor delaying the thermal reactor data testing is that the final {sup 235}U evaluation has not yet been officially released--only an earlier evaluation with a constant low-energy eta value (like in version V) is currently available. The official version VI {sup 235}U evaluation (scheduled for release as Mod-1) gives a drooping eta variation at low energy; i.e., eta decreases with decreasing energy. This behavior was suggested by European studies to improve the calculation of temperature coefficients in LWRs.

Williams, M.L. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Kahler, A.C. [Bettis Atomic Power Lab., West Mifflin, PA (United States); MacFarlane, R.E. [Los Alamos National Lab., NM (United States); Milgram, M. [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Wright, R.Q. [Oak Ridge National Lab., TN (United States)

1991-12-31T23:59:59.000Z

202

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

DOE Green Energy (OSTI)

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.

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

1990-10-01T23:59:59.000Z

203

SRF Test Areas Cryogenic System Controls Graphical User Interface  

SciTech Connect

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.

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

2011-06-09T23:59:59.000Z

204

Recent Progress of RF Cavity Study at Mucool Test Area  

DOE Green Energy (OSTI)

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.

Yonehara, Katsuya; /Fermilab

2011-12-02T23:59:59.000Z

205

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

Open Energy Info (EERE)

Flow Test At Raft River Geothermal Area (2008) Flow Test At Raft River Geothermal Area (2008) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Raft River Geothermal Area (2008) Exploration Activity Details Location Raft River Geothermal Area Exploration Technique Flow Test Activity Date 2008 Usefulness not indicated DOE-funding Unknown Exploration Basis To confirm resource using flow tests Notes Both production and injection wells were flow tested. Aslo includes interference testing across the well field. References Glaspey, Douglas J. (30 January 2008) Final Technical Resource Confirmation Testing at the Raft River Geothermal Project, Cassia County, Idaho Retrieved from "http://en.openei.org/w/index.php?title=Flow_Test_At_Raft_River_Geothermal_Area_(2008)&oldid=473856

206

An Experimental Shield Test Facility for the Development of Minimum Weight Shields for Compact Reactor Power Systems  

SciTech Connect

Discussions are given of the characteristics of fission-source plate, graphite reactor, and pool-type reactor facilities applicable to development studies of minimum weight shielding materials. Advantages of a proposed SNAP dual-purpose shielding facility are described in terms of a disk-shaped fission-source plate, reactor, and building. A program for the study of advanced shielding materials is discussed for materials and configuations to be evaluted with the fission-source plate, the testing of the prototype at high-power levels, and full-power tests on the actual reactor.

Tomlinson, R.L.

1959-08-07T23:59:59.000Z

207

300 AREA URANIUM CONTAMINATION  

SciTech Connect

{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

BORGHESE JV

2009-07-02T23:59:59.000Z

208

Criticality Safety Evaluation for the Advanced Test Reactor U-Mo Demonstration Elements  

SciTech Connect

The Reduced Enrichment Research Test Reactors (RERTR) fuel development program is developing a high uranium density fuel based on a (LEU) uranium-molybdenum alloy. Testing of prototypic RERTR fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. Two RERTR-Full Size Demonstration fuel elements based on the ATR-Reduced YA elements (all but one plate fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). The two fuel elements will be irradiated in alternating cycles such that only one element is loaded in the reactor at a time. Existing criticality analyses have analyzed Standard (HEU) ATR elements (all plates fueled) from which controls have been derived. This criticality safety evaluation (CSE) documents analysis that determines the reactivity of the Demonstration fuel elements relative to HEU ATR elements and shows that the Demonstration elements are bound by the Standard HEU ATR elements and existing HEU ATR element controls are applicable to the Demonstration elements.

Leland M. Montierth

2010-12-01T23:59:59.000Z

209

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

SciTech Connect

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

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

2012-02-01T23:59:59.000Z

210

Initial confinement studies of ohmically heated plasmas in the tokamak fusion test reactor  

DOE Green Energy (OSTI)

Initial operation of the tokamak fusion test reactor has concentrated upon confinement studies of ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (tau/sub E/) are 0.1--0.2 s for a line-average density range (n-bar/sub e/) of (1--2.5) x 10/sup 19/ m/sup -3/ with electron temperatures of T/sub e/(o)approx.1.2--2.2 keV, ion temperatures of T/sub i/(0)approx.0.9--1.5 keV, and Z/sub eff/approx.3. A comparison of Princeton large torus, poloidal divertor experiment, and tokamak fusion test reactor plasma confinement supports a dimension-cubed scaling law.

Efthimion, P.C.; Bell, M.; Blanchard, W.R.; Bretz, N.; Cecchi, J.L.; Coonrod, J.; Davis, S.; Dylla, H.F.; Fonck, R.; Furth, H.P.

1984-04-23T23:59:59.000Z

211

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

SciTech Connect

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.

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

1982-03-01T23:59:59.000Z

212

Results of Electric Survey in the Area of Hawaii Geothermal Test Well HGP-A  

Open Energy Info (EERE)

Electric Survey in the Area of Hawaii Geothermal Test Well HGP-A 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 in the Area of Hawaii Geothermal Test Well HGP-A Abstract N/A Authors James Kauahikaua and Douglas Klein Published Journal Geothermal Resources Council, TRANSACTIONS, 1978 DOI Not Provided Check for DOI availability: http://crossref.org Online Internet link for Results of Electric Survey in the Area of Hawaii Geothermal Test Well HGP-A Citation James Kauahikaua,Douglas Klein. 1978. Results of Electric Survey in the Area of Hawaii Geothermal Test Well HGP-A. Geothermal Resources Council, TRANSACTIONS. 2:363-366. Retrieved from "http://en.openei.org/w/index.php?title=Results_of_Electric_Survey_in_the_Area_of_Hawaii_Geothermal_Test_Well_HGP-A&oldid=682499

213

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

SciTech Connect

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.

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

2012-08-01T23:59:59.000Z

214

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

Science Conference Proceedings (OSTI)

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.

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

215

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

DOE Green Energy (OSTI)

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.

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

2006-01-25T23:59:59.000Z

216

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

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.

Hans Gougar

2010-02-01T23:59:59.000Z

217

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

Open Energy Info (EERE)

Flow Test At Lake City Hot Springs Area (Benoit Et Al., 2005) Flow Test At Lake City Hot Springs Area (Benoit Et Al., 2005) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Lake City Hot Springs Area (Benoit Et Al., 2005) Exploration Activity Details Location Lake City Hot Springs Area Exploration Technique Flow Test Activity Date Usefulness useful DOE-funding Unknown Notes Core holes enabled injection and flow testing up to 70 gpm. References Dick Benoit, Joe Moore, Colin Goranson, David Blackwell (2005) Core Hole Drilling And Testing At The Lake City, California Geothermal Field Retrieved from "http://en.openei.org/w/index.php?title=Flow_Test_At_Lake_City_Hot_Springs_Area_(Benoit_Et_Al.,_2005)&oldid=386872" Category: Exploration Activities What links here Related changes

218

200 Area Treated Effluent Disposal Facility operational test specification. Revision 2  

Science Conference Proceedings (OSTI)

This document identifies the test specification and test requirements for the 200 Area Treated Effluent Disposal Facility (200 Area TEDF) operational testing activities. These operational testing activities, when completed, demonstrate the functional, operational and design requirements of the 200 Area TEDF have been met. The technical requirements for operational testing of the 200 Area TEDF are defined by the test requirements presented in Appendix A. These test requirements demonstrate the following: pump station No.1 and associated support equipment operate both automatically and manually; pump station No. 2 and associated support equipment operate both automatically and manually; water is transported through the collection and transfer lines to the disposal ponds with no detectable leakage; the disposal ponds accept flow from the transfer lines with all support equipment operating as designed; and the control systems operate and status the 200 Area TEDF including monitoring of appropriate generator discharge parameters.

Crane, A.F.

1995-02-09T23:59:59.000Z

219

Flow Test At Black Warrior Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Linked Data Page Edit History Share this page on Facebook icon Twitter icon Flow Test At Black Warrior Area (DOE GTP) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal...

220

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

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Flint Geothermal Area (DOE GTP) Exploration Activity Details Location Flint...

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

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

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Mcgee Mountain Area (DOE GTP) Exploration Activity Details Location Mcgee Mountain...

222

Flow Test At Crump's Hot Springs Area (DOE GTP) | Open Energy...  

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Crump's Hot Springs Area (DOE GTP) Exploration Activity Details Location Crump's Hot...

223

Flow Test At San Emidio Desert Area (DOE GTP) | Open Energy Informatio...  

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At San Emidio Desert Area (DOE GTP) Exploration Activity Details Location San Emidio...

224

Flow Test At Steamboat Springs Area (Combs, Et Al., 1999) | Open...  

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Steamboat Springs Area (Combs, Et Al., 1999) Exploration Activity Details Location...

225

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

Open Energy Info (EERE)

Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Newberry Caldera Area (DOE GTP) Exploration Activity Details Location Newberry...

226

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

Open Energy Info (EERE)

DOE GTP) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Soda Lake Area (DOE GTP) Exploration Activity Details Location Soda Lake...

227

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

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.

David W. Nigg; Sean R. Morrell

2012-09-01T23:59:59.000Z

228

Injectivity Test At Fenton Hill Hdr Geothermal Area (Dash, Et Al., 1983) |  

Open Energy Info (EERE)

Injectivity Test At Fenton Hill Hdr Geothermal Area (Dash, Et Al., 1983) Injectivity Test At Fenton Hill Hdr Geothermal Area (Dash, Et Al., 1983) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Injectivity Test At Fenton Hill Hdr Geothermal Area (Dash, Et Al., 1983) Exploration Activity Details Location Fenton Hill Hdr Geothermal Area Exploration Technique Injectivity Test Activity Date Usefulness useful DOE-funding Unknown Notes Fenton Hill HDR site. References Z. V. Dash, H. D. Murphy, R. L. Aamodt, R. G. Aguilar, D. W. Brown, D. A. Counce, H. N. Fisher, C. O. Grigsby, H. Keppler, A. W. Laughlin, R. M. Potter, J. W. Tester, P. E. Trujillo Jr, G. Zyvoloski (1983) Hot Dry Rock Geothermal Reservoir Testing- 1978 To 1980 Retrieved from "http://en.openei.org/w/index.php?title=Injectivity_Test_At_Fenton_Hill_Hdr_Geothermal_Area_(Dash,_Et_Al.,_1983)&oldid=511316"

229

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

Science Conference Proceedings (OSTI)

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.

Hakan Ozaltun & Herman Shen

2011-11-01T23:59:59.000Z

230

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

Science Conference Proceedings (OSTI)

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.

Donna P. Guillen

2012-07-01T23:59:59.000Z

231

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

DOE Green Energy (OSTI)

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

K. B. Campbell

2002-06-01T23:59:59.000Z

232

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

SciTech Connect

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.

Patrick Matthews

2010-03-01T23:59:59.000Z

233

Operational Philosophy for the Advanced Test Reactor National Scientific User Facility  

SciTech Connect

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.

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

2013-02-01T23:59:59.000Z

234

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

SciTech Connect

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.

Blaine Grover

2012-10-01T23:59:59.000Z

235

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

SciTech Connect

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

D. L. Gustafason

2001-02-01T23:59:59.000Z

236

The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology  

Science Conference Proceedings (OSTI)

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.

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

2009-05-01T23:59:59.000Z

237

Injectivity Test At Fenton Hill Hdr Geothermal Area (Grigsby, Et Al., 1983)  

Open Energy Info (EERE)

Injectivity Test At Fenton Hill Hdr Geothermal Area (Grigsby, Et Al., 1983) Injectivity Test At Fenton Hill Hdr Geothermal Area (Grigsby, Et Al., 1983) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Injectivity Test At Fenton Hill Hdr Geothermal Area (Grigsby, Et Al., 1983) Exploration Activity Details Location Fenton Hill Hdr Geothermal Area Exploration Technique Injectivity Test Activity Date Usefulness not indicated DOE-funding Unknown References C. O. Grigsby, J. W. Tester, P. E. Trujillo, D. A. Counce, J. Abbott, C. E. Holley, L. A. Blatz (1983) Rock-Water Interactions In Hot Dry Rock Geothermal Systems- Field Investigations Of In Situ Geochemical Behavior Retrieved from "http://en.openei.org/w/index.php?title=Injectivity_Test_At_Fenton_Hill_Hdr_Geothermal_Area_(Grigsby,_Et_Al.,_1983)&oldid=511318

238

EA-1050: Test Area North Pool Stabilization Project, Idaho Falls, Idaho |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

50: Test Area North Pool Stabilization Project, Idaho Falls, 50: Test Area North Pool Stabilization Project, Idaho Falls, Idaho EA-1050: Test Area North Pool Stabilization Project, Idaho Falls, Idaho SUMMARY This EA evaluates the environmental impacts of the U.S. Department of Energy's Idaho National Engineering Laboratory's proposal to remove 344 canisters of Three Mile Island core debris and commercial fuels from the Test Area North Pool and transfer them to the Idaho Chemical Processing Plant for interim dry storage until an alternate storage location other than INEL, or a permanent federal spent nuclear fuel repository is available. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD May 6, 1996 EA-1050: Finding of No Significant Impact Test Area North Pool Stabilization Project

239

Flow Test At Coso Geothermal Area (1985-1986) | Open Energy Information  

Open Energy Info (EERE)

Flow Test At Coso Geothermal Area (1985-1986) Flow Test At Coso Geothermal Area (1985-1986) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Coso Geothermal Area (1985-1986) Exploration Activity Details Location Coso Geothermal Area Exploration Technique Flow Test Activity Date 1985 - 1986 Usefulness not indicated DOE-funding Unknown Exploration Basis Understand the connectivity of the production and injection wells. Notes A long-term flow test was conducted involving one producing well (well 43-7), one injector (well 88-1), and two observation wells (well 66-6 and California Energy Co's well 71A-7). The flow test included a well production metering system and a water injection metering system. References Sanyal, S.; Menzies, A.; Granados, E.; Sugine, S.; Gentner, R.

240

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

Open Energy Info (EERE)

Flow Test At Lightning Dock Area (Cunniff & Bowers, 2005) Flow Test At Lightning Dock Area (Cunniff & Bowers, 2005) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Lightning Dock Area (Cunniff & Bowers, 2005) Exploration Activity Details Location Lightning Dock Area Exploration Technique Flow Test Activity Date Usefulness useful DOE-funding Unknown Notes After the Welaco temperature survey was completed for TG52-7, preparations were completed for a controlled airlift test. This test was completed in the period from 19-20 September 2003 for some 23 hours. The well produced steady state flow of about 320-325 gpm at a wellhead temperature of 126.7degrees C (260degreesF). This production rate is equivalent to about 162,000 pounds per hour, with the production temperature producing usable

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

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

Open Energy Info (EERE)

Flow Test At Raft River Geothermal Area (2004) Flow Test At Raft River Geothermal Area (2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Raft River Geothermal Area (2004) Exploration Activity Details Location Raft River Geothermal Area Exploration Technique Flow Test Activity Date 2004 Usefulness useful DOE-funding GRED II Notes Geothermal Resource Exploration and Definition Projects Raft River (GRED II): Re-assessment and testing of previously abandoned production wells. The objective of the U.S. Geothermal effort is to re-access the available wellbores, assess their condition, perform extensive testing of the reservoir to determine its productive capacity, and perform a resource utilization assessment. At the time of this paper, all five wells had been

242

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

Open Energy Info (EERE)

Flow Test At Raft River Geothermal Area (2006) Flow Test At Raft River Geothermal Area (2006) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Raft River Geothermal Area (2006) Exploration Activity Details Location Raft River Geothermal Area Exploration Technique Flow Test Activity Date 2006 Usefulness not indicated DOE-funding Unknown Exploration Basis Determine field hydraulic conductivity using borehole impeller flowmeter data Notes A quantitative evaluation of borehole-impeller flowmeter data leads to estimated field hydraulic conductivity. Data were obtained during an injection test of a geothermal well at the Raft River geothermal test site in Idaho. Both stationary and trolling calibrations of the flowmeter were made in the well. Methods were developed to adjust for variations in hole

243

Coupled hydrodynamic-structural analysis of an integral flowing sodium test loop in the TREAT reactor  

SciTech Connect

A hydrodynamic-structural response analysis of the Mark-IICB loop was performed for the TREAT (Transient Reactor Test Facility) test AX-1. Test AX-1 is intended to provide information concerning the potential for a vapor explosion in an advanced-fueled LMFBR. The test will be conducted in TREAT with unirradiated uranium-carbide fuel pins in the Mark-IICB integral flowing sodium loop. Our analysis addressed the ability of the experimental hardware to maintain its containment integrity during the reference accident postulated for the test. Based on a thermal-hydraulics analysis and assumptions for fuel-coolant interaction in the test section, a pressure pulse of 144 MPa maximum pressure and pulse width of 1.32 ms has been calculated as the reference accident. The response of the test loop to the pressure transient was obtained with the ICEPEL and STRAW codes. Modelling of the test section was completed with STRAW and the remainder of the loop was modelled by ICEPEL.

Zeuch, W.R.; A-Moneim, M.T.

1979-01-01T23:59:59.000Z

244

Bench-scale reactor tests of low-temperature, catalytic gasification of wet, industrial wastes  

DOE Green Energy (OSTI)

Bench-scale reactor tests are under way at Pacific Northwest Laboratory to develop a low-temperature, catalytic gasification system. The system, licensed under the trade name Thermochemical Environmental Energy System (TEES{reg sign}), is designed for to a wide variety of feedstocks ranging from dilute organics in water to waste sludges from food processing. The current research program is focused on the use of a continuous-feed, tubular reactor. The catalyst is nickel metal on an inert support. Typical results show that feedstocks such as solutions of 2% para-cresol or 5% and 10% lactose in water or cheese whey can be processed to >99% reduction of chemical oxygen demand (COD) at a rate of up to 2 L/hr. The estimated residence time is less than 5 min at 360{degree}C and 3000 psig, not including 1 to 2 min required in the preheating zone of the reactor. The liquid hourly space velocity has been varied from 1.8 to 2.9 L feedstock/L catalyst/hr depending on the feedstock. The product fuel gas contains 40% to 55% methane, 35% to 50% carbon dioxide, and 5% to 10% hydrogen with as much as 2% ethane, but less than 0.1% ethylene or carbon monoxide, and small amounts of higher hydrocarbons. The byproduct water stream carries residual organics amounting to less than 500 mg/L COD. 9 refs., 1 fig., 4 tabs.

Elliott, D.C.; Neuenschwander, G.G.; Baker, E.G.; Butner, R.S.; Sealock, L.J.

1990-04-01T23:59:59.000Z

245

Waste Heat Recovery from the Advanced Test Reactor Secondary Coolant Loop  

Science Conference Proceedings (OSTI)

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.

Donna Post Guillen

2012-11-01T23:59:59.000Z

246

IMPROVED COMPUTATIONAL NEUTRONICS METHODS AND VALIDATION PROTOCOLS FOR THE ADVANCED TEST REACTOR  

SciTech Connect

The Idaho National Laboratory (INL) is in the process of modernizing 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 purpose. Of particular importance, a set of as-run core depletion HELIOS calculations for all ATR cycles since August 2009 was successfully completed during 2011. This demonstration supported a decision late in the year to proceed with the phased incorporation of the HELIOS methodology into the ATR fuel cycle management process beginning in 2012. 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 were 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 a flexible, easily-repeatable ATR physics code validation protocol that is consistent with applicable ASTM standards.

David W. Nigg; Joseph W. Nielsen; Benjamin M. Chase; Ronnie K. Murray; Kevin A. Steuhm

2012-04-01T23:59:59.000Z

247

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

SciTech Connect

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)

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

248

F/H Area ETF effluent (H-016 outfall), ceriodaphnia survival/reproduction test, test date: March 21, 1991  

SciTech Connect

This toxicity test was conducted to determine if the effluent from the F/H area at Savannah River Plant affects the survival or reproduction of the test organisms during a seven day period. The test involved exposing the test organisms to a series of dilutions of the effluent. At each dilution the survival and reproduction of ten test organisms was recorded. Each effluent dilution was compared to a control set of test organisms. Survival data were analyzed by Fisher`s Exact Test and the Trimmed Spearman-Karber test to determine the effluent concentration necessary to cause statistically significant (p = 0.05) mortality. Reproduction data was analyzed for normality, homogeneity of variance and equality of replicates among dilutions to determine the appropriate statistical test for analysis of statistical differences in reproduction among dilutions. Results are summarized.

Specht, W.L.

1991-08-01T23:59:59.000Z

249

F/H Area ETF effluent (H-016 outfall), ceriodaphnia survival/reproduction test, test date: March 21, 1991  

SciTech Connect

This toxicity test was conducted to determine if the effluent from the F/H area at Savannah River Plant affects the survival or reproduction of the test organisms during a seven day period. The test involved exposing the test organisms to a series of dilutions of the effluent. At each dilution the survival and reproduction of ten test organisms was recorded. Each effluent dilution was compared to a control set of test organisms. Survival data were analyzed by Fisher's Exact Test and the Trimmed Spearman-Karber test to determine the effluent concentration necessary to cause statistically significant (p = 0.05) mortality. Reproduction data was analyzed for normality, homogeneity of variance and equality of replicates among dilutions to determine the appropriate statistical test for analysis of statistical differences in reproduction among dilutions. Results are summarized.

Specht, W.L.

1991-08-01T23:59:59.000Z

250

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

SciTech Connect

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.

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

1982-03-01T23:59:59.000Z

251

Measurements of Nonlinear Energy Transfer in Turbulence in the Tokamak Fusion Test Reactor  

SciTech Connect

The application of a new bispectral analysis technique to density fluctuation measurements in the core of the Tokamak Fusion Test Reactor indicates that the peak in the autopower spectrum usually lies in a region of linear stability. Large changes in the linear and nonlinear characteristics of the turbulence are observed as the plasma toroidal rotation and/or confinement properties are varied, while estimates of the turbulence-driven diffusivity varies only slightly with rotation. These observations are consistent with the operation of a global organizing property that may be related to the observation of Bohm-like scaling of ion thermal transport. {copyright} {ital 1997} {ital The American Physical Society}

Kim, J.S.; Fonck, R.J.; Durst, R.D. [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin, Madison, Wisconsin 53706 (United States)] [Department of Nuclear Engineering and Engineering Physics, University of Wisconsin, Madison, Wisconsin 53706 (United States); Fernandez, E.; Terry, P.W. [Department of Physics, University of Wisconsin, Madison, Wisconsin 53706 (United States)] [Department of Physics, University of Wisconsin, Madison, Wisconsin 53706 (United States); Paul, S.F.; Zarnstorff, M.C. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)] [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

1997-08-01T23:59:59.000Z

252

Flow Test At Chena Area (Benoit, Et Al., 2007) | Open Energy Information  

Open Energy Info (EERE)

Chena Area (Benoit, Et Al., 2007) Chena Area (Benoit, Et Al., 2007) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Chena Area (Benoit, Et Al., 2007) Exploration Activity Details Location Chena Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding Unknown References Dick Benoit, Gwen Holdmann, David Blackwell (2007) Low Cost Exploration, Testing, And Development Of The Chena Geothermal Resource Retrieved from "http://en.openei.org/w/index.php?title=Flow_Test_At_Chena_Area_(Benoit,_Et_Al.,_2007)&oldid=387083" Category: Exploration Activities What links here Related changes Special pages Printable version Permanent link Browse properties About us Disclaimers Energy blogs Linked Data Developer services OpenEI partners with a broad range of international organizations to grow

253

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

254

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

E-Print Network (OSTI)

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

255

A series of low-altitude aerial radiological surveys of selected regions within Areas 3, 5, 8, 9, 11, 18, and 25 at the Nevada Test Site  

SciTech Connect

A series of low-altitude, aerial radiological surveys of selected regions within Areas 3, 5, 8, 9, 11, 18,and 25 of the Nevada Test Site was conducted from December 1996 through June 1999. The surveys were conducted for the US Department of Energy by the Remote Sensing Laboratory, located in Las Vegas, Nevada, and maintained and operated by Bechtel Nevada. The flights were conducted at a nominal altitude of 15 meters above ground level along a set of parallel flight lines spaced 23 meters apart. The purpose of these low-altitude surveys was to measure, map, and define the areas of americium-241 activity. The americium contamination will be used to determine the areas of plutonium contamination. Americium-241 activity was detected within 8 of the 11 regions. The three regions where americium-241 was not detected were in the inactive Nuclear Rocket Development Station complex in Area 25, which encompassed the Test Cell A and Test Cell C reactor test stands and the Reactor Maintenance Assembly and Disassembly facility.

Colton, D.P.

1999-12-01T23:59:59.000Z

256

Corrective Action Investigation Plan for Corrective Action Unit 232: Area 25 Sewage Lagoons, Nevada Test Site, Nevada, Revision 0  

Science Conference Proceedings (OSTI)

The Corrective Action Investigation Plan for Corrective Action Unit 232, Area 25 Sewage Lagoons, 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 232 consists of Corrective Action Site 25-03-01, Sewage Lagoon. Corrective Action Unit 232, Area 25 Sewage Lagoons, received sanitary effluent from four buildings within the Test Cell ''C'' Facility from the mid-1960s through approximately 1996. The Test Cell ''C'' Facility was used to develop nuclear propulsion technology by conducting nuclear test reactor studies. Based on the site history collected to support the Data Quality Objectives process, contaminants of potential concern include volatile organic compounds, semivolatile organic compounds, Resource Conservation and Recovery Act metals, petroleum hydrocarbons, polychlorinated biphenyls, pesticides, herbicides, gamma emitting radionuclides, isotopic plutonium, isotopic uranium, and strontium-90. A detailed conceptual site model is presented in Section 3.0 and Appendix A of this Corrective Action Investigation Plan. The conceptual model serves as the basis for the sampling strategy. Under the Federal Facility Agreement and Consent Order, the Corrective Action Investigation Plan will be submitted to the Nevada Division of Environmental Protection for approval. Field work will be conducted following approval of the plan. The results of the field investigation will support a defensible evaluation of corrective action alternatives in the Corrective Action Decision Document.

USDOE /NV

1999-05-01T23:59:59.000Z

257

Construction of MV-6 Well Pad at the Central Nevada Test Area Completed |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Construction of MV-6 Well Pad at the Central Nevada Test Area Construction of MV-6 Well Pad at the Central Nevada Test Area Completed Construction of MV-6 Well Pad at the Central Nevada Test Area Completed October 22, 2013 - 6:10pm Addthis What does this project do? Goal 1. Protect human health and the environment A new groundwater monitoring/validation (MV) well was installed at the Central Nevada Test Area (CNTA) in September 2013. LM proposed this well to the Nevada Division of Environmental Protection (NDEP) to enhance the existing monitoring network and to expedite the Federal Facility Agreement and Consent Order (FFACO) closure process for the CNTA Subsurface Corrective Action Unit. CNTA is located in Hot Creek Valley in Nye County, Nevada, adjacent to U.S. Highway 6, about 30 miles north of Warm Springs, Nevada. CNTA was the site of "Project Faultless," a test site where a

258

Construction of MV-6 Well Pad at the Central Nevada Test Area Completed |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Construction of MV-6 Well Pad at the Central Nevada Test Area Construction of MV-6 Well Pad at the Central Nevada Test Area Completed Construction of MV-6 Well Pad at the Central Nevada Test Area Completed October 22, 2013 - 6:10pm Addthis What does this project do? Goal 1. Protect human health and the environment A new groundwater monitoring/validation (MV) well was installed at the Central Nevada Test Area (CNTA) in September 2013. LM proposed this well to the Nevada Division of Environmental Protection (NDEP) to enhance the existing monitoring network and to expedite the Federal Facility Agreement and Consent Order (FFACO) closure process for the CNTA Subsurface Corrective Action Unit. CNTA is located in Hot Creek Valley in Nye County, Nevada, adjacent to U.S. Highway 6, about 30 miles north of Warm Springs, Nevada. CNTA was the site of "Project Faultless," a test site where a

259

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

Open Energy Info (EERE)

Flow Test At Raft River Geothermal Area (1979) Flow Test At Raft River Geothermal Area (1979) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Raft River Geothermal Area (1979) Exploration Activity Details Location Raft River Geothermal Area Exploration Technique Flow Test Activity Date 1979 Usefulness useful DOE-funding Unknown Exploration Basis To allow for the lateral and vertical extrapolation of core and test data and bridged the gap between surface geophysical data and core analyses. Notes Temperature and flowmeter logs provide evidence that these fractures and faults are conduits that conduct hot water to the wells. One of the intermediate depth core holes penetrated a hydrothermally altered zone that includes several fractures producing hot water. This altered production

260

F/H Area ETF effluent (H-016 outfall) ceriodaphnia survival/reproduction test, test date: December 28, 1989  

SciTech Connect

This toxicity test was conducted to determine if the effluent from the H/F area of Savannah River Plant affect the survival or reproduction of the test organisms during a seven day period. The test involved exposing the test organisms to a series of dilutions of the effluent. At each dilution the survival and reproduction of ten test organisms was recorded. Each effluent dilution was compared to a control set of test organisms. Survival data were analyzed by Fisher`s Exact Test and Probit Analysis to determine the effluent concentration necessary to cause statistically significant (p=0.05) mortality. Reproduction data was analyzed for normality, homogeneity of variance and equality of replicates among dilutions to determine the appropriate statistical test for analysis of statistical differences in reproduction among dilutions. Results are summarized.

Specht, W.L.

1991-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

F/H Area ETF effluent (H-016 outfall) ceriodaphnia survival/reproduction test, test date: December 28, 1989  

SciTech Connect

This toxicity test was conducted to determine if the effluent from the H/F area of Savannah River Plant affect the survival or reproduction of the test organisms during a seven day period. The test involved exposing the test organisms to a series of dilutions of the effluent. At each dilution the survival and reproduction of ten test organisms was recorded. Each effluent dilution was compared to a control set of test organisms. Survival data were analyzed by Fisher's Exact Test and Probit Analysis to determine the effluent concentration necessary to cause statistically significant (p=0.05) mortality. Reproduction data was analyzed for normality, homogeneity of variance and equality of replicates among dilutions to determine the appropriate statistical test for analysis of statistical differences in reproduction among dilutions. Results are summarized.

Specht, W.L.

1991-08-01T23:59:59.000Z

262

Corrective action investigation plan for CAU Number 453: Area 9 Landfill, Tonopah Test Range  

SciTech Connect

This Corrective Action Investigation Plan (CAIP) contains the environmental sample collection objectives and criteria for conducting site investigation activities at the Area 9 Landfill, Corrective Action Unit (CAU) 453/Corrective Action (CAS) 09-55-001-0952, 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 Area 9 Landfill is located northwest of Area 9 on the TTR. The landfill cells associated with CAU 453 were excavated to receive waste generated from the daily operations conducted at Area 9 and from range cleanup which occurred after test activities.

NONE

1997-05-14T23:59:59.000Z

263

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

SciTech Connect

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.

John Jackson; Todd Allen; Frances Marshall; Jim Cole

2013-03-01T23:59:59.000Z

264

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

Science Conference Proceedings (OSTI)

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

Douglas M. Gerstner

2009-05-01T23:59:59.000Z

265

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

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.

KIRK WINTERHOLLER

2008-02-25T23:59:59.000Z

266

ENVIRONMENTAL IlONITORING REPORT FOR THE NEVADA TEST SITE AND OTHER TEST AREAS USED FOR UNDERGROUND NUCLEAR DETONATIONS  

Office of Legacy Management (LM)

IlONITORING REPORT FOR THE NEVADA TEST SITE IlONITORING REPORT FOR THE NEVADA TEST SITE AND OTHER TEST AREAS USED FOR UNDERGROUND NUCLEAR DETONATIONS January through December 1975 Nonitoring Operations Division Environmental Monitoring and Support Laboratory U.S. ENVIRONMENTAL PROTECTION AGENCY Las Vegas, Nevada 89114 APRIL 1976 This work performed under a Memorandum of Understanding No. AT(26-1)-539 for the U . S . ENERGY RESEARCH & DEVELOPMENT ADMINISTRATION EMSL-LV-5 39-4 May 1976 ENVIRONMENTAL 14ONITORING REPORT FOR THE NEVADA TEST SITE AND OTHER TEST AREAS USED FOR UNDERGROUND NUCLEAR DETONATIONS January through December I975 Monitoring Operations Division Environmental Monitoring and Support Laboratory U.S. ENVIRONMENTAL PROTECTION AGENCY Las Vegas, Nevada 89114 APRIL 1976 This work performed under a Memorandum of

267

Hot-gas filter testing with the transport reactor demonstration unit  

Science Conference Proceedings (OSTI)

The objectives of the hot-gas cleanup (HGC) work on the transport reactor demonstration unit (TRDU) located at the Energy & Environmental Research Center (EERC) is to demonstrate acceptable performance of hot-gas filter elements in a pilot-scale system prior to long-term demonstration tests. The primary focus of the experimental effort in the 2-year project will be the testing of hot-gas filter element performance (particulate collection efficiency, filter pressure differential, filter cleanability, and durability) as a function of temperature and filter face velocity during short-term operation (100-200 hours). This filter vessel will be utilized in combination with the TRDU to evaluate the performance of selected hot-gas filter elements under gasification operating conditions. This work will directly support the power systems development facility (PSDF) utilizing the M.W. Kellogg transport reactor located at Wilsonville, Alabama and, indirectly, the Foster Wheeler advanced pressurized fluid-bed combustor, also located at Wilsonville.

Mann, M.D.; Swanson, M.L.; Ness, R.O.; Haley, J.S.

1995-11-01T23:59:59.000Z

268

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

Science Conference Proceedings (OSTI)

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.

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

2011-03-01T23:59:59.000Z

269

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

SciTech Connect

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.

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

2006-06-01T23:59:59.000Z

270

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

SciTech Connect

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.

K. B. Campbell

2003-03-01T23:59:59.000Z

271

Fuel development activities of the US RERTR Program. [Reduced Enrichment Research and Test Reactor  

SciTech Connect

Progress in the development and irradiation testing of high-density fuels for use with low-enriched uranium in research and test reactors is reported. Swelling and blister-threshold temperature data obtained from the examination of miniature fuel plates containing UAl/sub x/, U/sub 3/O/sub 8/, U/sub 3/Si/sub 2/, or U/sub 3/Si dispersed in an aluminum matrix are presented. Combined with the results of metallurgical examinations, these data show that these four fuel types will perform adequately to full burnup of the /sup 235/U contained in the low-enriched fuel. The exothermic reaction of the uranium-silicide fuels with aluminum has been found to occur at about the same temperature as the melting of the aluminum matrix and cladding and to be essentially quenched by the melting endotherm. A new series of miniature fuel plate irradiations is also discussed.

Snelgrove, J.L.; Domagala, R.F.; Wiencek, T.C.; Copeland, G.L.

1983-01-01T23:59:59.000Z

272

Aerial Photography At Nevada Test And Training Range Area (Sabin, Et Al.,  

Open Energy Info (EERE)

2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Aerial Photography At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Exploration Activity Details Location Nevada Test And Training Range Area Exploration Technique Aerial Photography Activity Date Usefulness not indicated DOE-funding Unknown Notes We re-examined most of the area using newer orthophotography, SPOT, and Thematic Mapper images, and identified several areas of possible late Quaternary surface faulting (Figure 3). References A. E. Sabin, J. D. Walker, J. Unruh, F. C. Monastero (2004) Toward The Development Of Occurrence Models For Geothermal Resources In The Western United States Retrieved from "http://en.openei.org/w/index.php?title=Aerial_Photography_At_Nevada_Test_And_Training_Range_Area_(Sabin,_Et_Al.,_2004)&oldid=386843

273

Flow Test At Pilgrim Hot Springs Area (DOE GTP) | Open Energy Information  

Open Energy Info (EERE)

Area (DOE GTP) Area (DOE GTP) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Pilgrim Hot Springs Area (DOE GTP) Exploration Activity Details Location Pilgrim Hot Springs Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding Unknown References (1 January 2011) GTP ARRA Spreadsheet Retrieved from "http://en.openei.org/w/index.php?title=Flow_Test_At_Pilgrim_Hot_Springs_Area_(DOE_GTP)&oldid=402456" Categories: Exploration Activities DOE Funded Activities ARRA Funded Activities What links here Related changes Special pages Printable version Permanent link Browse properties 429 Throttled (bot load) Error 429 Throttled (bot load) Throttled (bot load) Guru Meditation: XID: 1863028959 Varnish cache server

274

Hot-Gas Filter Testing with a Transport Reactor Development Unit  

Science Conference Proceedings (OSTI)

The objective of the hot-gas cleanup (HGC) work on the transport reactor demonstration unit (TRDU) located at the Environmental Research Center is to demonstrate acceptable performance of hot-gas filter elements in a pilot-scale system prior to long-term demonstration tests. The primary focus of the experimental effort in the 2-year project will be the testing of hot- gas filter elements as a function of particulate collection efficiency, filter pressure differential, filter cleanability, and durability during relatively short-term operation (100-200 hours). A filter vessel will be used in combination with the TRDU to evaluate the performance of selected hot- gas filter elements under gasification operating conditions. This work will directly support the Power Systems Development Facility utilizing the M.W. Kellogg transport reactor located at Wilsonville, Alabama and indirectly the Foster Wheeler advanced pressurized fluid-bed combustor, also located at Wilsonville and the Clean Coal IV Pinon Pine IGCC Power Project. This program has a phased approach involving modification and upgrades to the TRDU and the fabrication, assembly, and operation of a hot-gas filter vessel (HGFV) capable of operating at the outlet design conditions of the TRDU. Phase 1 upgraded the TRDU based upon past operating experiences. Additions included a nitrogen supply system upgrade, upgraded LASH auger and 1807 coal feed lines, the addition of a second pressurized coal feed hopper and a dipleg ash hopper, and modifications to spoil the performance of the primary cyclone. Phase 2 included the HGFV design, procurement, and installation. Phases 3 through 5 consist of 200-hour hot-gas filter tests under gasification conditions using the TRDU at temperatures of 540-650{degrees}C (1000-1200{degrees}F), 9.3 bar, and face velocities of 1.4, 2. and 3.8 cm/s, respectively. The increased face velocities are achieved by removing candles between each test.

Swanson, M.L.; Ness, R.O., Jr. [North Dakota Univ., Grand Forks, ND (United States). Energy and Environmental Research Center

1996-12-31T23:59:59.000Z

275

Geodetic Survey At Nevada Test And Training Range Area (Sabin, Et Al.,  

Open Energy Info (EERE)

Nevada Test And Training Range Area (Sabin, Et Al., 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 Training Range Area (Sabin, Et Al., 2004) Exploration Activity Details Location Nevada Test And Training Range Area Exploration Technique Geodetic Survey Activity Date Usefulness not indicated DOE-funding Unknown Notes NAFR straddles the boundary of the Walker Lane belt and the Basin and Range extensional province. Neotectonic motions are inferred from GPS and seismic observations. GPS velocities indicate that the strain field changes from the east-west extension typical of the Basin and Range to the northwest-southeast-directed transtension characteristic of the Walker Lane belt across the region.

276

Injectivity Test At Newberry Caldera Area (Combs, Et Al., 1999) | Open  

Open Energy Info (EERE)

Newberry Caldera Area (Combs, Et Al., 1999) Newberry Caldera Area (Combs, Et Al., 1999) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Injectivity Test At Newberry Caldera Area (Combs, Et Al., 1999) Exploration Activity Details Location Newberry Caldera Area Exploration Technique Injectivity Test Activity Date Usefulness useful DOE-funding Unknown Notes After circulating the mud out of the hole and replacing it with clear water, we attempted two injection tests; one into the open hole section (51 16'- 5360') below the HQ liner, and one into the annulus outside the uncemented part (2748' - -4800') of the liner. References Jim Combs, John T. Finger, Colin Goranson, Charles E. Hockox Jr., Ronald D. Jacobsen, Gene Polik (1999) Slimhole Handbook- Procedures And Recommendations For Slimhole Drilling And Testing In Geothermal Exploration

277

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

Open Energy Info (EERE)

Lassen Volcanic National Park Area (Janik & Mclaren, 2010) Lassen Volcanic National Park Area (Janik & Mclaren, 2010) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Lassen Volcanic National Park Area (Janik & Mclaren, 2010) Exploration Activity Details Location Lassen Volcanic National Park Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding Unknown Notes Water samples were collected during nitrogen-stimulated flow tests in 1978, but no information was provided on sampling conditions. The well was flowed again for the last time in 1982, but the flow test lasted only 1 h (Thompson, 1985). References Cathy J. Janik, Marcia K. McLaren (2010) Seismicity And Fluid Geochemistry At Lassen Volcanic National Park, California- Evidence For Two

278

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

Open Energy Info (EERE)

1978 DOI Not Provided Check for DOI availability: http:crossref.org Online Internet link for Results of Electric Survey in the Area of Hawaii Geothermal Test Well HGP-A...

279

Aquifer Testing Recommendations for Supporting Phase II of the T Area Technetium-99 Data Objectives Process  

Science Conference Proceedings (OSTI)

Aquifer characterization needs are currently being assessed to optimize pump-and-treat remedial strategies within the 200-ZP-1 operable unit, specifically for the immediate area of the 241-T Tank Farm. This report provides a general discussion of the six identified hydrologic test methods for possible subsequent characterization within the 241-T Tank Farm area and details for implementing the large-scale recovery test after terminating pumping at the 241-Tank Farm extraction well locations.

Spane, Frank A.

2008-04-02T23:59:59.000Z

280

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

SciTech Connect

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

Bechtel Nevada

1998-08-31T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

Production test IP-278-A: Verification of BPA loss bulk temperature surge at the DE-Reactor. Supplement A  

SciTech Connect

This report details planning to run a second outage test at the DR-Reactor using the same instrumentation and procedure as an earlier test but increasing the trip-out level from 800 MW up to a maximum of 1200 MW.

Jones, S.S.

1960-01-07T23:59:59.000Z

282

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

Science Conference Proceedings (OSTI)

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.

C. M. Obi

2000-12-01T23:59:59.000Z

283

Closure report for CAU 339: Area 12 Fleet Operations steam-cleaning discharge area, Nevada Test Site  

SciTech Connect

This Closure Report (CR) provides documentation of the completed corrective action at the Area 12 Fleet Operations site located in the southeast portion of the Area 12 Camp at the Nevada Test Site (NTS). Field work was performed in July 1997 as outlined in the Corrective Action Plan (CAP). The CAP was approved by the Nevada Division of Environmental Protection (NDEP) in June 1997. This site is identified in the Federal Facility Agreement and Consent Order (FFACO) as Corrective Action Site (CAS) Number 12-19-01 and is the only CAS in Corrective Action Unit (CAU) 339. The former Area 12 Fleet Operations Building 12-16 functioned as a maintenance facility for light- and heavy-duty vehicles from approximately 1965 to January 1993. Services performed at the site included steam-cleaning, tire service, and preventative maintenance on vehicles and equipment. Past activities impacted the former steam-cleaning discharge area with volatile organic compounds (VOCs) and total petroleum hydrocarbons (TPH) as oil.

NONE

1997-12-01T23:59:59.000Z

284

Interim report VII, production test IP-549-A half-plant low alum feed water treatment at F Reactor  

SciTech Connect

A half-plant low alum water treatment test began at F Reactor on January 16, 1963. The test, which had been prompted by the analysis of ledge corrosion attack on fuel elements, will demonstrate whether or not high alum feed is responsible for increasing the frequency of ledge and groove corrosion attack on fuel element surfaces. The effect will be evaluated by comparing visual examination results obtained from the normal production fuel irradiated in process water treated with two different alum feed rates. Six 20-column fuel discharges, ten columns from each side of the reactor, have been taken during the test as follows: (1) One discharge prior to the start of the test. (2) One discharge such that the test side was exposed to coolant treated with both high and low alum feed. (3) Four discharges under test conditions. This report discusses the results obtained from the fifth discharge under test conditions.

Geier, R.G.

1964-03-18T23:59:59.000Z

285

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

SciTech Connect

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).

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

286

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

SciTech Connect

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).

David W. Nigg

2013-09-01T23:59:59.000Z

287

Engineering considerations in the selection of the tokamak to follow the Tokamak Fusion Test Reactor (TFTR)  

SciTech Connect

The tokamak to follow the Tokamak Fusion Test Reactor (TFTR) should satisfy two important objectives. First, it should be a significant step in physics and engineering goals in order to maintain the level of progress which the US has established as the world leader in fusion energy development. The second objective should be to provide the information necessary to support the strategy and goals of the long-range Department of Energy (DOE) Fusion Program. In their Comprehensive Program Management Plan, the DOE identifies the need for a reactor technology program in the 1990s in which the major goal is to prove engineering feasibility. In this paper, the specific engineering needs are identified which have been developed through the tokamak design studies over the past decade. On the basis of these needs, it appears that several options are available for the next tokamak to follow TFTR. The final choice of the concept will involve consideration of the technical needs and the reality of the Fusion Program budget.

Shannon, T.E.

1983-01-01T23:59:59.000Z

288

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

E-Print Network (OSTI)

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 testing in order to more specifically document potential impacts of Marcellus Shale gas development

Manning, Sturt

289

Preparations for deuterium--tritium experiments on the Tokamak Fusion Test Reactor*  

Science Conference Proceedings (OSTI)

The final hardware modifications for tritium operation have been completed for the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. [bold 21], 1324 (1992)]. These activities include preparation of the tritium gas handling system, installation of additional neutron shielding, conversion of the toroidal field coil cooling system from water to a Fluorinert[sup TM] system, modification of the vacuum system to handle tritium, preparation, and testing of the neutral beam system for tritium operation and a final deuterium--deuterium (D--D) run to simulate expected deuterium--tritium (D--T) operation. Testing of the tritium system with low concentration tritium has successfully begun. Simulation of trace and high power D--T experiments using D--D have been performed. The physics objectives of D--T operation are production of [approx]10 MW of fusion power, evaluation of confinement, and heating in deuterium--tritium plasmas, evaluation of [alpha]-particle heating of electrons, and collective effects driven by alpha particles and testing of diagnostics for confined [alpha] particles. Experimental results and theoretical modeling in support of the D--T experiments are reviewed.

Hawryluk, R.J.; Adler, H.; Alling, P.; Ancher, C.; Anderson, H.; Anderson, J.L.; Anderson, J.W.; Arunasalam, V.; Ascione, G.; Aschroft, D.; Barnes, C.W.; Barnes, G.; Batchelor, D.B.; Bateman, G.; Batha, S.; Baylor, L.A.; Beer, M.; Bell, M.G.; Biglow, T.S.; Bitter, M.; Blanchard, W.; Bonoli, P.; Bretz, N.L.; Brunkhorst, C.; Budny, R.; Burgess, T.; Bush, H.; Bush, C.E.; Camp, R.; Caorlin, M.; Carnevale, H.; Chang, Z.; Chen, L.; Cheng, C.Z.; Chrzanowski, J.; Collazo, I.; Collins, J.; Coward, G.; Cowley, S.; Cropper, M.; Darrow, D.S.; Daugert, R.; DeLooper, J.; Duong, H.; Dudek, L.; Durst, R.; Efthimion, P.C.; Ernst, D.; Faunce, J.; Fonck, R.J.; Fredd, E.; Fredrickson, E.; Fromm, N.; Fu, G.Y.; Furth, H.P.; Garzotto, V.; Gentile, C.; Gettelfinger, G.; Gilbert, J.; Gioia, J.; Goldfinger, R.C.; Golian, T.; Gorelenkov, N.; Gouge, M.J.; Grek, B.; Grisham, L.R.; Hammett, G.; Hanson, G.R.; Heidbrink, W.; Hermann, H.W.; Hill, K.W.; Hirshman, S.; Hoffman, D.J.; Hosea, J.; Hulse, R.A.; Hsuan, H.; Ja

1994-05-01T23:59:59.000Z

290

Injectivity Test At Steamboat Springs Area (Combs, Et Al., 1999) | Open  

Open Energy Info (EERE)

Steamboat Springs Area (Combs, Et Steamboat Springs Area (Combs, Et Al., 1999) Exploration Activity Details Location Steamboat Springs Area Exploration Technique Injectivity Test Activity Date Usefulness not indicated DOE-funding Unknown Notes Part of the injection testing used downhole packers for isolating various zones and evaluating their permeability. By running the packers into the hole on N-rod ( 2.75"+K610 OD), the annulus was roughly the same cross-sectional area as the inside of the pipe. It was then possible to inject into either the zone above the packer or the one below, and compare the infectivity of those intervals. References Jim Combs, John T. Finger, Colin Goranson, Charles E. Hockox Jr., Ronald D. Jacobsen, Gene Polik (1999) Slimhole Handbook- Procedures And Recommendations For Slimhole Drilling And Testing In Geothermal Exploration

291

Thermal single-well injection-withdrawal tracer tests for determining fracture-matrix heat transfer area  

E-Print Network (OSTI)

Testing for Estimating Heat Transfer Area in FracturedFRACTURE-MATRIX HEAT TRANSFER AREA Karsten Pruess andimprove the flow and heat transfer characteristics of the

Pruess, K.

2011-01-01T23:59:59.000Z

292

Radiation shielding calculations for MuCool test area at Fermilab  

DOE Green Energy (OSTI)

The MuCool Test Area (MTA) is an intense primary beam facility derived directly from the Fermilab Linac to test heat deposition and other technical concerns associated with the liquid hydrogen targets being developed for cooling intense muon beams. In this shielding study the results of Monte Carlo radiation shielding calculations performed using the MARS14 code for the MuCool Test Area and including the downstream portion of the target hall and berm around it, access pit, service building, and parking lot are presented and discussed within the context of the proposed MTA experimental configuration.

Igor Rakhno; Carol Johnstone

2004-05-26T23:59:59.000Z

293

Corrective Action Plan for Corrective Action Unit 490: Station 44 Burn Area, Tonopah Test Range, Nevada  

Science Conference Proceedings (OSTI)

Corrective Action Unit (CAU) 490, Station 44 Burn Area is located on the Tonopah Test Range (TTR). CAU 490 is listed in the Federal Facility Agreement and Consent Order (FFACO, 1996) and includes for Corrective Action Sites (CASs): (1) Fire Training Area (CAS 03-56-001-03BA); (2) Station 44 Burn Area (CAS RG-56-001-RGBA); (3) Sandia Service Yard (CAS 03-58-001-03FN); and (4) Gun Propellant Burn Area (CAS 09-54-001-09L2).

K. B. Campbell

2002-04-01T23:59:59.000Z

294

THE COMPONENT TEST FACILITY – A NATIONAL USER FACILITY FOR TESTING OF HIGH TEMPERATURE GAS-COOLED REACTOR (HTGR) COMPONENTS AND SYSTEMS  

DOE Green Energy (OSTI)

The Next Generation Nuclear Plant (NGNP) and other High-Temperature Gas-cooled Reactor (HTGR) Projects require research, development, design, construction, and operation of a nuclear plant intended for both high-efficiency electricity production and high-temperature industrial applications, including hydrogen production. During the life cycle stages of an HTGR, plant systems, structures and components (SSCs) will be developed to support this reactor technology. To mitigate technical, schedule, and project risk associated with development of these SSCs, a large-scale test facility is required to support design verification and qualification prior to operational implementation. As a full-scale helium test facility, the Component Test facility (CTF) will provide prototype testing and qualification of heat transfer system components (e.g., Intermediate Heat Exchanger, valves, hot gas ducts), reactor internals, and hydrogen generation processing. It will perform confirmation tests for large-scale effects, validate component performance requirements, perform transient effects tests, and provide production demonstration of hydrogen and other high-temperature applications. Sponsored wholly or in part by the U.S. Department of Energy, the CTF will support NGNP and will also act as a National User Facility to support worldwide development of High-Temperature Gas-cooled Reactor technologies.

David S. Duncan; Vondell J. Balls; Stephanie L. Austad

2008-09-01T23:59:59.000Z

295

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

Open Energy Info (EERE)

Vale Hot Springs Area (Combs, Et Vale Hot Springs Area (Combs, Et Al., 1999) Exploration Activity Details Location Vale Hot Springs Area Exploration Technique Injectivity Test Activity Date Usefulness useful DOE-funding Unknown Notes Analysis of the two injection tests performed at the exploration slimhole site during May, 1995 yielded estimates for the permeability-thickness product (transmissivity) kh of 0.25 and 0.23 Da-fi, based on pressure fall off after injection (see Section IV-a). Using the pressure buildup for the second test, a transmissivity of 0.610 Da-ft was estimated. These estimates are approximately an order of magnitude smaller than the kh values estimated for the nearby A-Alt well which was tested in 1994. References Jim Combs, John T. Finger, Colin Goranson, Charles E. Hockox Jr.,

296

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

Open Energy Info (EERE)

Reese River Area (Henkle & Ronne, Reese River Area (Henkle & Ronne, 2008) Exploration Activity Details Location Reese River Area Exploration Technique Injectivity Test Activity Date Usefulness not indicated DOE-funding Unknown Notes On March 22, 2007 a brief injectivity test was preformed after the slotted liner had been installed. Water was injected at flow rates of 6.3 l/s, 13 l/s and 19 l/s and the pressure and temperature was recorded down hole at a depth of 926 m. At the higher flow rate, the test was interrupted several times to repair leaks at the surface. From the recorded pressure an approximate injectivity index of 10 l/s/MPa was calculated. References William R. Henkle, Joel Ronne (2008) Phase 2 Reese River Geothermal Project Slim Well 56-4 Drilling And Testing Retrieved from

297

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

Open Energy Info (EERE)

source source History View New Pages Recent Changes All Special Pages Semantic Search/Querying Get Involved Help Apps Datasets Community Login | Sign Up Search Page Edit History Facebook icon Twitter icon » Flow Test At Long Valley Caldera Area (Farrar, Et Al., 2003) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Long Valley Caldera Area (Farrar, Et Al., 2003) Exploration Activity Details Location Long Valley Caldera Area Exploration Technique Flow Test Activity Date Usefulness useful DOE-funding Unknown Notes The pressure data collected during a 50-h-long flow test at LVEW in September 2001 are best matched using solutions for a flow system consisting of a steeply dipping fracture with infinite hydraulic conductivity, surrounded by a finite-conductivity rock matrix. At shallow

298

Initial confinement studies of ohmically heated plasmas in the Tokamak Fusion Test Reactor  

DOE Green Energy (OSTI)

Initial operation of the Tokamak Fusion Test Reactor (TFTR) has concentrated upon confinement studies of ohmically heated hydrogen and deuterium plasmas. Total energy confinement times (tau/sub E/) are 0.1 to 0.2 s for a line-average density range (anti n/sub e/) of 1 to 2.5 x 10/sup 19/ m/sup -3/ with electron temperatures of T/sub e/(o) approx. 1.2 to 2.2 keV, ion temperatures of T/sub i/(o) approx. 0.9 to 1.5 keV, and Z/sub eff/ approx. 3. A comparison of PLT, PDX, and TFTR plasma confinement supports a dimension-cubed scaling law.

Efthimion, P.C.; Bell, M.; Blanchard, W.R.; Bretz, N.; Cecchi, J.L.; Coonrod, J.; Davis, S.; Dylla, H.F.; Fonck, R.; Furth, H.P.

1984-06-01T23:59:59.000Z

299

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

Science Conference Proceedings (OSTI)

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)

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

300

ARMY GAS-COOLED REACTOR SYSTEMS PROGRAM. INITIAL FULL POWER AND LIMITED ENDURANCE TESTS OF THE ML-1 NUCLEAR POWER PLANT. Final Test Report  

SciTech Connect

The evaluation of the data generated during the full power and limited endurance tests of the ML-1 mobile nuclear power plant indicates that the reactor performs in accordance with the design specifications. During the 101 hr test period, the reactor attained a maximum power of 3.44 Mw( and 247 kw(e) was measured at the output shaft of the turbine-compressor set. No operating limits were exceeded during these tests and all systems performed satisfactorily Except for the known performance deficiency of the turbinecompressor set, which prevented the attainment of design output power, no operational, stability, or control problems were encountered. All test objectives were achieved and the tests were considered completely successful. (auth)

Kattchee, N.

1963-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


301

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

DOE Green Energy (OSTI)

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.

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

2007-03-30T23:59:59.000Z

302

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

SciTech Connect

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.

S. Blaine Grover

2009-05-01T23:59:59.000Z

303

First-wall, blanket, and shield engineering test program for magnetically confined fusion power reactors  

Science Conference Proceedings (OSTI)

The key engineering areas identified for early study relate to FW/B/S system thermal-hydraulics, thermomechnics, nucleonics, electromagnetics, assembly, maintenance, and repair. Programmatic guidance derived frm planning exercises involving over thirty organizations (laboratories, industries, and universities) has indicated (1) that meaningful near term engineering testing should be feasible within the bounds of a modest funding base, (2) that there are existing facilities and expertise which can be profitably utilized in this testing, and (3) that near term efforts should focus on the measurement of engineering data and the verification/calibration of predictive methods for anticipated normal operational and transient FW/B/S conditions. The remainder of this paper discusses in more detail the planning strategies, proposed approach to near term testing, and longer range needs for integrated FW/B/S test facilities.

Maroni, V.A.

1980-01-01T23:59:59.000Z

304

Flow Test At Fenton Hill Hdr Geothermal Area (Grigsby, Et Al., 1983) | Open  

Open Energy Info (EERE)

Grigsby, Et Al., 1983) Grigsby, Et Al., 1983) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Fenton Hill Hdr Geothermal Area (Grigsby, Et Al., 1983) Exploration Activity Details Location Fenton Hill Hdr Geothermal Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding Unknown References C. O. Grigsby, J. W. Tester, P. E. Trujillo, D. A. Counce, J. Abbott, C. E. Holley, L. A. Blatz (1983) Rock-Water Interactions In Hot Dry Rock Geothermal Systems- Field Investigations Of In Situ Geochemical Behavior Retrieved from "http://en.openei.org/w/index.php?title=Flow_Test_At_Fenton_Hill_Hdr_Geothermal_Area_(Grigsby,_Et_Al.,_1983)&oldid=511312" Category: Exploration Activities What links here Related changes

305

Modeling-Computer Simulations At Nevada Test And Training Range Area  

Open Energy Info (EERE)

source source History View New Pages Recent Changes All Special Pages Semantic Search/Querying Get Involved Help Apps Datasets Community Login | Sign Up Search Page Edit History Facebook icon Twitter icon » Modeling-Computer Simulations At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Modeling-Computer Simulations At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Exploration Activity Details Location Nevada Test And Training Range Area Exploration Technique Modeling-Computer Simulations Activity Date Usefulness not indicated DOE-funding Unknown Notes Nellis Air Force Range (NAFR) occupies over 3 million acres in southern Nevada (Figure 1). We recently assessed potential utility-grade geothermal

306

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

Open Energy Info (EERE)

source source History View New Pages Recent Changes All Special Pages Semantic Search/Querying Get Involved Help Apps Datasets Community Login | Sign Up Search Page Edit History Facebook icon Twitter icon » Flow Test At Lake City Hot Springs Area (Warpinski, Et Al., 2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Flow Test At Lake City Hot Springs Area (Warpinski, Et Al., 2004) Exploration Activity Details Location Lake City Hot Springs Area Exploration Technique Flow Test Activity Date Usefulness not indicated DOE-funding Unknown Notes The Lake City site, which is located in far northeastern California, consists of a previously identified geothermal site that has been explored with both geophysics and drilling (Hedel, 1981), but has not been

307

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

SciTech Connect

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.

None

1996-12-01T23:59:59.000Z

308

Underground Test Area Subproject Phase I Data Analysis Task. Volume IV - Hydrologic Parameter Data Documentation Package  

Science Conference Proceedings (OSTI)

Volume IV 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 hydrologic parameter 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.

None

1996-09-01T23:59:59.000Z

309

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

Science Conference Proceedings (OSTI)

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.

None

1996-12-01T23:59:59.000Z

310

Lithology and Stratigraphy of Holes Drilled in LANL-Use Areas of the Nevada Test Site  

SciTech Connect

Geologic data for ten holes drilled in areas used by Los Alamos National Laboratory at the Nevada Test Site are presented in this report. The holes include emplacement holes, instrumentation holes, and Underground Test Area wells drilled during calendar years 1991 through 1995. For each hole a stratigraphic log, a detailed lithologic log, and one or two geologic cross sections are presented, along with a supplemental data sheet containing information about the drilling operations, geology, or references. For three of the holes, graphic data summary sheets with geologic and geophysical data are provided as plates.

Lance B. Prothro; Sigmund L. Drellack, Jr.; Brian M. Allen

1999-07-01T23:59:59.000Z

311

Plutonium-aerosol emission rates and potential inhalation exposure during cleanup and treatment test at Area 11, Nevada Test Site  

SciTech Connect

A Cleanup and Treatment (CAT) test was conducted in 1981 at Area 11, Nevada Test Site. Its purpose was to evaluate the effectiveness of using a large truck-mounted vacuum cleaner similar to those used to clean paved streets for cleaning radiological contamination from the surface of desert soils. We found that four passes with the vehicle removed 97% of the alpha contamination and reduced resuspension by 99.3 to 99.7%. Potential exposure to cleanup workers was slight when compared to natural background exposure. 7 refs., 1 fig., 2 tabs.

Shinn, J.H.; Homan, D.N.

1985-08-13T23:59:59.000Z

312

Compilation of modal analyses of volcanic rocks from the Nevada Test Site area, Nye County, Nevada  

SciTech Connect

Volcanic rock samples collected from the Nevada Test Site, Nye County, Nevada, between 1960 and 1985 were analyzed by thin section to obtain petrographic mode data. In order to provide rapid accessibility to the entire database, all data from the cards were entered into a computerized database. This computer format will enable workers involved in stratigraphic studies in the Nevada Test Site area and other locations in southern Nevada to perform independent analyses of the data. The data were compiled from the mode cards into two separate computer files. The first file consists of data collected from core samples taken from drill holes in the Yucca Mountain area. The second group of samples were collected from measured sections and surface mapping traverses in the Nevada Test Site area. Each data file is composed of computer printouts of tables with mode data from thin section point counts, comments on additional data, and location data. Tremendous care was taken in transferring the data from the cards to computer, in order to preserve the original information and interpretations provided by the analyzer. In addition to the data files above, a file is included that consists of Nevada Test Site petrographic data published in other US Geological Survey and Los Alamos National Laboratory reports. These data are presented to supply the user with an essentially complete modal database of samples from the volcanic stratigraphic section in the Nevada Test Site area. 18 refs., 4 figs.

Page, W.R.

1990-10-01T23:59:59.000Z

313

Use of the WECC WAMS in Wide Area Probing Tests for Validation of System Performance & Modeling  

Science Conference Proceedings (OSTI)

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.

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

2009-02-01T23:59:59.000Z

314

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

SciTech Connect

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.

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

2012-03-01T23:59:59.000Z

315

Radiation shielding issues for MuCool test area at Fermilab  

DOE Green Energy (OSTI)

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

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

2005-03-01T23:59:59.000Z

316

Deviation to the Test Program and Procedures for the 710 Critical Experiment Reactor Control Drum Mockup Experiment  

SciTech Connect

This document describes a deviation from the "Test Program and Procedures for the 710 Critical Experiment Reactor Control Drum Mockup Experiment," TM-64-3-706, which was made in accordance with ITS Standard Practice J80-81 on September 14, 1964. The deviation did not involve a significant change in the safety of the operation.

Sims, F.L.

1964-09-14T23:59:59.000Z

317

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

SciTech Connect

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.

S.J. Roberts

2007-03-20T23:59:59.000Z

318

System Engineering Program Applicability for the High Temperature Gas-Cooled Reactor (HTGR) Component Test Capability (CTC)  

SciTech Connect

This white paper identifies where the technical management and systems engineering processes and activities to be used in establishing the High Temperature Gas-cooled Reactor (HTGR) Component Test Capability (CTC) should be addressed and presents specific considerations for these activities under each CTC alternative

Jeffrey Bryan

2009-06-01T23:59:59.000Z

319

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

Science Conference Proceedings (OSTI)

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.

Robert M. Edwards

2003-09-28T23:59:59.000Z

320

Thermal analysis for a spent reactor fuel storage test in granite  

Science Conference Proceedings (OSTI)

A test is conducted in which spent fuel assemblies from an operating commercial nuclear power reactor are emplaced in the Climax granite at the US Department of Energy`s Nevada Test Site. In this generic test, 11 canisters of spent PWR fuel are emplaced vertically along with 6 electrical simulator canisters on 3 m centers, 4 m below the floor of a storage drift which is 420 m below the surface. Two adjacent parallel drifts contain electrical heaters, operated to simulate (in the vicinity of the storage drift) the temperature fields of a large repository. This test, planned for up to five years duration, uses fairly young fuel (2.5 years out of core) so that the thermal peak will occur during the time frame of the test and will not exceed the peak that would not occur until about 40 years of storage had older fuel (5 to 15 years out of core) been used. This paper describes the calculational techniques and summarizes the results of a large number of thermal calculations used in the concept, basic design and final design of the spent fuel test. The results of the preliminary calculations show the effects of spacing and spent fuel age. Either radiation or convection is sufficient to make the drifts much better thermal conductors than the rock that was removed to create them. The combination of radiation and convection causes the drift surfaces to be nearly isothermal even though the heat source is below the floor. With a nominal ventilation rate of 2 m{sup 3}/s and an ambient rock temperature of 23{sup 0}C, the maximum calculated rock temperature (near the center of the heat source) is about 100{sup 0}C while the maximum air temperature in the drift is around 40{sup 0}C. This ventilation (1 m{sup 3}/s through the main drift and 1/2 m{sup 3}/s through each of the side drifts) will remove about 1/3 of the heat generated during the first five years of storage.

Montan, D.N.

1980-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Corrective Action Decision Document for Corrective Action Unit 417: Central Nevada Test Area Surface, Nevada Appendix D - Corrective Action Investigation Report, Central Nevada Test Area, CAU 417  

SciTech Connect

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.

U.S. Department of Energy, Nevada Operations office

1999-04-02T23:59:59.000Z

322

Offsite environmental monitoring report. Radiation monitoring around United States nuclear test areas, calendar year 1982  

Science Conference Proceedings (OSTI)

A principal activity of the Offsite Radiological Safety Program is routine environmental monitoring for radioactive materials in various media and for radiation in areas which may be affected by nuclear tests. It is conducted to document compliance with standards, to identify trends, and to provide information to the public. This report summarizes these activities for CY 1982.

Black, S. C.; Grossman, R. F.; Mullen, A. A.; Potter, G. D.; Smith, D. D. [comps.

1983-07-01T23:59:59.000Z

323

Closure report for housekeeping category, Corrective Action Unit 349, Area 12, Nevada Test Site  

Science Conference Proceedings (OSTI)

This Closure Report summarizes the corrective actions which were completed at the Corrective Action Sites within Corrective Action Unit 349 Area 12 at the Nevada Test Site. Current site descriptions, observations and identification of wastes removed are included on FFACO Corrective Action Site housekeeping closure verification forms.

NONE

1998-01-01T23:59:59.000Z

324

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

Science Conference Proceedings (OSTI)

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

None

1997-10-01T23:59:59.000Z

325

Corrective Action Plan for Corrective Action Unit 453: Area 9 UXO Landfill, Tonopah Test Range, Nevada  

Science Conference Proceedings (OSTI)

This corrective action plan proposes the closure method for the area 9 unexploded Ordnance landfill, corrective action unit 453 located at the Tonopah Test Range. The area 9 UXO landfill consists of corrective action site no. 09-55-001-0952 and is comprised of three individual landfill cells designated as A9-1, A9-2, and A9-3. The three landfill cells received wastes from daily operations at area 9 and from range cleanups which were performed after weapons testing. Cell locations and contents were not well documented due to the unregulated disposal practices commonly associated with early landfill operations. However, site process knowledge indicates that the landfill cells were used for solid waste disposal, including disposal of UXO.

Bechtel Nevada

1998-09-30T23:59:59.000Z

326

Closure Report for Corrective Action Unit 407: Roller Coaster RADSAFE Area, Tonopah Test Range, Nevada  

SciTech Connect

This closure report (CR) provides documentation for the closure of the Roller Coaster RADSAFE Area (RCRSA) Corrective Action Unit (CAU) 407 identified in the Federal Facility Agreement and Consent Order (FFACO) (Nevada Division of Environmental Protection [NDEP] et al., 1996). CAU 407 is located at the Tonopah Test Range (TTR), Nevada. The TTR is approximately 225 kilometers (km) (140 miles [mi]) northwest of Las Vegas, Nevada (Figure 1). The RCRSA is located on the northeast comer of the intersection of Main Road and Browne's Lake Road, which is approximately 8 km (5 mi) south of Area 3 (Figure 1). The RCRSA was used during May and June of 1963 to decontaminate vehicles, equipment, and personnel from the Double Tracks and Clean Slate tests. Investigation of the RCRSA was conducted from June through November of 1998. A Corrective Action Decision Document (CADD) (U.S. Department of Energy, Nevada Operations Office [DOEN], 1999) was approved in October of 1999. The purpose of this CR is to: Document the closure activities as proposed in the Corrective Action Plan (CAP) (DOEM, 2000). Obtain a Notice of Completion from the NDEP. Recommend the movement of CAU 407 from Appendix III to Appendix IV of the FFACO. The following is the scope of the closure actions implemented for CAU 407: Removal and disposal of surface soils which were over three times background for the area. Soils identified for removal were disposed of at the Area 5 Radioactive Waste Management Site (RWMS) at the Nevada Test Site (NTS). Excavated areas were backfilled with clean borrow soil located near the site. A soil cover was constructed over the waste disposal pit area, where subsurface constituents of concern remain. The site was fenced and posted as an ''Underground Radioactive Material'' area.

T. M. Fitzmaurice

2001-12-01T23:59:59.000Z

327

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

SciTech Connect

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.

Irene Farnham

2011-05-01T23:59:59.000Z

328

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

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Emission Testing of Washington Emission Testing of Washington Metropolitan Area Transit Authority (WMATA) Natural Gas and Diesel Transit Buses M. Melendez, J. Taylor, and J. Zuboy National Renewable Energy Laboratory W.S. Wayne West Virginia University D. Smith U.S. Department of Energy Technical Report NREL/TP-540-36355 December 2005 Emission Testing of Washington Metropolitan Area Transit Authority (WMATA) Natural Gas and Diesel Transit Buses M. Melendez, J. Taylor, and J. Zuboy National Renewable Energy Laboratory W.S. Wayne West Virginia University D. Smith U.S. Department of Energy Prepared under Task No. FC05-9000 Technical Report NREL/TP-540-36355 December 2005 National Renewable Energy Laboratory 1617 Cole Boulevard, Golden, Colorado 80401-3393 303-275-3000 * www.nrel.gov

329

AREA  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

AREA AREA FAQ # Question Response 316 vs DCAA FAQ 1 An inquiry from CH about an SBIR recipient asking if a DCAA audit is sufficient to comply with the regulation or if they need to add this to their audit they have performed yearly by a public accounting firm. 316 audits are essentially A-133 audits for for-profit entities. They DO NOT replace DCAA or other audits requested by DOE to look at indirect rates or incurred costs or closeouts. DCAA would never agree to perform A-133 or our 316 audits. They don't do A-133 audits for DOD awardees. The purpose of the audits are different, look at different things and in the few instances of overlap, from different perspectives. 316

330

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

Science Conference Proceedings (OSTI)

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.

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

2006-09-30T23:59:59.000Z

331

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

SciTech Connect

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).

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

2012-02-01T23:59:59.000Z

332

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

Science Conference Proceedings (OSTI)

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.

Noel Duckwitz

2011-06-01T23:59:59.000Z

333

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

Science Conference Proceedings (OSTI)

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.

Noel Duckwitz

2011-06-01T23:59:59.000Z

334

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

Science Conference Proceedings (OSTI)

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.

Noel Duckwitz

2011-06-01T23:59:59.000Z

335

Investigation of global Alfven instabilities in the Tokamak Fusion Test Reactor  

SciTech Connect

Toroidal Alfven eigenmodes (TAE) were excited by the energetic neutral beam ions tangentially injected into plasmas at low magnetic field in the Tokamak Fusion Test Reactor (TFTR) ({ital Proceedings} {ital of} {ital the} 11{ital th} {ital International} {ital Conference} {ital on} {ital Plasma} {ital Physics} {ital and} {ital Controlled} {ital Fusion} {ital Research} (IAEA, Vienna, 1987), Vol. 1, p. 51). The injection velocities were comparable to the Alfven speed. The modes were identified by measurements from Mirnov coils and beam emission spectroscopy (BES). TAE modes appear in bursts whose repetition rate increases with beam power. The neutron emission rate exhibits sawtoothlike behavior and the crashes always coincide with TAE bursts. This indicates ejection of fast ions from the plasma until these modes are stabilized. The dynamics of growth and stabilization were investigated at various plasma currents and magnetic fields. The results indicate that the instability can effectively clamp the number of energetic ions in the plasmas. The observed instability threshold is discussed in light of recent theories. In addition to these TAE modes, intermittent oscillations at three times the fundamental TAE frequency were observed by Mirnov coils, but no corresponding signal was found in BES. It appears that these high-frequency oscillations do not have a direct effect on the plasma neutron source strength.

Wong, K.L.; Durst, R.; Fonck, R.J.; Paul, S.F.; Roberts, D.R.; Fredrickson, E.D.; Nazikian, R.; Park, H.K.; Bell, M.; Bretz, N.L.; Budny, R.; Cheng, C.Z.; Cohen, S.; Hammett, G.W.; Jobes, F.C.; Johnson, L.; Meade, D.M.; Medley, S.S.; Mueller, D.; Nagayama, Y.; Owens, D.K.; Sabbagh, S.; Synakowski, E.J. (Princeton Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States))

1992-07-01T23:59:59.000Z

336

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

Science Conference Proceedings (OSTI)

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.

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

337

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

Science Conference Proceedings (OSTI)

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.

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

2011-03-01T23:59:59.000Z

338

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

SciTech Connect

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.

NONE

1994-02-01T23:59:59.000Z

339

Corrective Action Investigation Plan for Corrective Action Unit 271: Areas 25, 26, and 27 Septic Systems, Nevada Test Site, Nevada (Rev. 0, April 2001)  

DOE Green Energy (OSTI)

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) 271 under the Federal Facility Agreement and Consent Order. Corrective Action Unit 271 consists of 15 Corrective Action Sites (CASs) including: thirteen Septic Systems (25-04-01, 25-04-03, 25-04-04, 25-04-08, 25-04-09, 25-04-10, 25-04-11, 26-04-01, 26-04-02, 26-05-03, 26-05-04, 26-05-05, and 27-05-02), one Contaminated Water Reservoir (26-03-01), and one Radioactive Leachfield (26-05-01). The CASs addressed by CAU 271 are located at Guard Station 500, the Reactor Control Point (RCP), Bare Reactor Experiment - Nevada Tower, and Engine Test State-1 (ETS-1) facilities in Area 25; the Port Gaston and Project Pluto facilities in Area 26; and the Baker Site in Area 27 of the Nevada Test Site. Between 1 958 and 1973, the RCP and ETS-1 facilities supported the development and testing of nuclear reactors for space propulsion as part of the Nuclear Rocket Development Station. The Project Pluto facilities supported nuclear reactor testing for use as a ramjet propulsion system between 1961 and 1964, followed by similar use for other projects through the early 1980s. The Baker Site facilities were constructed in the 1960s to serve as the staging point where the manufactured components of nuclear devices were assembled, disassembled, and modified. The scope of the investigation strategy at these sites will involve biased and random soil sampling in leachfields using excavation (with drilling as a contingency), collection of soil samples underlying the base of proximal and distal ends of septic tanks and distal ends of distribution structures, defining the lateral and vertical extent of contamination through discrete field and possible stepout location sampling, collection system line inspection, and geotechnical/ hydrological laboratory analyses. Based on site history and existing characterization data gathered to support the Data Quality Objectives process, it is believed that the contaminants of potential concern at these CASs may include: volatile organic compounds, fecal coliform bacteria, semivolatile organic compounds, petroleum hydrocarbons, Resource Conservation and Recovery Act metals, gamma-emitting radionuclides, isotopic uranium, strontium-90, tritium, beryllium, polychlorinated biphenyls, pesticides, and herbicides. The results of this field investigation will support a defensible evaluation of corrective action alternatives in the Corrective Action Decision Document.

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

2001-04-09T23:59:59.000Z

340

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

SciTech Connect

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.

David W. Nigg; Devin A. Steuhm

2011-09-01T23:59:59.000Z

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341

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

SciTech Connect

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.

David W. Nigg; Devin A. Steuhm

2011-09-01T23:59:59.000Z

342

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

SciTech Connect

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.

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

2012-09-01T23:59:59.000Z

343

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  

DOE Green Energy (OSTI)

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).

Susan Evans

2004-11-01T23:59:59.000Z

344

Borehole and geohydrologic data for test hole USW UZ-6, Yucca Mountain area, Nye County, Nevada  

SciTech Connect

Test hole USW UZ-6, located 1.8 kilometers west of the Nevada Test Site on a major north-trending ridge at Yucca Mountain, was dry drilled in Tertiary tuff to a depth of 575 meters. The area near this site is being considered by the US Department of Energy for potential construction of a high-level, radioactive-waste repository. Test hole USW UZ-6 is one of seven test holes completed in the unsaturated zone as part of the US Geological Survey`s Yucca Mountain Project to characterize the potential repository site. Data pertaining to borehole drilling and construction, lithology of geologic units penetrated, and laboratory analyses for hydrologic characteristics of samples of drill-bit cuttings are included in this report.

Whitfield, M.S. Jr.; Loskot, C.L. [Geological Survey, Denver, CO (United States); Cope, C.M. [Foothill Engineering Consultants, Inc., Golden, CO (United States)

1993-04-01T23:59:59.000Z

345

Geothermometry At Nevada Test And Training Range Area (Sabin, Et Al., 2004)  

Open Energy Info (EERE)

2004) 2004) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Geothermometry At Nevada Test And Training Range Area (Sabin, Et Al., 2004) Exploration Activity Details Location Nevada Test And Training Range Area Exploration Technique Geothermometry Activity Date Usefulness not indicated DOE-funding Unknown Notes Groundwater data are limited to a portion of NAFR; data are more plentiful beyond the range boundaries. Geothermometry yields calculated groundwater temperatures generally ranging from 30 to 105degrees C, with a rough correlation between the SiO2-chalcedony and the Na-K-Na (Mg-corrected) geothermometers. References A. E. Sabin, J. D. Walker, J. Unruh, F. C. Monastero (2004) Toward The Development Of Occurrence Models For Geothermal Resources In The

346

Treatability Test Plan for 300 Area Uranium Stabilization through Polyphosphate Injection  

Science Conference Proceedings (OSTI)

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.

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

347

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

Science Conference Proceedings (OSTI)

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

NSTec Environmental Restoration

2008-08-01T23:59:59.000Z

348

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

Science Conference Proceedings (OSTI)

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.

Dautel, W.A.

1996-10-01T23:59:59.000Z

349

Hydrologic Resources Management Program and Underground Test Area Project FY2005 Progress Report  

Science Conference Proceedings (OSTI)

This report describes FY 2005 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 five 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 Bechtel Nevada (BN).

Eaton, G F; Genetti, V; Hu, Q; Hudson, G B; Kersting, A B; Lindvall, R E; Moran, J E; Nimz, G J; Ramon, E C; Rose, T P; Shuller, L; Williams, R W; Zavarin, M; Zhao, P

2007-03-23T23:59:59.000Z

350

Materials Reliability Program, Reactor Vessel Head Boric Acid Corrosion Testing (MRP-165)  

Science Conference Proceedings (OSTI)

Pressurized water reactor (PWR) coolant leakage from stress corrosion cracking of an Alloy 600 control rod drive mechanism (CRDM) penetration has led to one case of severe corrosion and cavity formation in a low-alloy steel reactor vessel head (RVH). The detailed progression of RVH wastage following initial leakage is complicated and probably involves several corrosion mechanisms. The Materials Reliability Program (MRP) has completed three tasks of a comprehensive program to examine postulated sequential...

2005-12-14T23:59:59.000Z

351

McClellan Nuclear Radiation Center (MNRC) TRIGA reactor: The national organization of test research and training reactors  

SciTech Connect

This year's TRTR conference is being hosted by the McClellan Nuclear Radiation Center. The conference will be held at the Red Lion Hotel in Sacramento, CA. The conference dates are scheduled for October 11-14, 1994. Deadlines for sponsorship commitment and papers have not been set, but are forthcoming. The newly remodeled Red Lion Hotel provides up-to-date conference facilities and one of the most desirable locations for dining, shopping and entertainment in the Sacramento area. While attendees are busy with the conference activities, a spouses program will be available. Although the agenda has not been set, the Sacramento area offers outings to San Francisco, Pier 39, Ghirardelli Square (famous for their chocolate), and a chance to discover 'El Dorado' in the gold country. Not to forget our own bit of history with visits to 'Old Sacramento and Old Folsom', where antiquities abound, to the world renown train museum and incredible eating establishments. (author)

Kiger, Kevin M. [SWI-ALC/TIR, 5335 Price Ave., McClellan Air Force Base Sacramento, CA 95652-2504 (United States)

1994-07-01T23:59:59.000Z

352

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

SciTech Connect

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.

Owen, M.B.

1997-04-01T23:59:59.000Z

353

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

Science Conference Proceedings (OSTI)

In Appendix A the second sentence of the first paragraph on Page A-1-1 of the Streamlined Approach for Environmental Restoration Plan for Corrective Action Unit 425: Area 9 Main Lake Construction Debris Disposal Area, Tonopah Test Range, Nevada, erroneously cites the EPA DQO guidance outline as (EPA, 1994). The correct citation is (EPA, 2000).

K. B. Campbell

2003-03-01T23:59:59.000Z

354

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

DOE Green Energy (OSTI)

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.

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

1993-02-01T23:59:59.000Z

355

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

SciTech Connect

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

D. L. Porter; H.C. Tsai

2012-08-01T23:59:59.000Z

356

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

Science Conference Proceedings (OSTI)

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.

K. B. Campbell

2003-04-01T23:59:59.000Z

357

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

SciTech Connect

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.

Stoller-Navarro Joint Venture

2004-06-24T23:59:59.000Z

358

Corrective Action Plan for Corrective Action Unit 261: Area 25 Test Cell A Leachfield System, Nevada Test Site, Nevada  

Science Conference Proceedings (OSTI)

This Corrective Action Plan (CAP) has been prepared for the Corrective Action Unit (CAU)261 Area 25 Test Cell A Leachfield System in accordance with the Federal Facility and Consent Order (Nevada Division of Environmental Protection [NDEP] et al., 1996). This CAP provides the methodology for implementing the approved corrective action alternative as listed in the Corrective Action Decision Document (U.S. Department of Energy, Nevada Operations Office, 1999). Investigation of CAU 261 was conducted from February through May of 1999. There were no Constituents of Concern (COCs) identified at Corrective Action Site (CAS) 25-05-07 Acid Waste Leach Pit (AWLP). COCs identified at CAS 25-05-01 included diesel-range organics and radionuclides. The following closure actions will be implemented under this plan: Because COCs were not found at CAS 25-05-07 AWLP, no action is required; Removal of septage from the septic tank (CAS 25-05-01), the distribution box and the septic tank will be filled with grout; Removal of impacted soils identified near the initial outfall area; and Upon completion of this closure activity and approval of the Closure Report by NDEP, administrative controls, use restrictions, and site postings will be used to prevent intrusive activities at the site.

T. M. Fitzmaurice

2000-08-01T23:59:59.000Z

359

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

Science Conference Proceedings (OSTI)

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.

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

360

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

SciTech Connect

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.

NSTec Environmental Management

2008-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

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

Science Conference Proceedings (OSTI)

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.

NONE

1997-04-01T23:59:59.000Z

362

Standard Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reactor Surveillance, E706(IIIB)  

E-Print Network (OSTI)

1.1 This test method describes the use of solid-state track recorders (SSTRs) for neutron dosimetry in light-water reactor (LWR) applications. These applications extend from low neutron fluence to high neutron fluence, including high power pressure vessel surveillance and test reactor irradiations as well as low power benchmark field measurement. (1) This test method replaces Method E 418. This test method is more detailed and special attention is given to the use of state-of-the-art manual and automated track counting methods to attain high absolute accuracies. In-situ dosimetry in actual high fluence-high temperature LWR applications is emphasized. 1.2 This test method includes SSTR analysis by both manual and automated methods. To attain a desired accuracy, the track scanning method selected places limits on the allowable track density. Typically good results are obtained in the range of 5 to 800 000 tracks/cm2 and accurate results at higher track densities have been demonstrated for some cases. (2) Trac...

American Society for Testing and Materials. Philadelphia

2003-01-01T23:59:59.000Z

363

Three-Dimensional Bayesian Geostatistical Aquifer Characterization at the Hanford 300 Area using Tracer Test Data  

SciTech Connect

Tracer testing under natural or forced gradient flow holds the potential to provide useful information for characterizing subsurface properties, through monitoring, modeling and interpretation of the tracer plume migration in an aquifer. Non-reactive tracer experiments were conducted at the Hanford 300 Area, along with constant-rate injection tests and electromagnetic borehole flowmeter (EBF) profiling. A Bayesian data assimilation technique, the method of anchored distributions (MAD) [Rubin et al., 2010], was applied to assimilate the experimental tracer test data with the other types of data and to infer the three-dimensional heterogeneous structure of the hydraulic conductivity in the saturated zone of the Hanford formation. In this study, the Bayesian prior information on the underlying random hydraulic conductivity field was obtained from previous field characterization efforts using the constant-rate injection tests and the EBF data. The posterior distribution of the conductivity field was obtained by further conditioning the field on the temporal moments of tracer breakthrough curves at various observation wells. MAD was implemented with the massively-parallel three-dimensional flow and transport code PFLOTRAN to cope with the highly transient flow boundary conditions at the site and to meet the computational demands of MAD. A synthetic study proved that the proposed method could effectively invert tracer test data to capture the essential spatial heterogeneity of the three-dimensional hydraulic conductivity field. Application of MAD to actual field data shows that the hydrogeological model, when conditioned on the tracer test data, can reproduce the tracer transport behavior better than the field characterized without the tracer test data. This study successfully demonstrates that MAD can sequentially assimilate multi-scale multi-type field data through a consistent Bayesian framework.

Chen, Xingyuan; Murakami, Haruko; Hahn, Melanie S.; Hammond, Glenn E.; Rockhold, Mark L.; Zachara, John M.; Rubin, Yoram

2012-06-01T23:59:59.000Z

364

Solar reforming of methane in a direct absorption catalytic reactor on a parabolic dish: I-test and analysis  

DOE Green Energy (OSTI)

The concept of solar driven chemical reaction in a commercial-scale volumetric receiver/reactor on a parabolic concentrator was successfully demonstrated in the CAtalytically Enhanced Solar Absorption Receiver (CAESAR) test. Solar reforming of methane (CH[sub 4]) with carbon dioxide (CO[sub 2]) was achieved in a 64 cm diameter direct absorption reactor on a parabolic dish capable of 150 kW solar power. The reactor was a catalytic volumetric absorber consisting of a multilayered, porous alumina foam disk coated with rhodium (Rh) catalyst. The system was operated during both steady-state and solar transient (cloud passage) conditions. The total solar power absorbed reached values up to 97 kW and the maximum methane conversion was 70%. Receiver thermal efficiencies ranged up to 85% and chemical efficiencies peaked at 54%. The absorber performed satisfactorily in promoting the reforming reaction during the tests without carbon formation. However, problems of cracking and degradation of the porous matrix, nonuniform dispersion of the Rh through the absorber, the catalyst deactivation due to sintering and possible encapsulation, must be resolved to achieve long-term operation and eventual commercialization.

Muir, J.F.; Hogan, R.E. Jr.; Skocypec, R.D. (Sandia National Lab., Albuquerque, NM (United States)); Buck, R. (DLR-ITT, Stuttgart (Germany))

1994-06-01T23:59:59.000Z

365

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

Science Conference Proceedings (OSTI)

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).

T. M. Fitzmaurice

2001-04-01T23:59:59.000Z

366

Standard Test Method for Gravimetric Determination of Nonvolatile Residue (NVR) in Environmentally Controlled Areas for Spacecraft  

E-Print Network (OSTI)

1.1 This test method covers the determination of nonvolatile residue (NVR) fallout in environmentally controlled areas used for the assembly, testing, and processing of spacecraft. 1.2 The NVR of interest is that which is deposited on sampling plate surfaces at room temperature: it is left to the user to infer the relationship between the NVR found on the sampling plate surface and that found on any other surfaces. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. 1.4 The values stated in SI units are to be regarded as standard. No other units of measurement are included in this standard.

American Society for Testing and Materials. Philadelphia

2008-01-01T23:59:59.000Z

367

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

SciTech Connect

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.

NSTec Environmental Management

2008-01-01T23:59:59.000Z

368

How accurately can one test CPT conservation with reactor and solar neutrino experiments?  

E-Print Network (OSTI)

We show that the combined data from solar neutrino experiments and from the KamLAND reactor neutrino experiment can establish an upper limit on, or detect, potential CPT violation in the neutrino sector of order 10^{-20} GeV to 10^{-21} GeV.

John N. Bahcall; V. Barger; Danny Marfatia

2002-01-23T23:59:59.000Z

369

Knoxville Area Transit: Propane Hybrid ElectricTrolleys; Advanced Technology Vehicles in Service, Advanced Vehicle Testing Activity (Fact Sheet)  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

website and in print publications. website and in print publications. TESTING ADVANCED VEHICLES KNOXVILLE AREA TRANSIT â—† PROPANE HYBRID ELECTRIC TROLLEYS Knoxville Area Transit PROPANE HYBRID ELECTRIC TROLLEYS NREL/PIX 13795 KNOXVILLE AREA TRANSIT (KAT) is recognized nationally for its exceptional service to the City of Knoxville, Tennessee. KAT received the American Public Transportation Associa- tion's prestigious Outstanding Achievement Award in 2004.

370

Corrective action plan for CAU Number 339: Area 12 Fleet Operations, Steam Cleaning Discharge Area, Nevada Test Site  

Science Conference Proceedings (OSTI)

The purpose of this Corrective Action Plan (CAP) is to provide the method for implementing the corrective action alternative as provided in the Corrective Action Decision Document (CADD). Detailed information of the site history and results of previous characterizations can be found in the Work Plan, the Preliminary Investigation Report, and the Phase 2 Characterization Report. Previous characterization investigations were completed as a condition of the Temporary Water Pollution Control Permit issued by the Nevada Division of Environmental Protection (NDEP) on July 14, 1992. The scope of this report is to prepare a CAP based upon the selected remedial alternative for closure of the Area 12, Building 12-16 Fleet Operations steam cleaning discharge area. The effluent discharge area has been impacted by volatile organic compounds (VOCs) and total petroleum hydrocarbons (TPH) as oil. The maximum hydrocarbon and VOC concentrations detected in the Preliminary and Phase 2 Site Characterization Investigations are summarized.

NONE

1997-05-01T23:59:59.000Z

371

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

SciTech Connect

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.

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

2012-10-01T23:59:59.000Z

372

PROCEEDINGS OF THE AEC SYMPOSIUM FOR CHEMICAL PROCESSING OF IRRADIATED FUELS FROM POWER, TEST, AND RESEARCH REACTORS, RICHLAND, WASHINGTON, OCTOBER 20 AND 21, 1959  

SciTech Connect

A review is presented in this symposium of the technology currently available for processing spent fuels from research, test, and power reactors. Twenty-one papers are included. Separate abstracts have been prepared for each paper. (W.L.H.)

1960-01-01T23:59:59.000Z

373

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

SciTech Connect

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.

David E. Korns

2012-05-01T23:59:59.000Z

374

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

Science Conference Proceedings (OSTI)

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.

IT Corporation, Las Vegas

2002-04-24T23:59:59.000Z

375

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  

Science Conference Proceedings (OSTI)

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.

Not Available

1992-07-01T23:59:59.000Z

376

AND OTHER TEST AREAS USED FOR U N D E R G R O U N  

Office of Legacy Management (LM)

AND OTHER TEST AREAS USED FOR U AND OTHER TEST AREAS USED FOR U N D E R G R O U N D NUCLEAR .DETONATIONS -9.\c January through December 1996 by the Monitoring Applications Laboratory National Enviscnmental Research Center LT . S . EFTtPlgRO%RIENFA& PROTECTIQN AGENCY LPS Vegas, Nevada -"& -% ~d*.".::. Published Hay 1975 This work p e r f o w e d under a Memorandum o f Understanding No. AT(26-1)-539) for the U. S. ENERGY RESEARCH B D E V E L O P M E X T ABMINISTRATIQN d ~ P v . a - r . . . - . -.. . . . . * . "+ . - : I - : : - ... 1-11.; ~ ~ ~ % ~ ! ~ $ ' ; : L : ; ~ : ~ ~ ~ ~ . . T h i s r e p o r t was prepared a s an account of work sponsored by t h e United S t a t e s Government. N e i t h e r t h e United S t a t e s nor t h e United S t a t e s Energy Research and Development A d m i n i s t r a t i o n , n o r any of t h e i r employees, nor any of t h e i r c o n

377

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

Science Conference Proceedings (OSTI)

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.

D. H. Cox

2001-06-01T23:59:59.000Z

378

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

SciTech Connect

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

PETERSEN SW

2009-09-24T23:59:59.000Z

379

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

Science Conference Proceedings (OSTI)

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

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

2007-06-30T23:59:59.000Z

380

Corrective Action Plan for Corrective Action Unit 407: Roller Coaster RADSAFE Area, Tonopah Test Range, Nevada  

Science Conference Proceedings (OSTI)

This Corrective Action Plan (CAP) has been prepared for the Roller Coaster RADSAFE Area Corrective Action Unit 407 in accordance with the Federal Facility and Consent Order (Nevada Division of Environmental Protection [NDEP] et al., 1996). This CAP provides the methodology for implementing the approved Corrective Action Alternative as listed in the Corrective Action Decision Document (U.S. Department of Energy, Nevada Operations Office, 1999). The RCRSA was used during May and June of 1963 to decontaminate vehicles, equipment, and personnel from the Clean Slate tests. The Constituents of Concern (COCs) identified during the site characterization include plutonium, uranium, and americium. No other COCS were identified. The following closure actions will be implemented under this plan: (1) Remove and dispose of surface soils which are over three times background for the area. Soils identified for removal will be disposed of at an approved disposal facility. Excavated areas will be backfilled with clean borrow soil fi-om a nearby location. (2) An engineered cover will be constructed over the waste disposal pit area where subsurface COCS will remain. (3) Upon completion of the closure and approval of the Closure Report by NDEP, administrative controls, use restrictions, and site postings will be used to prevent intrusive activities at the site. Barbed wire fencing will be installed along the perimeter of this unit. Post closure monitoring will consist of site inspections to determine the condition of the engineered cover. Any identified maintenance and repair requirements will be remedied within 90 working days of discovery and documented in writing at the time of repair. Results of all inspections/repairs for a given year will be addressed in a single report submitted annually to the NDEP.

T. M. Fitzmaurice

2000-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "test reactor area" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Utilization of MIT research reactor by Boston-area universities. Final report, July 1, 1978-June 30, 1980  

SciTech Connect

The guest institutions which used the MITR reactor during 1978-1979 are listed. The participation during previous years since the program began at MIT is also indicated. A summary of the type and amount of participation by local institutions is given.

Clark, L. Jr.; Janghorbani, M.

1981-07-01T23:59:59.000Z

382

FIRST SODIUM REACTOR EXPERIMENT (SRE) TEST OF HALLAM NUCLEAR POWER FACILITY (HNPF) CONTROL MATERIALS  

SciTech Connect

An experiment was conducted in the SRE to measure temperatures and neutron flux levels in and near a boron-containing simulated control rod. The data are being used to check analytical methods developed for prediction of control rod heat generation rates and maximum temperatures in this type of control rod in the Hallam Nuclear Power Facility. The maximum observed temperatures with a reactor power level of 20 Mw were 1363 deg F for a boron-- nickel alloy ring having a 0.105-in. radial clearance with the thimble and 1100 deg F for a boron -nickel alloy ring having a 0.020-in. radial clearance. The maximum temperature difference between the coolant and the control rod was 473 deg F. It is concluded that the expected greater heat generation rates in the Hallam reactor would prohibit the use of boron-containing absorber materials in a combined a him-safety rod. (auth)

Arneson, S.O.

1959-06-01T23:59:59.000Z

383

Variation of the magnetic susceptibility of artificial graphite with exposure in the materials testing reactor  

SciTech Connect

The magnetic susceptibility of artificial graphite was determined as a function of exposure in the MTR. Specimens were studied with exposures ranging from 0.07 to 82 {times} 10{sup18} nvt. Fluxes were determined by means of x-ray measurements and resistivity measurements. The dependence of the magnetic susceptibility on exposure in the MTR and also in a Hanford reactor are graphed, and an equivalence factor is calculated.

McCelland, J.D.

1955-02-23T23:59:59.000Z

384

Materials Reliability Program: Reactor Vessel Head Boric Acid Corrosion Testing (MRP-199)  

Science Conference Proceedings (OSTI)

PWR coolant leakage from stress corrosion cracking of an Alloy 600 control rod drive mechanism (CRDM) penetration has led to one case of severe corrosion and cavity formation in a low-alloy steel reactor vessel head (RVH). The detailed progression of RVH wastage following initial leakage is complicated and probably involves several corrosion mechanisms. The Materials Reliability Program (MRP) has completed three tasks of a comprehensive program to examine postulated sequential stages of boric acid corros...

2007-06-27T23:59:59.000Z

385

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

Science Conference Proceedings (OSTI)

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.

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

386

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

SciTech Connect

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.).

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

387

Closure plan for CAU No. 93: Area 6 steam cleaning effluent ponds, Nevada Test Site  

SciTech Connect

The steam cleaning effluent ponds (SCEP) waste unit is located in Area 6 at the Nevada Test Site (NTS). Nevada Operations Office operates the NTS and has entered into a trilateral agreement with the State of Nevada and the Defense Special Weapons Agency (DSWA). The trilateral agreement provides a framework for identifying, characterizing, remediating, and closing environmental sites on the NTS and associated bombing ranges. The SCEP waste unit consists of: two steam cleaning effluent ponds; layout pad and associated grease trap; Building 6-623 steam cleaning pad; test pad; Building 6-623 grease trap; Building 6-800 steam cleaning pad; Building 6-800 separator; Building 6-621 sump; and the concrete asbestos piping connecting these components to both SCEPs. Clean closure is the recommended closure strategy for the majority of the components within this CAU. Four components of the unit (Building 6-621 Sump, Test Pad Grease Trap, Building 6-623 Steam Cleaning Pad, and North SCEP pipeline) are recommended to be closed in place. This closure plan provides the strategy and backup information necessary to support the clean closure of each of the individual components within CAU 93. Analytical data generated during the characterization field work and earlier sampling events indicates the majority of CAU 93 soil and infrastructure is non-hazardous (i.e., impacted primarily with petroleum hydrocarbons).

NONE

1997-04-01T23:59:59.000Z

388

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

SciTech Connect

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

Field, J.G.

1994-06-10T23:59:59.000Z

389

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

Science Conference Proceedings (OSTI)

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.

Chaloud, D.J; Daigler, D.M.; Davis, M.G. [and others

1996-06-01T23:59:59.000Z

390

On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor  

Science Conference Proceedings (OSTI)

IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

Ayman I. Hawari; Mohamed A. Bourham

2010-04-22T23:59:59.000Z

391

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

SciTech Connect

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.

Blaine Grover

2012-10-01T23:59:59.000Z

392

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

Science Conference Proceedings (OSTI)

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.

Blaine Grover

2012-10-01T23:59:59.000Z

393

Brief summary of reactor core component welding for the Fast Flux Test Facility (FFTF)  

SciTech Connect

Included are descriptions of welding methods and joint design, welding equipment, and qualification tests. (DG)

Brown, W.F.

1974-04-15T23:59:59.000Z

394

In Situ Bioremediation Interim Remedial Action Report, Test Area North, Operable Unit 1-07B  

E-Print Network (OSTI)

This Interim Remedial Action Report is for the in situ bioremediation remedial component of Operable Unit 1-07B at Test Area North at the Idaho National Laboratory. Under U.S. Environmental Protection Agency guidance, an interim report for a long-term groundwater remedial action provides a chronology of events and a description of the remedial action facilities, systems, components, and operating documents that lead to a declaration that the system is operational and functional. It is the conclusion of this report that the in situ bioremediation remedial component includes the infrastructure and programs necessary to achieve the objectives of the in situ bioremediation remedial component for contaminated groundwater in the vicinity of the TSF-05 well; therefore, it can be deemed operational and functional. iii ivCONTENTS ABSTRACT.................................................................................................................................................iii

Unit -b; Prepared For The

2009-01-01T23:59:59.000Z

395

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

Science Conference Proceedings (OSTI)

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.

NSTec Environmental Management

2007-09-01T23:59:59.000Z