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

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

2

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

3

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

4

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

5

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

6

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

7

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

8

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

9

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

10

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

11

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

12

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

13

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

14

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

15

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

16

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

17

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

18

NUCLEAR REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel plate is designed for incorporation into control rods of the type utilized in high-flux test reactors. The fuel plate is designed so that the portion nearest the poison section of the control rod contains about one-half as much fissionable material as in the rest of the plate, thereby eliminating dangerous flux peaking in that portion. (AEC)

Currier, E.L. Jr.; Nicklas, J.H.

1963-06-11T23:59:59.000Z

19

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

20

FUEL ASSAY REACTOR  

DOE Patents (OSTI)

A reactor having maximum sensitivity to perturbations is described comprising a core consisting of a horizontally disposed, rectangular, annular fuel zone containing enriched uranium dioxide dispersed in graphite, the concentration of uranium dioxide increasing from the outside to the inside of the fuel zone, an internal reflector of graphite containing an axial test opening disposed within the fuel zone, an external graphite reflector, means for changing the neutron spectrum in the test opening, and means for measuring perturbations in the neutron flux caused by the introduction of different fuel elements into the test opening. (AEC)

Spinrad, B.I.; Sandmeier, H.A.; Martens, F.H.

1962-12-25T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

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

22

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

23

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

24

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

25

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

26

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

27

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

28

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

29

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

30

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

31

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

32

NUCLEAR REACTOR FUEL SYSTEMS  

DOE Patents (OSTI)

Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

1959-09-15T23:59:59.000Z

33

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

34

Power Ramp Testing of Additive Fuel Rods in the Halden Reactor  

Science Conference Proceedings (OSTI)

Additive fuel provides an effective remedy against pellet-cladding interaction (PCI) failure and a means of mitigating severe secondary degradation in the form of long axial splits in the cladding after fuel failure. This report—part of EPRI's Fuel Reliability Program—summarizes five years worth of data on additive fuel properties, pre-irradiation corrosion and creep tests, microstructure characterization, and subsequent in-core power ramp tests performed on segments retrieved after two cycle...

2003-11-14T23:59:59.000Z

35

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

36

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel slug for a reactor which acts as a safety device is described. The fuel slug is an aluminum tube with a foil lining the inside surface of the tube, the foil being fabricated of uranium in a lead matrix.

Horning, W.A.; Lanning, D.D.; Donahue, D.J.

1959-10-01T23:59:59.000Z

37

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

38

A disposition strategy for highly enriched, aluminum-based fuel from research and test reactors  

SciTech Connect

The strategy proposed in this paper offers the Department of Energy an approach for disposing of aluminum-based, highly enriched uranium (HEU) spent fuels from foreign and domestic research reactors. The proposal is technically, socially, and economically sound. If implemented, it would advance US non-proliferation goals while also disposing of the spent fuel`s waste by timely and proven methods using existing technologies and facilities at SRS without prolonged and controversial storage of the spent fuel. The fuel would be processed through 221-H. The radioactive fission products (waste) would be treated along with existing SRS high level waste by vitrifying it as borosilicate glass in the Defense Waste Processing Facility (DWPF) for disposal in the national geological repository. The HEU would be isotopically diluted, during processing, to low-enriched uranium (LEU) which can not be used to make weapons, thus eliminating proliferation concerns. The LEU can be sold to fabricators of either research reactor fuel or commercial power fuel. This proposed processing-LEU recycle approach has several important advantages over other alternatives, including: Lowest capital investment; lowest net total cost; quickest route to acceptable waste form and final geologic disposal; and likely lowest safety, health, and environmental impacts.

McKibben, J.M.; Gould, T.H.; McDonell, W.R.; Bickford, W.E.

1994-11-01T23:59:59.000Z

39

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

40

Cermet fuel reactors  

Science Conference Proceedings (OSTI)

Cermet fueled nuclear reactors are attractive candidates for high performance space power systems. The cermet fuel consists of tungsten-urania hexagonal fuel blocks characterized by high strength at elevated temperatures, a high thermal conductivity and resultant high thermal shock resistance. Key features of the cermet fueled reactor design are (1) the ability to achieve very high coolant exit temperatures, and (2) thermal shock resistance during rapid power changes, and (3) two barriers to fission product release - the cermet matrix and the fuel element cladding. Additionally, thre is a potential for achieving a long operating life because of (1) the neutronic insensitivity of the fast-spectrum core to the buildup of fission products and (2) the utilization of a high strength refractory metal matrix and structural materials. These materials also provide resistance against compression forces that potentially might compact and/or reconfigure the core. In addition, the neutronic properties of the refractory materials assure that the reactor remains substantially subcritical under conditions of water immersion. It is concluded that cermet fueled reactors can be utilized to meet the power requirements for a broad range of advanced space applications. 4 refs., 4 figs., 3 tabs.

Cowan, C.L.; Palmer, R.S.; Van Hoomissen, J.E.; Bhattacharyya, S.K.; Barner, J.O.

1987-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

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

42

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A reactor fuel element of the capillary tube type is described. The element consists of a thin walled tube, sealed at both ends, and having an interior coatlng of a fissionable material, such as uranium enriched in U-235. The tube wall is gas tight and is constructed of titanium, zirconium, or molybdenum.

Kesselring, K.A.; Seybolt, A.U.

1958-12-01T23:59:59.000Z

43

NEUTRONIC REACTOR FUEL PUMP  

DOE Patents (OSTI)

A reactor fuel pump is described which offers long life, low susceptibility to radiation damage, and gaseous fission product removal. An inert-gas lubricated bearing supports a journal on one end of the drive shsft. The other end has an impeller and expansion chamber which effect pumping and gas- liquid separation. (T.R.H.)

Cobb, W.G.

1959-06-01T23:59:59.000Z

44

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

45

Results of recent reactor-material tests on dispersal of oxide fuel from a disrupted core  

Science Conference Proceedings (OSTI)

The results of experimental investigations and related analyses are reported addressing the dispersal of molten oxide fuel from a disrupted core via various available pathways for the CRBR system. These investigations included the GAPFLOW tests in which pressure-driven and gravity drainage tests were performed using dispersal pathways mocking up the intersubassembly gaps, the CAMEL C6 and C7 tests in which molten fuel entered sodium-filled control assembly ducts under prototypic thermal-hydraulic conditions, and the Lower Internals Drainage (LID) tests in which molten fuel drained downward through simulated below-core structure (orifice plate stacks) as the bottom of control assembly ducts. The results of SHOTGUN tests addressing basic freezing of molten UO/sub 2/ and UO/sub 2//metal mixtures flowing through circular tubes are also reported. Test results have invariably shown the existance of stable UO/sub 2/ crusts on the inside surfaces of the flow paths. Appreciable removal of fuel was indicated prior to freezing-induced immobilization. Application of heat transfer models based upon the presence of stable, insulating fuel crusts tends to overpredict the removal process.

Spencer, B.W.; Wilson, R.J.; Vetter, D.L.; Erickson, E.G.; Dewey, G.

1985-01-01T23:59:59.000Z

46

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

47

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

48

Fossil fuel furnace reactor  

DOE Patents (OSTI)

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

49

Water reactor fuel cladding  

Science Conference Proceedings (OSTI)

This patent describes a nuclear reactor fuel element cladding tube. It comprises: an outer cylindrical layer of a first zirconium alloy selected from the group consisting of Zircaloy-2 and Zircaloy-4; an inner cylindrical layer of a second zirconium alloy consisting essentially of about 0.19 to 0.6 wt.% tin, about 0.19 to less than 0.5 wt.% iron, about 100 to 700 ppm oxygen, less than 2000 ppm total impurities, and the remainder essentially zirconium; the inner layer characterized by aqueous corrosion resistance substantially the same as the first zirconium alloy; the inner layer characterized by improved resistance to PCI crack propagation under reactor operating conditions compared to the first zirconium alloy and substantially the same PCI crack propagation resistance compared to unalloyed zirconium; and the inner cylindrical layer is metallurgically bonded to the outer layer.

Foster, J.P.; McDonald, S.G.

1990-06-12T23:59:59.000Z

50

NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT  

DOE Patents (OSTI)

A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

Currier, E.L. Jr.; Nicklas, J.H.

1962-08-14T23:59:59.000Z

51

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

SciTech Connect

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

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

1987-10-01T23:59:59.000Z

52

Analysis of the Reactor Physics of Low-Enrichment Fuel for the INL Advanced Test Reactor in support of RERTR  

Science Conference Proceedings (OSTI)

Analysis of the performance of the ATR with a LEU fuel design shows promise in terms of a core design that will yield the same neutron sources in target locations. A proposed integral cladding burnable absorber design appears to meet power profile requirements that will satisfy power distributions for safety limits. Performance of this fuel design is ongoing; the current work is the initial evaluation of the core performance of this fuel design with increasing burnup. Results show that LEU fuel may have a longer lifetime that HEU fuel however, such limits may be set by mechanical performance of the fuel rather that available reactivity. Changes seen in the radial fuel power distribution with burnup in LEU fuel will require further study to ascertain the impact on neutron fluxes in target locations. Source terms for discharged fuel have also been studied. By its very nature, LEU fuel produces much more plutonium than is present in HEU fuel at discharge. However, the effect of the plutonium inventory appears to have little affect on radiotoxicity or decay heat in the fuel.

Mark DeHart; William Skerjanc; Sean Morrell

2012-06-01T23:59:59.000Z

53

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

SciTech Connect

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

Timothy A. Hyde

2012-06-01T23:59:59.000Z

54

Technique for preparation of prototypic fast test reactor mixed-oxide fuel pins for ceramography and metallography  

Science Conference Proceedings (OSTI)

Irradiated liquid metal fast breeder reactor prototypic mixed-oxide fuel and cladding have been sectioned and examined in remote hot cell facilities for ceramographic and metallographic examinations. Metallography, shielded electron microprobe examinations, electron microscopy from replicas and autoradiography are routinely employed to obtain postirradiation data for statistical evaluation. Selection of preparatory techniques are based on fuel type, burnup, customer requirements and fuel pin condition.

Chastain, S.A.

1981-01-01T23:59:59.000Z

55

Fuel Reformation: Microchannel Reactor Design  

DOE Green Energy (OSTI)

Fuel processing is used to extract hydrogen from conventional vehicle fuel and allow fuel cell powered vehicles to use the existing petroleum fuel infrastructure. Kilowatt scale micro-channel steam reforming, water-gas shift and preferential oxida-tion reactors have been developed capable of achieving DOE required system performance metrics. Use of a microchannel design effectively supplies heat to the highly endothermic steam reforming reactor to maintain high conversions, controls the temperature profile for the exothermic water gas shift reactor, which optimizes the overall reaction conversion, and removes heat to prevent the unwanted hydrogen oxidation in the prefer-ential oxidation reactor. The reactors combined with micro-channel heat exchangers, when scaled to a full sized 50 kWe automotive system, will be less than 21 L in volume and 52 kg in weight.

Brooks, Kriston P.; Davis, James M.; Fischer, Christopher M.; King, David L.; Pederson, Larry R.; Rawlings, Gregg C.; Stenkamp, Victoria S.; TeGrotenhuis, Ward E.; Wegeng, Robert S.; Whyatt, Greg A.

2005-09-01T23:59:59.000Z

56

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

57

DESTRUCTIVE EXAMINATION OF 3-CYCLE LWR (LIGHT WATER REACTOR) FUEL RODS FROM TURKEY POINT UNIT 3 FOR THE CLIMAX - SPENT FUEL TEST  

DOE Green Energy (OSTI)

The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reator fuel rods with similar burnups (28 GWd/MTU) and operating histories.

ATKIN SD

1981-06-01T23:59:59.000Z

58

Gas Test Loop Booster Fuel Hydraulic Testing  

SciTech Connect

The Gas Test Loop (GTL) project is for the design of an adaptation to the Advanced Test Reactor (ATR) to create a fast-flux test space where fuels and materials for advanced reactor concepts can undergo irradiation testing. Incident to that design, it was found necessary to make use of special booster fuel to enhance the neutron flux in the reactor lobe in which the Gas Test Loop will be installed. Because the booster fuel is of a different composition and configuration from standard ATR fuel, it is necessary to qualify the booster fuel for use in the ATR. Part of that qualification is the determination that required thermal hydraulic criteria will be met under routine operation and under selected accident scenarios. The Hydraulic Testing task in the GTL project facilitates that determination by measuring flow coefficients (pressure drops) over various regions of the booster fuel over a range of primary coolant flow rates. A high-fidelity model of the NW lobe of the ATR with associated flow baffle, in-pile-tube, and below-core flow channels was designed, constructed and located in the Idaho State University Thermal Fluids Laboratory. A circulation loop was designed and constructed by the university to provide reactor-relevant water flow rates to the test system. Models of the four booster fuel elements required for GTL operation were fabricated from aluminum (no uranium or means of heating) and placed in the flow channel. One of these was instrumented with Pitot tubes to measure flow velocities in the channels between the three booster fuel plates and between the innermost and outermost plates and the side walls of the flow annulus. Flow coefficients in the range of 4 to 6.5 were determined from the measurements made for the upper and middle parts of the booster fuel elements. The flow coefficient for the lower end of the booster fuel and the sub-core flow channel was lower at 2.3.

Gas Test Loop Hydraulic Testing Staff

2006-09-01T23:59:59.000Z

59

MOLTEN FLUORIDE NUCLEAR REACTOR FUEL  

DOE Patents (OSTI)

Molten-salt reactor fuel compositions consisting of mixtures of fluoride salts are reported. In its broadest form, the composition contains an alkali fluoride such as sodium fluoride, zirconium tetrafluoride, and a uranium fluoride, the latter being the tetrafluoride or trifluoride or a mixture of the two. An outstanding property of these fuel compositions is a high coeffieient of thermal expansion which provides a negative temperature coefficient of reactivity in reactors in which they are used.

Barton, C.J.; Grimes, W.R.

1960-01-01T23:59:59.000Z

60

Out-of-Reactor Corrosion Tests of Fuel Cladding Materials: Corrosion as a Function of Hydrogen Overpressure  

Science Conference Proceedings (OSTI)

EPRI has sponsored laboratory experiments to investigate whether an increased dissolved hydrogen (DH) level in the reactor coolant of pressurized water reactors (PWR) would result in increased hydrogen pickup (HPU) by the fuel cladding and spacer weld structure materials. This report documents exposure of clean, modern zirconium-based alloys for up to 730 days at three DH levels as well as exposure of Zircaloy 4 (Zry-4) specimens with different types of nickel contacts for 100 days at three DH ...

2013-11-27T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Fuel Management of Reactors Other Than Light Water Reactors

Ehud Greenspan; Pavel Hejzlar; Hiroshi Sekimoto; Georgy Toshinsky; David Wade

62

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

Science Conference Proceedings (OSTI)

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

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

2011-06-07T23:59:59.000Z

63

FUSED REACTOR FUELS  

DOE Patents (OSTI)

This invention relates to a nuciear reactor fuel composition comprising (1) from about 0.01 to about 50 wt.% based on the total weight of said composition of at least one element selected from the class consisting of uranium, thorium, and plutonium, wherein said eiement is present in the form of at least one component selected from the class consisting of oxides, halides, and salts of oxygenated anions, with components comprising (2) at least one member selected from the class consisting of (a) sulfur, wherein the sulfur is in the form of at least one entity selected irom the class consisting of oxides of sulfur, metal sulfates, metal sulfites, metal halosulfonates, and acids of sulfur, (b) halogen, wherein said halogen is in the form of at least one compound selected from the class of metal halides, metal halosulfonates, and metal halophosphates, (c) phosphorus, wherein said phosphorus is in the form of at least one constituent selected from the class consisting of oxides of phosphorus, metal phosphates, metal phosphites, and metal halophosphates, (d) at least one oxide of a member selected from the class consisting of a metal and a metalloid wherein said oxide is free from an oxide of said element in (1); wherein the amount of at least one member selected from the class consisting of halogen and sulfur is at least about one at.% based on the amount of the sum of said sulfur, halogen, and phosphorus atom in said composition; and wherein the amount of said 2(a), 2(b) and 2(c) components in said composition which are free from said elements of uranium, thorium, arid plutonium, is at least about 60 wt.% based on the combined weight of the components of said composition which are free from said elements of uranium, thorium, and plutonium. (AEC)

Mayer, S.W.

1962-11-13T23:59:59.000Z

64

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A fuel element particularly adapted for use in nuclear reactors of high power density is offered. It has fissionable fuel pellet segments mounted in a tubular housing and defining a central passage in the fuel element. A burnable poison element extends through the central passage, which is designed to contain more poison material at the median portion than at the end portions thereby providing a more uniform hurnup and longer reactivity life.

Bassett, C.H.

1961-05-16T23:59:59.000Z

65

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A fuel element is designed which is particularly adapted for reactors of high power density used to generate steam for the production of electricity. The fuel element consists of inner and outer concentric tubes forming an annular chamber within which is contained fissionable fuel pellet segments, wedge members interposed between the fuel segments, and a spring which, acting with wedge members, urges said fuel pellets radially into contact against the inner surface of the outer tube. The wedge members may be a fertile material convertible into fissionable fuel material by absorbing neutrons emitted from the fissionable fuel pellet segments. The costly grinding of cylindrical fuel pellets to close tolerances for snug engagement is reduced because the need to finish the exact size is eliminated. (AEC)

Bassett, C.H.

1961-11-21T23:59:59.000Z

66

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A nuclear reactor fuel element comprising high density ceramic fissionable material enclosed in a tubular cladding of corrosion-resistant material is described. The fissionable material is in the form of segments of a tube which have cooperating tapered interfaces which produce outward radial displacement when the segments are urged axially together. A resilient means is provided within the tubular housing to constantly urge the fuel segments axially. This design maintains the fuel material in tight contacting engagement against the inner surface of the outer cladding tube to eliminate any gap therebetween which may be caused by differential thermal expansion between the fuel material and the material of the tube.

Bassett, C.H.

1961-05-01T23:59:59.000Z

67

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A fuel element possessing good stability and heat conducting properties is described. The fuel element comprises an outer tube formed of material selected from the group consisting of stainhess steel, V, Ti. Mo. or Zr, a fuel tube concentrically fitting within the outer tube and containing an oxide of an isotope selected from the group consisting of U/sup 235/, U/sup 233/, and Pu/sup 239/, and a hollow, porous core concentrically fitting within the fuel tube and formed of an oxide of an element selected from the group consisting of Mg, Be, and Zr.

Shackleford, M.H.

1958-12-16T23:59:59.000Z

68

NEUTRONIC REACTOR FUEL ELEMENT  

DOE Patents (OSTI)

A nuclear fuel element comprising a plurality of nuclear fuel bearing strips is presented. The strips are folded along their longitudinal axes to an angle of about 60 deg and are secured at each end by ferrule to form an elongated assembly suitable for occupying a cylindrical coolant channel.

Gurinsky, D.H.; Powell, R.W.; Fox, M.

1959-11-24T23:59:59.000Z

69

Advanced Gas Reactor Fuel Program's TRISO Particle Fuel Sets A New World  

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

Advanced Gas Reactor Fuel Program's TRISO Particle Fuel Sets A New Advanced Gas Reactor Fuel Program's TRISO Particle Fuel Sets A New World Record For Irradiation Performance Advanced Gas Reactor Fuel Program's TRISO Particle Fuel Sets A New World Record For Irradiation Performance November 16, 2009 - 1:12pm Addthis As part of the Office of Nuclear Energy's Next Generation Nuclear Plant (NGNP) Program, the Advanced Gas Reactor (AGR) Fuel Development Program has achieved a new international record for irradiation testing of next-generation particle fuel for use in high temperature gas reactors (HTGRs). The AGR Fuel Development Program was initiated by the Department of Energy in 2002 to develop the advanced fabrication and characterization technologies, and provide irradiation and safety performance data required to license TRISO particle fuel for the NGNP and future HTGRs. The AGR

70

FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A method is described whereby fuel tubes or pins are cut, loaded with fuel pellets and a heat transfer medium, sealed at each end with slotted fittings, and assembled into a rectangular tube bundle to form a fuel element. The tubes comprising the fuel element are laterally connected between their ends by clips and tabs to form a linear group of spaced parallel tubes, which receive their vertical support by resting on a grid. The advantages of this method are that it permits elimination of structural material (e.g., fuel-element cans) within the reactor core, and removal of at least one fuel pin from an element and replacement thereof so that a burnable poison may be utilized during the core lifetime. (AEC)

Dickson, J.J.

1963-09-24T23:59:59.000Z

71

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

Science Conference Proceedings (OSTI)

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

Dawn Scates

2010-10-01T23:59:59.000Z

72

Nuclear reactor composite fuel assembly  

DOE Patents (OSTI)

A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

1980-01-01T23:59:59.000Z

73

Summary report on the HFED (High-Uranium-Loaded Fuel Element Development) miniplate irradiations for the RERTR (Reduced Enrichment Research and Test Reactor) Program  

SciTech Connect

An experiment to evaluate the irradiation characteristics of various candidate low-enriched, high-uranium content fuels for research and test reactors was performed for the US Department of Energy Reduced Enrichment Research and Test Reactor Program. The experiment included the irradiation of 244 miniature fuel plates (miniplates) in a core position in the Oak Ridge Research Reactor. The miniplates were aluminum-based, dispersion-type plates 114.3 mm long by 50.8 mm wide with overall plate thicknesses of 1.27 or 1.52 mm. Fuel core dimensions varied according to the overall plate thicknesses with a minimum clad thickness of 0.20 mm. Tested fuels included UAl/sub x/, UAl/sub 2/, U/sub 3/O/sub 8/, U/sub 3/SiAl, U/sub 3/Si, U/sub 3/Si/sub 1.5/, U/sub 3/Si/sub 2/, U/sub 3/SiCu, USi, U/sub 6/Fe, and U/sub 6/Mn/sub 1.3/ materials. Although most miniplates were made with low-enriched uranium (19.9%), some with medium-enriched uranium (40 to 45%), a few with high-enriched uranium (93%), and a few with depleted uranium (0.2 to 0.4%) were tested for comparison. These fuel materials were irradiated to burnups ranging from /approximately/27 to 98 at. % /sup 235/U depletion. Operation of the experiment, measurement of miniplate thickness as the irradiation progressed, ultimate shipment of the irradiated miniplates to various hot cells, and preliminary results are reported here. 18 refs., 12 figs., 7 tabs.

Senn, R.L.

1989-04-01T23:59:59.000Z

74

RERTR program activities related to the development and application of new LEU fuels. [Reduced Enrichment Research and Test Reactor; low-enriched uranium  

SciTech Connect

The statue of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of RERTR Program objectives and goals, program accomplishments are discussed with emphasis on the development, demonstration and application of new LEU fuels. Most program activities have proceeded as planned, and a combination of two silicide fuels (U/sub 3/Si/sub 2/-Al and U/sub 3/Si-Al) holds excellent promise for achieving the long-term program goals. Current plans and schedules project the uranium density of qualified RERTR fuels for plate-type reactors to grow by approximately 1 g U/cm/sup 3/ each year, from the current 1.7 g U/cm/sup 3/ to the 7.0 g U/cm/sup 3/ which will be reached in late 1988. The technical needs of research and test reactors for HEU exports are also forecasted to undergo a gradual but dramatic decline in the coming years.

Travelli, A.

1983-01-01T23:59:59.000Z

75

NUCLEAR REACTOR FUEL ELEMENT ASSEMBLY  

DOE Patents (OSTI)

A method of fabricating nuclear reactor fuel element assemblies having a plurality of longitudinally extending flat fuel elements in spaced parallel relation to each other to form channels is presented. One side of a flat side plate is held contiguous to the ends of the elements and a welding means is passed along the other side of the platertransverse to the direction of the longitudinal extension of the elements. The setting and speed of travel of the welding means is set to cause penetration of the side plate with welds at bridge the gap in each channel between adjacent fuel elements with a weld-through bubble of predetermined size. The fabrication of a high strength, dependable fuel element is provided, and the reduction of distortion and high production costs are facilitated by this method. (AEC)

Stengel, F.G.

1963-12-24T23:59:59.000Z

76

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

77

Fuel handling apparatus for a nuclear reactor  

DOE Patents (OSTI)

Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

Hawke, Basil C. (Solana Beach, CA)

1987-01-01T23:59:59.000Z

78

Preliminary Study on Utilization of Carbon Dioxide as a Coolant of High Temperature Engineering Test Reactor with MOX and Minor Actinides Fuel  

Science Conference Proceedings (OSTI)

High temperature engineering test reactor (HTTR) is an uranium oxide (UO2) fuel, graphite moderator and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 950 deg. C. Instead of using helium gas, we have utilized carbon dioxide as a coolant in the present study. Beside that, uranium and plutonium oxide (mixed oxide, MOX) and minor actinides have been employed as a new fuel type of HTTR. Utilization of plutonium and minor actinide is one of the support system to non-proliferation issue in the nuclear development. The enrichment for uranium oxide has been varied of 6-20% with plutonium and minor actinides concentration of 10%. In this study, burnup period is 1100 days. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. Reactor core calculation was done by using CITATION module. The result shows that HTTR can achieve its criticality condition with 14% of {sup 235}U enrichment.

Fauzia, A. F.; Waris, A.; Novitrian [Bosscha Laboratory, Department of Physics, Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, INDONESIA Jl. Ganesa 10 Bandung 40132 (Indonesia)

2010-06-22T23:59:59.000Z

79

Fast Reactor Fuel Type and Reactor Safety Performance  

Science Conference Proceedings (OSTI)

Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

R. Wigeland; J. Cahalan

2009-09-01T23:59:59.000Z

80

Proliferation Resistant Nuclear Reactor Fuel  

Science Conference Proceedings (OSTI)

Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and we posit that the exploration, development, and implementation of intrinsic mechanisms such as discussed here are part of a balanced approach aimed at preventing the misuse of nuclear material for nuclear-energy applications.

Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

2011-02-18T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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.


81

Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories  

Office of Legacy Management (LM)

ADVANCED REACTORS DIVISION FUEL LABORATORIES CHESWICK, PENNSYLVANIA Department of Energy Office of Policy, Safety and Environment Office of Operational Safety Environmental...

82

Accident Tolerant Fuels for Light Water Reactors  

Science Conference Proceedings (OSTI)

Presentation Title, Accident Tolerant Fuels for Light Water Reactors. Author(s), Steven J. Zinkle, Kurt A. Terrani, Lance L. Snead. On-Site Speaker (Planned) ...

83

Rethinking the light water reactor fuel cycle  

E-Print Network (OSTI)

The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to ...

Shwageraus, Evgeni, 1973-

2004-01-01T23:59:59.000Z

84

Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2  

Science Conference Proceedings (OSTI)

The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

2002-09-01T23:59:59.000Z

85

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

86

DECONTAMINATION OF NEUTRON-IRRADIATED REACTOR FUEL  

DOE Patents (OSTI)

A pyrometallurgical method of decontaminating neutronirradiated reactor fuel is presented. In accordance with the invention, neutron-irradiated reactor fuel may be decontaminated by countercurrently contacting the fuel with a bed of alkali and alkaine fluorides under an inert gas atmosphere and inductively melting the fuel and tracking the resulting descending molten fuel with induction heating as it passes through the bed. By this method, a large, continually fresh surface of salt is exposed to the descending molten fuel which enhances the efficiency of the scrubbing operation.

Buyers, A.G.; Rosen, F.D.; Motta, E.E.

1959-12-22T23:59:59.000Z

87

Irradiation and Testing of Fuels and Cladding Materials  

Science Conference Proceedings (OSTI)

Mar 14, 2012 ... Mechanical Performance of Materials for Current and Advanced Nuclear Reactors: Irradiation and Testing of Fuels and Cladding Materials

88

United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support  

SciTech Connect

The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

Douglas Morrell

2011-03-01T23:59:59.000Z

89

Simulated nuclear reactor fuel assembly  

DOE Patents (OSTI)

An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

Berta, Victor T. (Idaho Falls, ID)

1993-01-01T23:59:59.000Z

90

The DOE Advanced Gas Reactor Fuel Development and Qualification Program  

Science Conference Proceedings (OSTI)

The high outlet temperatures and high thermal-energy conversion efficiency of modular High Temperature Gas-cooled Reactors (HTGRs) enable an efficient and cost effective integration of the reactor system with non-electricity generation applications, such as process heat and/or hydrogen production, for the many petrochemical and other industrial processes that require temperatures between 300°C and 900°C. The Department of Energy (DOE) has selected the HTGR concept for the Next Generation Nuclear Plant (NGNP) Project as a transformative application of nuclear energy that will demonstrate emissions-free nuclear-assisted electricity, process heat, and hydrogen production, thereby reducing greenhouse-gas emissions and enhancing energy security. The objective of the DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification program is to qualify tristructural isotropic (TRISO)-coated particle fuel for use in HTGRs. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, post-irradiation examination (PIE) and safety testing, fuel performance modeling, and fission-product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete, fundamental understanding of the relationship between the fuel fabrication process and key fuel properties, the irradiation and accident safety performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. An overview of the program and recent progress is presented.

David Petti

2010-09-01T23:59:59.000Z

91

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

92

FUEL PROGRAMMING FOR SODIUM GRAPHITE REACTORS  

SciTech Connect

The effect of fuel programming, i.e., the scheme used for changing fuel in a core, on the reactivity and specific power of a sodium graphite reactor is discussed Fuel programs considered Include replacing fuel a core-load at a time or a radial zone at a time, replacing fuel to manutain the same average exposure of fuel elements throughout the core, and replacing and transferring fuel elements to maintain more highly exposed fuel in the center or at the periphery of the core. Flux and criticality calculations show the degree of power flattening and the concurrent decrease in effective multiplication which results from maintaining more exposed fuel toward the core center. Corverse effects are shown for the case of maintaining more exposed fuel near the core periphery. The excess reactivity which must be controlled in the various programs is considered. Illustrative schedules for implementing each of these programs in an SGR are presented. (auth)

Connolly, T.J.

1959-10-15T23:59:59.000Z

93

NEUTRONIC REACTOR FUEL ELEMENT AND CORE SYSTEM  

DOE Patents (OSTI)

This patent relates to neutronic reactors and in particular to an improved fuel element and a novel reactor core system for facilitating removal of contaminating fission products, as they are fermed, from association with the flssionable fuel, so as to mitigate the interferent effects of such fission products during reactor operation. The fuel elements are comprised of tubular members impervious to fluid and contatning on their interior surfaces a thin layer of fissionable material providing a central void. The core structure is comprised of a plurality of the tubular fuel elements arranged in parallel and a closed manifold connected to their ends. In the reactor the core structure is dispersed in a water moderator and coolant within a pressure vessel, and a means connected to said manifuld is provided for withdrawing and disposing of mobile fission product contamination from the interior of the feel tubes and manifold.

Moore, W.T.

1958-09-01T23:59:59.000Z

94

Overview of reduced enrichment fuels: Development, testing, and specification  

SciTech Connect

The US Reduced Enrichment Research and Test Reactor (RERTR) Program was established in 1978 to provide the technical means to operate research and test reactors with low enrichment uranium (LEU) fuels without significant penalty in experiment performance, operation costs, component modifications, or safety characteristics. This paper discusses relevant developments in fuel developments. 9 refs., 1 tab.

Snelgrove, J.L.

1987-01-01T23:59:59.000Z

95

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

E-Print Network (OSTI)

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

Fray, Elliott Shepard

2013-01-01T23:59:59.000Z

96

Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.  

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 in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

2009-03-01T23:59:59.000Z

97

Hawaii Fuel Cell Test Facility  

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

Fuel Cell Test Facility presented to DOE Hydrogen Codes and Standards Coordinating Committee Fuel Purity Specifications Workshop Renaissance Hollywood Hotel by Rick Rocheleau...

98

Fuel cycle problems in fusion reactors  

SciTech Connect

Fuel cycle problems of fusion reactors evolve around the breeding, recovery, containment, and recycling of tritium. These processes are described, and their implications and alternatives are discussed. Technically, fuel cycle problems are solvable; economically, their feasibility is not yet known. (auth)

Hickman, R.G.

1976-01-13T23:59:59.000Z

99

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...

100

Future Transient Testing of Advanced Fuels  

Science Conference Proceedings (OSTI)

The transient in-reactor fuels testing workshop was held on May 4–5, 2009 at Idaho National Laboratory. The purpose of this meeting was to provide a forum where technical experts in transient testing of nuclear fuels could meet directly with technical instrumentation experts and nuclear fuel modeling and simulation experts to discuss needed advancements in transient testing to support a basic understanding of nuclear fuel behavior under off-normal conditions. The workshop was attended by representatives from Commissariat à l'Énergie Atomique CEA, Japanese Atomic Energy Agency (JAEA), Department of Energy (DOE), AREVA, General Electric – Global Nuclear Fuels (GE-GNF), Westinghouse, Electric Power Research Institute (EPRI), universities, and several DOE national laboratories. Transient testing of fuels and materials generates information required for advanced fuels in future nuclear power plants. Future nuclear power plants will rely heavily on advanced computer modeling and simulation that describes fuel behavior under off-normal conditions. TREAT is an ideal facility for this testing because of its flexibility, proven operation and material condition. The opportunity exists to develop advanced instrumentation and data collection that can support modeling and simulation needs much better than was possible in the past. In order to take advantage of these opportunities, test programs must be carefully designed to yield basic information to support modeling before conducting integral performance tests. An early start of TREAT and operation at low power would provide significant dividends in training, development of instrumentation, and checkout of reactor systems. Early start of TREAT (2015) is needed to support the requirements of potential users of TREAT and include the testing of full length fuel irradiated in the FFTF reactor. The capabilities provided by TREAT are needed for the development of nuclear power and the following benefits will be realized by the refurbishment and restart of TREAT. •TREAT is an absolute necessity in the suite of reactor fuel test capabilities •TREAT yields valuable information on reactivity effects, margins to failure, fuel dispersal, and failure propagation •Most importantly, interpretation of TREAT experiment results is a stringent test of the integrated understanding of fuel performance.

Jon Carmack

2009-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

A Study of Fast Reactor Fuel Transmutation in a Candidate Dispersion Fuel Design  

SciTech Connect

Dispersion fuels represent a significant departure from typical ceramic fuels to address swelling and radiation damage in high burnup fuel. Such fuels use a manufacturing process in which fuel particles are encapsulated within a non-fuel matrix. Dispersion fuels have been studied since 1997 as part of an international effort to develop and test very high density fuel types for the Reduced Enrichment for Research and Test Reactors (RERTR) program.[1] The Idaho National Laboratory is performing research in the development of an innovative dispersion fuel concept that will meet the challenges of transuranic (TRU) transmutation by providing an integral fission gas plenum within the fuel itself, to eliminate the swelling that accompanies the irradiation of TRU. In this process, a metal TRU vector produced in a separations process is atomized into solid microspheres. The dispersion fuel process overcoats the microspheres with a mixture of resin and hollow carbon microspheres to create a TRUC. The foam may then be heated and mixed with a metal power (e.g., Zr, Ti, or Si) and resin to form a matrix metal carbide, that may be compacted and extruded into fuel elements. In this paper, we perform reactor physics calculations for a core loaded with the conceptual fuel design. We will assume a “typical” TRU vector and a reference matrix density. We will employ a fuel and core design based on the Advanced Burner Test Reactor (ABTR) design.[2] Using the CSAS6 and TRITON modules of the SCALE system [3] for preliminary scoping studies, we will demonstrate the feasibility of reactor operations. This paper will describe the results of these analyses.

Mark DeHart; Hongbin Zhang; Eric Shaber; Matthew Jesse

2010-11-01T23:59:59.000Z

102

Spent fuel test project, Climax granitic stock, Nevada Test Site  

SciTech Connect

The Spent Fuel Test-Climax (SFT-C) is a test of dry geologic storage of spent nuclear reactor fuel. The SFT-C is located at a depth of 420 m in the Climax granitic stock at the Nevada Test Site. Eleven canisters of spent commercial PWR fuel assemblies are to be stored for 3 to 5 years. Additional heat is supplied by electrical heaters, and more than 800 channels of technical information are being recorded. The measurements include rock temperature, rock displacement and stress, joint motion, and monitoring of the ventilation air volume, temperature, and dewpoint.

Ramspott, L.D.

1980-10-24T23:59:59.000Z

103

EVALUATION OF CORE PHYSICS ANALYSIS METHODS FOR CONVERSION OF THE INL ADVANCED TEST REACTOR TO LOW-ENRICHMENT FUEL  

Science Conference Proceedings (OSTI)

Computational neutronics studies to support the possible conversion of the ATR to LEU are underway. Simultaneously, INL is engaged in a physics methods upgrade project to put into place modern computational neutronics tools for future support of ATR fuel cycle and experiment analysis. A number of experimental measurements have been performed in the ATRC in support of the methods upgrade project, and are being used to validate the new core physics methods. The current computational neutronics work is focused on performance of scoping calculations for the ATR core loaded with a candidate LEU fuel design. This will serve as independent confirmation of analyses that have been performed previously, and will evaluate some of the new computational methods for analysis of a candidate LEU fuel for ATR.

Mark DeHart; Gray S. Chang

2012-04-01T23:59:59.000Z

104

Evaluation of core physics analysis methods for conversion of the INL advanced test reactor to low-enrichment fuel  

Science Conference Proceedings (OSTI)

Computational neutronics studies to support the possible conversion of the ATR to LEU are underway. Simultaneously, INL is engaged in a physics methods upgrade project to put into place modern computational neutronics tools for future support of ATR fuel cycle and experiment analysis. A number of experimental measurements have been performed in the ATRC in support of the methods upgrade project, and are being used to validate the new core physics methods. The current computational neutronics work is focused on performance of scoping calculations for the ATR core loaded with a candidate LEU fuel design. This will serve as independent confirmation of analyses that have been performed previously, and will evaluate some of the new computational methods for analysis of a candidate LEU fuel for ATR. (authors)

DeHart, M. D.; Chang, G. S. [Idaho National Laboratory, 2525 Fremont Street, Idaho Falls, ID 83415-3870 (United States)

2012-07-01T23:59:59.000Z

105

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

106

Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide  

Science Conference Proceedings (OSTI)

This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors’ spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

2012-09-01T23:59:59.000Z

107

Fuel assembly for nuclear reactors  

DOE Patents (OSTI)

A new and improved fuel assembly is formed to minimize the amount of parasitic structural material wherein a plurality of hollow tubular members are juxtaposed to the fuel elements of the assembly. The tubular members may serve as guide tubes for control elements and are secured to a number of longitudinally spaced grid members along the fuel assembly. The grid members include means thereon engaging each of the fuel elements to laterally position the fuel elements in a predetermined array. Openings in the bottom of each hollow member serve as a shock absorber to cushion shock transmitted to the structure when the control elements are rapidly inserted in their corresponding tubular members.

Creagan, Robert J. (Pitcairn, PA); Frisch, Erling (Pittsburgh, PA)

1977-01-01T23:59:59.000Z

108

TESTING AND ACCEPTANCE OF FUEL PLATES FOR RERTR FUEL DEVELOPMENT EXPERIMENTS  

SciTech Connect

This paper discusses how candidate fuel plates for RERTR Fuel Development experiments are examined and tested for acceptance prior to reactor insertion. These tests include destructive and nondestructive examinations (DE and NDE). The DE includes blister annealing for dispersion fuel plates, bend testing of adjacent cladding, and microscopic examination of archive fuel plates. The NDE includes Ultrasonic (UT) scanning and radiography. UT tests include an ultrasonic scan for areas of “debonds” and a high frequency ultrasonic scan to determine the "minimum cladding" over the fuel. Radiography inspections include identifying fuel outside of the maximum fuel zone and measurements and calculations for fuel density. Details of each test are provided and acceptance criteria are defined. These tests help to provide a high level of confidence the fuel plate will perform in the reactor without a breach in the cladding.

J.M. Wight; G.A. Moore; S.C. Taylor

2008-10-01T23:59:59.000Z

109

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

110

Current Projects for Reactor Physics and Fuel Cycle Analysis...  

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

Nuclear Systems Modeling and Design Analysis > Reactor Physics and Fuel Cycle Analysis > Current Projects Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics...

111

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

112

Remote Inspection Devices for Spent Reactor Enriched Uranium Fuel Elements  

SciTech Connect

A remote video inspection was developed and deployed in Argentina for the detailed inspection of highly radioactive spent reactor fuel (SNF) as a prerequisite to its shipment to the Savannah River Site (SRS) in the United States for long-term storage and disposition. The fuel is highly enriched uranium (HEU) spent assemblies dating from 1967 to 1989 and aluminum clad uranium-aluminum alloy of a typical material test reactor design. The specialized video system was designed for low cost, high portability, easy setup, and ease of usage, while accommodating the differing electrical systems (i.e. 110/60 Hz, 220/50 Hz) between the United States and Argentina.

Heckendorn, F.M.

2001-01-03T23:59:59.000Z

113

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

114

Strengthening the nuclear-reactor fuel cycle against proliferation  

SciTech Connect

Argonne National Laboratory (ANL) conducts several research programs that serve to reduce the risks of fissile-material diversion from the nuclear-reactor fuel cycle. The objectives are to provide economical and efficient neutron or power generation with the minimum of inherent risks, and to further minimize risks by utilizing sophisticated techniques to detect attempts at material diversion. This paper will discuss the Reduced Enrichment Research and Test Reactor (RERTR) Program, the Isotope Correlation Technique (ICT), and Proliferation-Resistant Closed-Cycle Reactors. The first two are sponsored by the DOE Office of Arms Control and Nonproliferation.

Travelli, A.; Snelgrove, J.; Persiani, P. [Argonne National Lab., IL (United States). Arms Control and Nonproliferation Program

1992-12-31T23:59:59.000Z

115

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.

116

AN ANALYSIS OF POWER REACTOR FUEL REPROCESSING  

SciTech Connect

This report presents an analysis of the projected economies and processing capacity requirements for a power reactor fuel reprocessing industry based on the recovery of fertile and fissionable materials from presently proposed power reactors within tbe confines of the continental United 8tates for the next five to ten years. An analysis of the present general state of development of a technology required for such an Industry is given. A summary of results of power reactor reprocessing chemical and engineering development at Oak Ridge National Laboratory from July 1955 through December 1956 is given. (auth)

Culler, F.L. Jr.; Blanco, R.E.; Goeller, H.E.; Watson, C.D.

1957-03-27T23:59:59.000Z

117

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

SciTech Connect

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

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

1983-03-01T23:59:59.000Z

118

Optimally moderated nuclear fission reactor and fuel source therefor  

DOE Patents (OSTI)

An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

Ougouag, Abderrafi M. (Idaho Falls, ID); Terry, William K. (Shelley, ID); Gougar, Hans D. (Idaho Falls, ID)

2008-07-22T23:59:59.000Z

119

Electrorefining {open_quotes}N{close_quotes} reactor fuel  

SciTech Connect

Principles of purifying of uranium metal by electrorefining are reviewed. Metal reactor fuel after irradiation is a form of impure uranium. Dissolution and deposition electrorefining processes were developed for spent metal fuel under the Integral Fast Reactor Program. Application of these processes to the conditioning of spent N-reactor fuel slugs is examined.

Gay, E.C.; Miller, W.E.

1995-02-01T23:59:59.000Z

120

PLUTONIUM METALLIC FUELS FOR FAST REACTORS  

Science Conference Proceedings (OSTI)

Early interest in metallic plutonium fuels for fast reactors led to much research on plutonium alloy systems including binary solid solutions with the addition of aluminum, gallium, or zirconium and low-melting eutectic alloys with iron and nickel or cobalt. There was also interest in ternaries of these elements with plutonium and cerium. The solid solution and eutectic alloys have most unusual properties, including negative thermal expansion in some solid-solution alloys and the highest viscosity known for liquid metals in the Pu-Fe system. Although metallic fuels have many potential advantages over ceramic fuels, the early attempts were unsuccessful because these fuels suffered from high swelling rates during burn up and high smearing densities. The liquid metal fuels experienced excessive corrosion. Subsequent work on higher-melting U-PuZr metallic fuels was much more promising. In light of the recent rebirth of interest in fast reactors, we review some of the key properties of the early fuels and discuss the challenges presented by the ternary alloys.

STAN, MARIUS [Los Alamos National Laboratory; HECKER, SIEGFRIED S. [Los Alamos National Laboratory

2007-02-07T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

Natural Fueling of a Tokamak Fusion Reactor  

E-Print Network (OSTI)

A natural fueling mechanism that helps to maintain the main core deuterium and tritium (DT) density profiles in a tokamak fusion reactor is discussed. In H-mode plasmas dominated by ion- temperature gradient (ITG) driven turbulence, cold DT ions near the edge will naturally pinch radially inward towards the core. This mechanism is due to the quasi-neutral heat flux dominated nature of ITG turbulence and still applies when trapped and passing kinetic electron effects are included. Fueling using shallow pellet injection or supersonic gas jets is augmented by an inward pinch of could DT fuel. The natural fueling mechanism is demonstrated using the three-dimensional toroidal electromagnetic gyrokinetic turbulence code GEM and is analyzed using quasilinear theory. Profiles similar to those used for conservative ITER transport modeling that have a completely flat density profile are examined and it is found that natural fueling actually reduces the linear growth rates and energy transport.

Wan, Weigang; Chen, Yang; Perkins, Francis W

2009-01-01T23:59:59.000Z

122

NEUTRONIC REACTOR AND FUEL ELEMENT THEREFOR  

DOE Patents (OSTI)

This patent relates to a reactor design of the type which employs solid fuel elements disposed in channels within the moderator through which channels and around the fuel elements is conveyed a coolant fiuid. The coolant channels are comprised of aluminum tubes extending through a solid moderator such as graphite and the fuel elements are comprised of an elongated solid body of natural uranium jacketed in an aluminum jacket with the ends thereof closed by aluminum caps of substantially greater thickness than the jacket was and in good thermal contact with the fuel material to facilitate the conduction of heat from the central portion of said ends to the coolant surrounding the fuel element to prevent overheating of said central portion.

Szilard, L.; Young, G.J.

1958-03-01T23:59:59.000Z

123

Nuclear reactor core and fuel element therefor  

SciTech Connect

This patent describes a nuclear reactor core. This core consists of vertical columns of disengageable fuel elements stacked one atop another. These columns are arranged in side-by-side relationship to form a substantially continuous horizontal array. Each of the fuel elements include a block of refractory material having relatively good thermal conductivity and neutron moderating characteristics. The block has a pair of parallel flat top and bottom end faces and sides which are substantially prependicular to the end faces. The sides of each block is aligned vertically within a vertical column, with the sides of vertically adjacent blocks. Each of the blocks contains fuel chambers, including outer rows containing only fuel chambers along the sides of the block have nuclear fuel material disposed in them. The blocks also contain vertical coolant holes which are located inside the fuel chambers in the outer rows and the fuel chambers which are not located in the outer rows with the fuel chambers and which extend axially completely through from end face to end face and form continuous vertical intracolumn coolant passageways in the reactor core. The blocks have vertical grooves extending along the sides of the blocks form interblock channels which align in groups to form continuous vertical intercolumn coolant passsageways in the reactor core. The blocks are in the form of a regular hexagonal prism with each side of the block having vertical gooves defining one half of one of the coolant interblock channels, six corner edges on the blocks have vertical groves defining one-third of an interblock channel, the vertical sides of the blocks defining planar vertical surfaces.

Fortescue, P.

1986-02-11T23:59:59.000Z

124

PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL  

DOE Patents (OSTI)

A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

Buyers, A.G.

1959-06-30T23:59:59.000Z

125

FUEL ELEMENT FOR NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear fuel element comprising a large number og wafers of fissionable material and a protective jacket having compartments holding these wafers is described. The compartments of the jacket aid the removal of heat from the wafers, keep the wafers or fragments thereof from migrating in the jacket, and permit the escape of gaseous fission products.

Carney, K.G. Jr.

1959-07-14T23:59:59.000Z

126

Welding austenitic steel clads for fast reactor fuel pins  

SciTech Connect

ABS>From symposium on fuel and elements for fast reactors; Brussels. Belgium (2 Jul 1973). Developmental programs aimed at fabrication of stainless steelclad PuO/sub 2/ fuel pins are described. Information and data are included on welding fast reactor fuel cans, methods of reducing the incidence of weld cracking, effects of weld stresses, and fuel plug design. (JRD)

Papeleux, P.; Flipot, A.J.; Lafontaine, I.

1973-01-01T23:59:59.000Z

127

Fusion reactors for synthetic fuels  

DOE Green Energy (OSTI)

Some of the types of processes now being considered for synthetic fuels production from fusion energy, together with an example of each type are listed. The process efficiency is defined as the chemical energy in the generated hydrogen (at the higher heating value (HHV)) divided by the total fusion energy release, including alpha particles and secondary neutron reactions in the blanket. Except where specifically noted, both high and low temperature blanket heats are counted as part of total fusion energy release.

Powell, J.R.

1979-01-01T23:59:59.000Z

128

Light water reactor mixed-oxide fuel irradiation experiment  

SciTech Connect

The United States Department of Energy Office of Fissile Materials Disposition is sponsoring and Oak Ridge National Laboratory (ORNL) is leading an irradiation experiment to test mixed uranium-plutonium oxide (MOX) fuel made from weapons-grade (WG) plutonium. In this multiyear program, sealed capsules containing MOX fuel pellets fabricated at Los Alamos National Laboratory (LANL) are being irradiated in the Advanced Test Reactor (ATR) at the Idaho National Engineering and Environmental Laboratory (INEEL). The planned experiments will investigate the utilization of dry-processed plutonium, the effects of WG plutonium isotopics on MOX performance, and any material interactions of gallium with Zircaloy cladding.

Hodge, S.A.; Cowell, B.S. [Oak Ridge National Lab., TN (United States); Chang, G.S.; Ryskamp, J.M. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.

1998-06-01T23:59:59.000Z

129

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

130

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

131

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.

132

Alternative Fuels Data Center: Fuel Economy Test Procedures and Labeling  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Fuel Economy Test Fuel Economy Test Procedures and Labeling to someone by E-mail Share Alternative Fuels Data Center: Fuel Economy Test Procedures and Labeling on Facebook Tweet about Alternative Fuels Data Center: Fuel Economy Test Procedures and Labeling on Twitter Bookmark Alternative Fuels Data Center: Fuel Economy Test Procedures and Labeling on Google Bookmark Alternative Fuels Data Center: Fuel Economy Test Procedures and Labeling on Delicious Rank Alternative Fuels Data Center: Fuel Economy Test Procedures and Labeling on Digg Find More places to share Alternative Fuels Data Center: Fuel Economy Test Procedures and Labeling on AddThis.com... More in this section... Federal State Advanced Search All Laws & Incentives Sorted by Type Fuel Economy Test Procedures and Labeling

133

Fuel Rod Cooling in Natural Uranium Reactors  

SciTech Connect

An analysis is presented of the transfer of heat from a cylindrical fuel rod surrounded by a fast flowing coolant in an annular duct, with maximum power output limited by fuel rod temperatures, coolant pressure drop and pumping power requirements. A method is also presented for comparing and evaluating various liquid and gaseous coolants within these limitations. The report also shows and discusses some calculated results obtained for the systems considred in the study of natural U reactors for the production of Pu and useful power (NAA-SR-137).

Trilling, C.A.

1952-01-28T23:59:59.000Z

134

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

A fuel element structure particularly useful in high temperature nuclear reactors is presented. Basically, the structure comprises two coaxial graphite sleeves integrally joined together by radial fins. Due to the high structural strength of graphite at high temperatures and the rigidity of this structure, nuclear fuel encased within the inner sleeve in contiguous relation therewith is supported and prevented from expanding radially at high temperatures. Thus, the necessity of relying on the usual cladding materials with relatively low temperature limitations for structural strength is removed. (AEC)

Davidson, J.K.

1963-11-19T23:59:59.000Z

135

Fuel handling system for a nuclear reactor  

DOE Patents (OSTI)

A pool type nuclear fission reactor has a core, with a plurality of core elements and a redan which confines coolant as a hot pool at a first end of the core separated from a cold pool at a second end of the core by the redan. A fuel handling system for use with such reactors comprises a core element storage basket located outside of the redan in the cold pool. An access passage is formed in the redan with a gate for opening and closing the passage to maintain the temperature differential between the hot pool and the cold pool. A mechanism is provided for opening and closing the gate. A lifting arm is also provided for manipulating the fuel core elements through the access passage between the storage basket and the core when the redan gate is open.

Saiveau, James G. (Hickory Hills, IL); Kann, William J. (Park Ridge, IL); Burelbach, James P. (Glen Ellyn, IL)

1986-01-01T23:59:59.000Z

136

NUCLEAR REACTOR AND THERMIONIC FUEL ELEMENT THEREFOR  

DOE Patents (OSTI)

The patent relates to the direct conversion of fission heat to electricity by use of thermionic plasma diodes having fissionable material cathodes, said diodes arranged to form a critical mass in a nuclear reactor. The patent describes a fuel element comprising a plurality of diodes each having a fissionable material cathode, an anode around said cathode, and an ionizable gas therebetween. Provision is made for flowing the gas and current serially through the diodes. (AEC)

Rasor, N.S.; Hirsch, R.L.

1963-12-01T23:59:59.000Z

137

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...

138

Post Irradiation Evaluation of BWR Fuel From Hope Creek Reactor  

Science Conference Proceedings (OSTI)

Occasionally, in some BWRs, fuel pellet washout from a single degraded fuel rod has resulted in high offgas levels that were sufficient to impede the reactor operation. In addition, certain sound fuel rods have exhibited high eddy-current liftoff values during routine poolside measurements. Investigators pursued these two recent BWR fuel issues by performing detailed hotcell examinations on selected fuel rods from the Hope Creek reactor. The results provided insights into the mechanisms involved and poss...

1997-03-12T23:59:59.000Z

139

PEBBLE-BED NUCLEAR REACTOR SYSTEM PHYSICS AND FUEL UTILIZATION  

E-Print Network (OSTI)

The Generation IV Pebble Bed Modular Reactor (PMBR) design may be used for electricity production, co-generation applications (industrial heat, hydrogen production, desalination, etc.), and could potentially eliminate some high level nuclear wastes. Because of these advantages, as well as the ability to build cost-effective small-to-medium sized reactors, this design is currently being considered for construction in many countries, from Japan, where test reactors are being analyzed, to China. The use of TRISO-coated micro-particles as a fuel in these reactors leads to multi-heterogeneity physics features that must be properly treated and accounted for. Inherent interrelationships of neutron interactions, temperature effects, and structural effects, further challenge computational evaluations of High Temperature Reactors (HTRs). The developed models and computational techniques have to be validated in code-to-code and, most importantly, code-to-experiment benchmark studies. This report quantifies the relative accuracy of various multi-heterogeneity treatments in whole-core 3D models for parametric studies of Generation IV Pebble Bed Modular Reactors as well as provide preliminary results of the PBMR performance analysis. Data is gathered from two different models, one based upon a benchmark for the African PBMR-400 design, and another based on the PROTEUS criticality experiment, since the African design is a more realistic power reactor, but the PROTEUS experiment model can be used for calculations that cannot be performed on the more complex model. Early data was used to refine final models, and the resulting final models were used to conduct parametric studies on composition and geometry optimization based on pebble bed reactor physics in order to improve fuel utilization.

Kelly, Ryan 1989-

2011-05-01T23:59:59.000Z

140

Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)  

SciTech Connect

The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

Jon Carmack (062056); Kemal O. Pasamehmetoglu (103171); David Alberstein

2008-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

Spent nuclear fuel discharges from U.S. reactors 1994  

Science Conference Proceedings (OSTI)

Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

NONE

1996-02-01T23:59:59.000Z

142

Updated Uranium Fuel Cycle Environmental Impacts for Advanced Reactor Designs  

Science Conference Proceedings (OSTI)

The purpose of this project was to update the environmental impacts from the uranium fuel cycle for select advanced (GEN III+) reactor designs.

Nitschke, R.

2004-10-03T23:59:59.000Z

143

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

144

Fuel assembly transfer basket for pool type nuclear reactor vessels  

DOE Patents (OSTI)

A fuel assembly transfer basket for a pool type, liquid metal cooled nuclear reactor having a side access loading and unloading port for receiving and relinquishing fuel assemblies during transfer.

Fanning, Alan W. (San Jose, CA); Ramsour, Nicholas L. (San Jose, CA)

1991-01-01T23:59:59.000Z

145

Preparations for the Integral Fast Reactor fuel cycle demonstration  

Science Conference Proceedings (OSTI)

Modifications to the Hot Fuel Examination Facility-South (HFEF/S) have been in progress since mid-1988 to ready the facility for demonstration of the unique Integral Fast Reactor (IFR) pyroprocess fuel cycle. This paper updates the last report on this subject to the American Nuclear Society and describes the progress made in the modifications to the facility and in fabrication of the new process equipment. The IFR is a breeder reactor, which is central to the capability of any reactor concept to contribute to mitigation of environmental impacts of fossil fuel combustion. As a fast breeder, fuel of course must be recycled in order to have any chance of an economical fuel cycle. The pyroprocess fuel cycle, relying on a metal alloy reactor fuel rather than oxide, has the potential to be economical even at small-scale deployment. Establishing this quantitatively is one important goal of the IFR fuel cycle demonstration.

Lineberry, M.J.; Phipps, R.D.

1989-01-01T23:59:59.000Z

146

Optimization of hydride fueled pressurized water reactor cores  

E-Print Network (OSTI)

This thesis contributes to the Hydride Fuels Project, a collaborative effort between UC Berkeley and MIT aimed at investigating the potential benefits of hydride fuel use in light water reactors (LWRs). This pursuit involves ...

Shuffler, Carter Alexander

2004-01-01T23:59:59.000Z

147

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

148

Nuclear reactor fuel rod attachment system  

DOE Patents (OSTI)

A reusable system for removably attaching a nuclear reactor fuel rod (12) to a support member (14). A locking cap (22) is secured to the fuel rod (12) and a locking strip (24) is fastened to the support member (14). The locking cap (22) has two opposing fingers (24a and 24b) shaped to form a socket having a body portion (26). The locking strip has an extension (36) shaped to rigidly attach to the socket's body portion (26). The locking cap's fingers are resiliently deflectable. For attachment, the locking cap (22) is longitudinally pushed onto the locking strip (24) causing the extension (36) to temporarily deflect open the fingers (24a and 24b) to engage the socket's body portion (26). For removal, the process is reversed.

Christiansen, David W. (Kennewick, WA)

1982-01-01T23:59:59.000Z

149

FUEL ASSEMBLY SHAKER TEST SIMULATION  

SciTech Connect

This report describes the modeling of a PWR fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that a can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL’s proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During FY 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL’s expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) are very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable. The SNL test data offers some opportunity for validation of the finite element model, but the specific loading conditions of the testing only excite the fuel assembly to respond in a limited manner. For example, the test accelerations were not strong enough to substantially drive the fuel assembly out of contact with the basket. Under this test scenario, the fuel assembly model does a reasonable job of approximating actual fuel assembly response, a claim that can be verified through direct comparison of model results to recorded test results. This does not offer validation for the fuel assembly model in all conceivable cases, such as high kinetic energy shock cases where the fuel assembly might lift off the basket floor to strike to basket ceiling. This type of nonlinear behavior was not witnessed in testing, so the model does not have test data to be validated against.a basis for validation in cases that substantially alter the fuel assembly response range. This leads to a gap in knowledge that is identified through this modeling study. The SNL shaker testing loaded a surrogate fuel assembly with a certain set of artificially-generated time histories. One thing all the shock cases had in common was an elimination of low frequency components, which reduces the rigid body dynamic response of the system. It is not known if the SNL test cases effectively bound all highway transportation scenarios, or if significantly greater rigid body motion than was tested is credible. This knowledge gap could be filled through modeling the vehicle dynamics of a used fuel conveyance, or by collecting acceleration time history data from an actual conveyance under highway conditions.

Klymyshyn, Nicholas A.; Sanborn, Scott E.; Adkins, Harold E.; Hanson, Brady D.

2013-05-30T23:59:59.000Z

150

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

151

Fuel performance comparison between Savannah River reactors and the US commercial nuclear reactors  

SciTech Connect

This document provides a review of fuel/target performance of the Savannah River Reactors which was made to compare their in-core performance with that of the commercial nuclear reactors in the US.

Paik, I.K.; Ellison, P.G.

1989-01-01T23:59:59.000Z

152

Fuel Summary Report: Shippingport Light Water Breeder Reactor  

SciTech Connect

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01T23:59:59.000Z

153

CHEMICAL ASPECTS OF PELLET-CLADDING INTERACTION IN LIGHT WATER REACTOR FUEL ELEMENTS  

E-Print Network (OSTI)

ANS/ENS Topical Meeting on Reactor Safety Aspects of FuelINTERACTION IN LiaiT-WATER-REACTOR FUEL ELEMENTS by D. R.PCI) in light water reactor fuel elements, the chemical

Olander, D.R.

2010-01-01T23:59:59.000Z

154

Thermionic Fuel Element performance: TFE Verification Program. Final test report  

DOE Green Energy (OSTI)

The program objective is to demonstrate the technology readiness of a Thermionic Fuel Element (TFE) suitable for use as the basic element in a thermionic reactor with electric power output in the 0.5 to 5.0 MW(e) range, and a full power life of 7 years. A TFE was designed that met the reliability and lifetime requirements for a 2 MW(e) conceptual reactor design. Analysis showed that this TFE could be used over the range of 0.5 to 5 megawatts. This was used as the basis for designing components for test and evaluation. The demonstration of a 7-year component lifetime capability was through the combined use of analytical models and accelerated, confirmatory tests in a fast test reactor. Iterative testing was performed in which the results of one test series led to evolutionary improvements in the next test specimens. The TFE components underwent screening and initial development testing in ex-reactor tests. Several design and materials options were considered for each component. As screening tests permitted, down selection occurred to very specific designs and materials. In parallel with ex-reactor testing, and fast reactor component testing, components were integrated into a TFE and tested in the TRIGA test reactor at GA. Realtime testing of partial length TFEs was used to test support, alignment and interconnective TFE components, and to verify TFE performance in-reactor with integral cesium reservoirs. Realtime testing was also used to verify the relation between TFE performance and fueled emitter swelling, to test the durability of intercell insulation, to check temperature distributions, and to verify the adequacy over time of the fission gas venting channels. Predictions of TFE lifetime rested primarily on the accelerated component testing results, as correlated and extended to realtime by the use of analytical models.

Not Available

1994-06-01T23:59:59.000Z

155

PROCESSING OF MOLTEN SALT POWER REACTOR FUEL  

SciTech Connect

ABS> Fuel reprocessing methods are being investigated for molten salt nuclear reactors which use LiF--BeF/sub 2/ salt as a solvent for UF/sub 4/ and ThF/sub 4/. A liquid HF dissolution procedure coupled with fluorination has been developed for recovery of the uranium and LiF- BeF/sub 2/ solvent salt which is highly enriched in Li/sup 7/. The recovered salt is decontaminated in the process from the major reactor poisons; namely, rare earths and neptunium. A brief investigation of alternate methods, including oxide precipitation, partial freezing, and metal reduction, indicated that such methods may give some separation of the solvent salt from reactor poisons, but they do not appear to be sufficiently quantitative for a simple processing operation. Solubilities of LiF and BeF/sub 2/ in aqueous 70t0 100% HF are presented. The BeF/sub 2/ solubility is appreciably increased in the presence of water and large amounts of LiF. Salt solubilities of 150 g/liter are attainable. Tracer experiments indicate that rare earth solubilities, relative to LiF-- BeF/sub 2/ solvent salt solubility, increase from about 10/sup -4/ mole% in 98% HF to 0.003 mole% in 80% HF. Fluorination of uranium from LiF--BeF/sub 2/ salt was demonstrated. This appears feasible also for the recovery of the relatively small ccncentration of uranium produced in the LiF- BeF/sub 2/ThF/sub 4/ blanket. A proposed chemical flowsheet is presented on the basis of this exploratory work as applied to the semicontinuous processing of a 600 Mw power reactor. (auth)

Campbell, D.O.; Cathers, G.I.

1959-04-01T23:59:59.000Z

156

Distribution of characteristics of LWR [light water reactor] spent fuel  

SciTech Connect

The purpose of this report is to develop a collective description of the entire spent fuel inventory in terms of various fuel properties relevant to Approved Testing Materials (ATMs) using information available from the Characteristics Data Base (CBD), which is sponsored by the US Department of Energy`s (DOE`s) Office of Civilian Radioactive Waste Management. A number of light-water reactor (LWR) characteristics were analyzed including assembly class representation, fuel burnup, enrichment, fuel fabrication data, defective fuel quantities, and, at PNL`s specific request, linear heat generation rate (LHGR) and the utilization of burnable poisons. A quantitative relationships was developed between burnup and enrichment for BWRs and PWRs. The relationship shows that the existing BWR ATM is near the center of the burnup-enrichment distribution, while the four PWR ATMs bracket the center of the burnup range but are on the low side of the enrichment range. Fuel fabrication data are based on vendor specifications for new fuel. Defective fuel distributions were analyzed in terms of assembly class and vendor design. LHGR values were calculated from utility data on burnup and effective full-power days; these calculations incorporate some unavoidable assumptions which may compromise the value of the results. Only a limited amount of data are available on burnable poisons at this time. Based on this distribution study, suggestions for additional ATMs are made. These are based on the class and design concepts and include BWR/2,3 barrier fuel, and the WE 17 {times} 17 class with integral burnable poison. Both should be at relatively high burnups. 16 refs., 5 figs., 15 tabs.

Reich, W.J.; Notz, K.J. [Oak Ridge National Lab., TN (USA); Moore, R.S. [Automated Sciences Group, Inc., Oak Ridge, TN (USA)

1991-01-01T23:59:59.000Z

157

PREPARING THE HIGH FLUX ISOTOPE REACTOR FOR CONVERSION TO LOW ENRICHED URANIUM FUEL ? EXTENDING CYCLE BURNUP  

Science Conference Proceedings (OSTI)

Reactor performance studies have been completed for conceptual plate designs and show that maintaining reactor performance while converting HFIR from high enriched to low enriched uranium (20 wt % 235U) fuel requires extending the end-of-life burnup value for HFIR fuel from the current nominal value of 2200 MWD to 2600 MWD. The current fuel fabrication procedure is discussed and changes that would be required to this procedure are identified. Design and safety related analyses that are required for the certification of a new fuel are identified. Qualification tests and comments regarding the regulatory approval process are provided along with a conceptual schedule.

Primm, Trent [ORNL; Chandler, David [ORNL

2009-01-01T23:59:59.000Z

158

FLOWSHEET EVALUATION FOR THE DISSOLVING AND NEUTRALIZATION OF SODIUM REACTOR EXPERIMENT USED NUCLEAR FUEL  

Science Conference Proceedings (OSTI)

This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

Daniel, W. E.; Hansen, E. K.; Shehee, T. C.

2012-10-30T23:59:59.000Z

159

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

160

Fuel cycle design and analysis of SABR: subrcritical advanced burner reactor.  

E-Print Network (OSTI)

??Various fuel cycles for a sodium-cooled, subcritical, fast reactor with a fusion neutron source for the transmutation of light water reactor spent fuel have been… (more)

Sommer, Christopher

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

Experience with non-fuel-bearing components in LWR (light-water reactor) fuel systems  

SciTech Connect

Many non-fuel-bearing components are so closely associated with the spent fuel assemblies that their integrity and behavior must be taken into consideration with the fuel assemblies, when handling spent fuel of planning waste management activities. Presented herein is some of the experience that has been gained over the past two decades from non-fuel-bearing components in light-water reactors (LWRs), both pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). Among the most important of these components are the control rod systems, the absorber and burnable poison rods, and the fuel assembly channels. 15 refs., 5 figs., 2 tabs.

Bailey, W.J.; Berting, F.M.

1990-12-01T23:59:59.000Z

162

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

A lattice type fissionable fuel structure for a nuclear reactor is described. The fissionable material is formed into a plurality of rod-llke bodies with each encased in a fluid-tight jacket. A plurality of spaced longitudinal fins are mounted on the exterior and extend radially from each jacket, with a portion of the fins extending radially beyond the remainder of the fins. A collar of short length for each body is mounted on the extended fins for spacing the bodies, and adjacent bodies abut each other through these collars. Should distortion of the bodies take place, coilapse of the outer fins is limited by the shorter flns, thereby insuring some coolant flow at all times. (AEC)

Duffy, J.G. Jr.

1961-05-30T23:59:59.000Z

163

FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

A lattice-type fissionable fuel structure for a nuclear reactor is offered. The fissionable material is formed into a plurality of rod-like bodies each encased in a fluid-tight jacket. A plurality of spaced longitudinal fins are mounted on the exterior of and extend radially from each jacket, and a portion of the fins extends radially beyond the remainder of the fins. A collar of short lengih for each body is mounted on the extended fins for spacing the bodies, and adjacent bodies abut each other through these collars. Should distortion of the bodies take place, collapse of the outer fins is limited by the shorter fins thereby insuring some coolant flow therethrough at all times.

Duffy, J.G. Jr.

1961-05-30T23:59:59.000Z

164

Preparation of high temperature gas-cooled reactor fuel element  

DOE Patents (OSTI)

This invention relates to a method for the preparation of high temperature gas-cooled reactor (HTGR) fuel elements wherein uncarbonized fuel rods are inserted in appropriate channels of an HTGR fuel element block and the entire block is inserted in an autoclave for in situ carbonization under high pressure. The method is particularly applicable to remote handling techniques.

Bradley, Ronnie A. (Oak Ridge, TN); Sease, John D. (Knoxville, TN)

1976-01-01T23:59:59.000Z

165

Conversion and standardization of university reactor fuels using low-enrichment uranium: Plans and schedules  

SciTech Connect

The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. To minimize this risk, the US Nuclear Regulatory Commission issued its final rule on ''Limiting the Use of Highly Enriched Uranium in Domestically Licensed Research and Test Reactors,'' in February 1986. This paper describes the plans and schedules developed by the US Department of Energy to coordinate an orderly transition from HEU to LEU fuel in most of these reactors. An important element in the planning process has been the desire to standardize the LEU fuels used in US university reactors and to enhance the performance and utilization of a number of these reactors. The program is estimated to cost about $10 million and to last about five years.

Young, H.H.; Brown, K.R.; Matos, J.E.

1986-01-01T23:59:59.000Z

166

PROTEUS - Simulation Toolset for Reactor Physics and Fuel Cycle Analysis  

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

Simulation Toolset for Simulation Toolset for Reactor Physics and Fuel Cycle Analysis PROTEUS Faster and more accurate neutronics calculations enable optimum reactor design... Argonne National Laboratory's powerful reactor physics toolset, PROTEUS, empowers users to create optimal reactor designs quickly, reliably and accurately. ...Reducing costs for designers of fast spectrum reactors. PROTEUS' long history of validation provides confidence in predictive simulations Argonne's simulation tools have more than 30 years of validation history against numerous experiments and measurements. The tools within PROTEUS work together, using the same interface files for easier integration of calculations. Multi-group Fast Reactor Cross Section Processing: MC 2 -3 No other fast spectrum multigroup generation tool

167

High Performance Fuel Desing for Next Generation Pressurized Water Reactors  

SciTech Connect

The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

Mujid S. Kazimi; Pavel Hejzlar

2006-01-31T23:59:59.000Z

168

Spent fuel utilization in a compact traveling wave reactor  

SciTech Connect

In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

Hartanto, Donny; Kim, Yonghee [Korea Advanced Institute of Science and Technology 373-1 Kusong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2012-06-06T23:59:59.000Z

169

Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering Division  

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

Analysis Analysis Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Overview Current Projects Software Nuclear Plant Dynamics and Safety Nuclear Data Program Advanced Reactor Development Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Reactor Physics and Fuel Cycle Analysis Bookmark and Share Reactor physics and fuel cycle analysis is a core competency of the Nuclear Engineering (NE) Division. The Division has played a major role in the design and analysis of advanced reactors, particularly liquid-metal-cooled reactors. NE researchers have concentrated on developing computer codes for

170

Hot Fuel Examination Facility's neutron radiography reactor  

SciTech Connect

Argonne National Laboratory-West is located near Idaho Falls, Idaho, and is operated by the University of Chicago for the United States Department of Energy in support of the Liquid Metal Fast Breeder Reactor Program, LMFBR. The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both nondestructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the nondestructive examination techniques utilized at HFEF is neutron radiography, which is provided by the NRAD reactor facility (a TRIGA type reactor) below the HFEF hot cell.

Pruett, D.P.; Richards, W.J.; Heidel, C.C.

1983-01-01T23:59:59.000Z

171

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

172

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

173

Fuel rod retention device for a nuclear reactor  

DOE Patents (OSTI)

A device is described for supporting a nuclear fuel rod in a fuel rod assembly which allows the rod to be removed without disturbing other rods in the assembly. A fuel rod cap connects the rod to a bolt which is supported in the assembly end fitting by means of a locking assembly. The device is designed so that the bolt is held securely during normal reactor operation yet may be easily disengaged and the fuel rod removed when desired.

Hylton, Charles L. (Madison Heights, VA)

1984-01-01T23:59:59.000Z

174

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...

175

A Blueprint for GNEP Advanced Burner Reactor Startup Fuel Fabrication Facility  

SciTech Connect

The purpose of this article is to identify the requirements and issues associated with design of GNEP Advanced Burner Reactor Fuel Facility. The report was prepared in support of providing data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives was to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept was proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR was proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu was assumed to be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) was being considered for fabrication of WG Pu fuel for the ABR. It was estimated that the facility will provide the startup fuel for 10-15 years and would take 3 to 5 years to construct.

S. Khericha

2010-12-01T23:59:59.000Z

176

EXTENDING SODIUM FAST REACTOR DRIVER FUEL USE TO HIGHER TEMPERATURES  

Science Conference Proceedings (OSTI)

Calculations of potential sodium-cooled fast reactor fuel temperatures were performed to estimate the effects of increasing the outlet temperature of a given fast reactor design by increasing pin power, decreasing assembly flow, or increasing inlet temperature. Based upon experience in the U.S., both metal and mixed oxide (MOX) fuel types are discussed in terms of potential performance effects created by the increased operating temperatures. Assembly outlet temperatures of 600, 650 and 700 °C were used as goal temperatures. Fuel/cladding chemical interaction (FCCI) and fuel melting, as well as challenges to the mechanical integrity of the cladding material, were identified as the limiting phenomena. For example, starting with a recent 1000 MWth fast reactor design, raising the outlet temperature to 650 °C through pin power increase increased the MOX centerline temperature to more than 3300 °C and the metal fuel peak cladding temperature to more than 700 °C. These exceeded limitations to fuel performance; fuel melting was limiting for MOX and FCCI for metal fuel. Both could be alleviated by design ‘fixes’, such as using a barrier inside the cladding to minimize FCCI in the metal fuel, or using annular fuel in the case of MOX. Both would also require an advanced cladding material with improved stress rupture properties. While some of these are costly, the benefits of having a high-temperature reactor which can support hydrogen production, or other missions requiring high process heat may make the extra costs justified.

Douglas L. Porter

2011-02-01T23:59:59.000Z

177

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

178

Reactor physics assessment of thick silicon carbide clad PWR fuels  

E-Print Network (OSTI)

High temperature tolerance, chemical stability and low neutron affinity make silicon carbide (SiC) a potential fuel cladding material that may improve the economics and safety of light water reactors (LWRs). "Thick" SiC ...

Bloore, David A. (David Allan)

2013-01-01T23:59:59.000Z

179

Innovative fuel designs for high power density pressurized water reactor  

E-Print Network (OSTI)

One of the ways to lower the cost of nuclear energy is to increase the power density of the reactor core. Features of fuel design that enhance the potential for high power density are derived based on characteristics of ...

Feng, Dandong, Ph. D. Massachusetts Institute of Technology

2006-01-01T23:59:59.000Z

180

IN-PILE GAS-COOLED FUEL ELEMENT TEST FACILITY  

SciTech Connect

Paper presented at American Nuclear Society Meeting, June I8-21, 1962, Boston, Mass. Design and operating problems of unclad and ceramic gas-cooled reactor fuels in high temperature circulating gas systems will be studied using a test facility now nearing completion at the Oak Ridge Research Reactor. A shielded air-tight cell houses a closed circuit gas system equipped for dealing with fission products circulating in the gas. Experiments can be conducted on fuel element performance and stability, fission product deposition, gas clean up, activity levels, component and system performance and shielding, and decontamination and maintenance of system hardware. (auth)

Zasler, J.; Huntley, W.R.; Gnadt, P.A.; Kress, T.S.

1962-07-10T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

Fabrication of advanced oxide fuels containing minor actinide for use in fast reactors  

Science Conference Proceedings (OSTI)

R and D of advanced fuel containing minor actinide for use in fast reactors is described related to the composite fuel with MgO matrix. Fabrication tests of MgO composite fuels containing Am were done by a practical process that could be adapted to the presently used commercial manufacturing technology. Am-containing MgO composite fuels having good characteristics, i.e., having no defects, a high density, a homogeneous dispersion of host phase, were obtained. As related technology, burn-up characteristics of a fast reactor core loaded with the present MgO composite fuel were also analyzed, mainly in terms of core criticality. Furthermore, phase relations of MA oxide which was assumed to be contained in MgO matrix fuel were experimentally investigated. (authors)

Miwa, Shuhei; Osaka, Masahiko; Tanaka, Kosuke; Ishi, Yohei; Yoshimochi, Hiroshi; Tanaka, Kenya [Oarai Research and Development Center, Japan Atomic Energy Agency, 4002 Oarai-machi, Higashi-ibaraki-gun, Ibaraki, 311-1393 (Japan)

2007-07-01T23:59:59.000Z

182

Yttrium and rare earth stabilized fast reactor metal fuel  

DOE Patents (OSTI)

To increase the operating temperature of a reactor, the melting point and mechanical properties of the fuel must be increased. For an actinide-rich fuel, yttrium, lanthanum and/or rare earth elements can be added, as stabilizers, to uranium and plutonium and/or a mixture of other actinides to raise the melting point of the fuel and improve its mechanical properties. Since only about 1% of the actinide fuel may be yttrium, lanthanum, or a rare earth element, the neutron penalty is low, the reactor core size can be reduced, the fuel can be burned efficiently, reprocessing requirements are reduced, and the nuclear waste disposal volumes reduced. A further advantage occurs when yttrium, lanthanum, and/or other rare earth elements are exposed to radiation in a reactor, they produce only short half life radioisotopes, which reduce nuclear waste disposal problems through much shorter assured-isolation requirements.

Guon, Jerold (Woodland Hills, CA); Grantham, LeRoy F. (Calabasas, CA); Specht, Eugene R. (Simi Valley, CA)

1992-01-01T23:59:59.000Z

183

Spent nuclear fuel discharges from US reactors 1993  

SciTech Connect

The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

Not Available

1995-02-01T23:59:59.000Z

184

VENTED FUEL ELEMENT FOR GAS-COOLED NEUTRONIC REACTORS  

DOE Patents (OSTI)

A hollow, porous-walled fuel element filled with fissionable fuel and provided with an outlet port through its wall is described. In operation in a gas-cooled reactor, the element is connected, through its outlet port, to the vacuum side of a pump that causes a portion of the coolant gas flowing over the exterior surface of the element to be drawn through the porous walls thereof and out through the outlet port. This continuous purging gas flow sweeps away gaseous fission products as they are released by the fissioning fuel. (AEC) A fuel element for a nuclear reactor incorporating a body of metal of melting point lower than the temperature of operation of the reactor and a nuclear fuel in finely divided form dispersed in the body of metal as a settled slurry is presented. (AEC)

Furgerson, W.T.

1963-12-17T23:59:59.000Z

185

The DOE Advanced Gas Reactor (AGR) Fuel Development and Qualification Program  

SciTech Connect

The Department of Energy has established the Advanced Gas Reactor Fuel Development and Qualification Program to address the following overall goals: Provide a baseline fuel qualification data set in support of the licensing and operation of the Next Generation Nuclear Plant (NGNP). Gas-reactor fuel performance demonstration and qualification comprise the longest duration research and development (R&D) task for the NGNP feasibility. The baseline fuel form is to be demonstrated and qualified for a peak fuel centerline temperature of 1250°C. Support near-term deployment of an NGNP by reducing market entry risks posed by technical uncertainties associated with fuel production and qualification. Utilize international collaboration mechanisms to extend the value of DOE resources. The Advanced Gas Reactor Fuel Development and Qualification Program consists of five elements: fuel manufacture, fuel and materials irradiations, postirradiation examination (PIE) and safety testing, fuel performance modeling, and fission product transport and source term evaluation. An underlying theme for the fuel development work is the need to develop a more complete fundamental understanding of the relationship between the fuel fabrication process, key fuel properties, the irradiation performance of the fuel, and the release and transport of fission products in the NGNP primary coolant system. Fuel performance modeling and analysis of the fission product behavior in the primary circuit are important aspects of this work. The performance models are considered essential for several reasons, including guidance for the plant designer in establishing the core design and operating limits, and demonstration to the licensing authority that the applicant has a thorough understanding of the in-service behavior of the fuel system. The fission product behavior task will also provide primary source term data needed for licensing. An overview of the program and recent progress will be presented.

David Petti; Hans Gougar; Gary Bell

2005-05-01T23:59:59.000Z

186

Assessment of innovative fuel designs for high performance light water reactors  

E-Print Network (OSTI)

To increase the power density and maximum allowable fuel burnup in light water reactors, new fuel rod designs are investigated. Such fuel is desirable for improving the economic performance light water reactors loaded with ...

Carpenter, David Michael

2006-01-01T23:59:59.000Z

187

Elastomer Compatibility Testing of Renewable Diesel Fuels  

DOE Green Energy (OSTI)

In this study, the integrity and performance of six elastomers were tested with ethanol-diesel and biodiesel fuel blends.

Frame, E.; McCormick, R. L.

2005-11-01T23:59:59.000Z

188

Gel-sphere-pac reactor fuel fabrication and its application to a variety of fuels  

SciTech Connect

The gel-sphere-pac fuel fabrication option was evaluated for its possible application to commercial scale fuel fabrication for 19 fuel element designs that use oxide fuel in metal clad rods. The dry gel spheres are prepared at the reprocessing plant and are then calcined, sintered, inspected, and loaded into fuel rods and packed by low-energy vibration. A fuel smear density of 83 to 88% theoretical can be obtained. All fuel fabrication process steps were defined and evaluated from fuel receiving to finished fuel element shipping. The evaluation also covers the feasibility of the process, the current status of technology, estimates of the required time and cost to develop the technology to commercial status, and the safety and licensability of commercial scale plants. The primary evaluation was for a Light-Water Reactor fuel element containing (U,Pu)O/sub 2/ fuel. The other 18 fuel element types - 3 for Light-Water Reactors, 1 for a Heavy-Water Reactor, 1 for a Gas-Cooled Fast Reactor, 7 for Liquid-Metal-Cooled Fast Breeder Reactors, and 3 pairs for Light-Water Prebreeder and Breeder Reactors - were compared with the Light-Water Reactor. The gel-sphere-pac option was found applicable to 17 of the 19 element types; the characteristics of a commercial scale plant were defined for these for making cost estimates for such plants. The evaluation clearly shows the gel-sphere-pac process to be a viable fuel fabrication option. Estimates indicate a significant potential fabrication cost advantage for the gel-sphere-pac process if a remotely operated and remotely maintained fuel fabrication plant is required.

Olsen, A.R.; Judkins, R.R. (comps.)

1979-12-01T23:59:59.000Z

189

Reference (Axially Graded) Low Enriched Uranium Fuel Design for the High Flux Isotope Reactor (HFIR)  

Science Conference Proceedings (OSTI)

During the past five years, staff at the Oak Ridge National Laboratory (ORNL) have studied the issue of whether the HFIR could be converted to low enriched uranium (LEU) fuel without degrading the performance of the reactor. Using state-of-the-art reactor physics methods and behind-the-state-of-the-art thermal hydraulics methods, the staff have developed fuel plate designs (HFIR uses two types of fuel plates) that are believed to meet physics and thermal hydraulic criteria provided the reactor power is increased from 85 to 100 MW. The paper will present a defense of the results by explaining the design and validation process. A discussion of the requirements for showing applicability of analyses to approval for loading the fuel to HFIR lead test core irradiation currently scheduled for 2016 will be provided. Finally, the potential benefits of upgrading thermal hydraulics methods will be discussed.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

190

PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY  

SciTech Connect

The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.

S. T. Khericha

2007-04-01T23:59:59.000Z

191

NEUTRON REACTOR FUEL ELEMENT UTILIZING ZIRCONIUM-BASE ALLOYS  

DOE Patents (OSTI)

This patent relates to clad fuel elements for use in neutronic reactors and is drawn to such a fuel element which consists of a core of fissionable material, comprised of an alloy of zirconium and U/sup 235/ enriched uranium, encased in a jacket of a binary zirconium-tin alloy in which the tin content ranges between 1 and 15% by weight.

Saller, H.A.; Keeler, J.R.; Szumachowski, E.R.

1957-11-12T23:59:59.000Z

192

Transient Testing of Nuclear Fuels and Materials in United States  

Science Conference Proceedings (OSTI)

The US Department of Energy (DOE) has been engaged in an effort to develop and qualify next generation LWR fuel with enhanced performance and safety and reduced waste generation since 2010. This program, which has emphasized collaboration between the DOE, U.S. national laboratories and nuclear industry, was refocused from enhanced performance to enhanced accident tolerance following the events at Fukushima in 2011. Accident tolerant fuels have been specifically described as fuels that, in comparison with standard UO2-Zircaloy, can tolerate loss of active cooling in the reactor core for a considerably longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, as well as design-basis and beyond design-basis events. The program maintains an ambitious goal to insert a lead test assembly (LTA) of the new design into a commercial power reactor by 2022 .

Daniel M. Wachs

2012-12-01T23:59:59.000Z

193

Nuclear reactor vessel fuel thermal insulating barrier  

DOE Patents (OSTI)

The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

2013-03-19T23:59:59.000Z

194

ORR irradiation experiment OF-1: accelerated testing of HTGR fuel  

SciTech Connect

The OF-1 capsule, the first in a series of High-Temperature Gas-Cooled Reactor fuel irradiations in the Oak Ridge Research Reactor, was irradiated for more than 9300 hr at full reactor power (30 MW). Peak fluences of 1.08 x 10/sup 22/ neutrons/cm/sup 2/ (> 0.18 MeV) were achieved. General Atomic Company's magazine P13Q occupied the upper two-thirds of the test space and the ORNL magazine OF-1 the lower one-third. The ORNL portion tested various HTGR recycle particles and fuel bonding matrices at accelerated flux levels under reference HTGR irradiation conditions of temperature, temperature gradient, and fast fluence exposure (> 0.18 MeV).

Tiegs, T.N.; Long, E.L. Jr.; Kania, M.J.; Thoms, K.R.; Allen, E.J.

1977-08-01T23:59:59.000Z

195

Processing of FRG high-temperature gas-cooled reactor fuel elements at General Atomic under the US/FRG cooperative agreement for spent fuel elements  

Science Conference Proceedings (OSTI)

The Federal Republic of Germany (FRG) and the United States (US) are cooperating on certain aspects of gas-cooled reactor technology under an umbrella agreement. Under the spent fuel treatment development section of the agreement, both FRG mixed uranium/ thorium and low-enriched uranium fuel spheres have been processed in the Department of Energy-sponsored cold pilot plant for high-temperature gas-cooled reactor (HTGR) fuel processing at General Atomic Company in San Diego, California. The FRG fuel spheres were crushed and burned to recover coated fuel particles suitable for further treatment for uranium recovery. Successful completion of the tests described in this paper demonstrated certain modifications to the US HTGR fuel burining process necessary for FRG fuel treatment. Results of the tests will be used in the design of a US/FRG joint prototype headend facility for HTGR fuel.

Holder, N.D.; Strand, J.B.; Schwarz, F.A.; Drake, R.N.

1981-11-01T23:59:59.000Z

196

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

197

Software: Reactor Physics and Fuel Cycle Analysis - Nuclear Engineering  

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

Analysis > Analysis > Software Capabilities Nuclear Systems Modeling and Design Analysis Reactor Physics and Fuel Cycle Analysis Overview Current Projects Software Nuclear Plant Dynamics and Safety Nuclear Data Program Advanced Reactor Development Nuclear Waste Form and Repository Performance Modeling Nuclear Energy Systems Design and Development Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Reactor Physics and Fuel Cycle Analysis Software Bookmark and Share An extensive powerful suite of computer codes developed and validated by the NE Division and its predecessor divisions at Argonne supports the development of fast reactors; many of these codes are also applicable to other reactor types. A brief description of these codes follows. Contact

198

REACTOR PHYSICS MODELING OF SPENT RESEARCH REACTOR FUEL FOR TECHNICAL NUCLEAR FORENSICS  

SciTech Connect

Technical nuclear forensics (TNF) refers to the collection, analysis and evaluation of pre- and post-detonation radiological or nuclear materials, devices, and/or debris. TNF is an integral component, complementing traditional forensics and investigative work, to help enable the attribution of discovered radiological or nuclear material. Research is needed to improve the capabilities of TNF. One research area of interest is determining the isotopic signatures of research reactors. Research reactors are a potential source of both radiological and nuclear material. Research reactors are often the least safeguarded type of reactor; they vary greatly in size, fuel type, enrichment, power, and burn-up. Many research reactors are fueled with highly-enriched uranium (HEU), up to {approx}93% {sup 235}U, which could potentially be used as weapons material. All of them have significant amounts of radiological material with which a radioactive dispersal device (RDD) could be built. Therefore, the ability to attribute if material originated from or was produced in a specific research reactor is an important tool in providing for the security of the United States. Currently there are approximately 237 operating research reactors worldwide, another 12 are in temporary shutdown and 224 research reactors are reported as shut down. Little is currently known about the isotopic signatures of spent research reactor fuel. An effort is underway at Savannah River National Laboratory (SRNL) to analyze spent research reactor fuel to determine these signatures. Computer models, using reactor physics codes, are being compared to the measured analytes in the spent fuel. This allows for improving the reactor physics codes in modeling research reactors for the purpose of nuclear forensics. Currently the Oak Ridge Research reactor (ORR) is being modeled and fuel samples are being analyzed for comparison. Samples of an ORR spent fuel assembly were taken by SRNL for analytical and radiochemical analysis. The fuel assembly was modeled using MONTEBURNS(MCNP5/ ORIGEN2.2) and MCNPX/CINDER90. The results from the models have been compared to each other and to the measured data.

Nichols, T.; Beals, D.; Sternat, M.

2011-07-18T23:59:59.000Z

199

Low-Enriched Fuel Design Concept for the Prismatic Very High Temperature Reactor Core  

SciTech Connect

A new non-TRISO fuel and clad design concept is proposed for the prismatic, heliumcooled Very High Temperature Reactor core. The new concept could substantially reduce the current 10-20 wt% TRISO uranium enrichments down to 4-6 wt% for both initial and reload cores. The proposed fuel form would be a high-temperature, high-density uranium ceramic, for example UO2, configured into very small diameter cylindrical rods. The small diameter fuel rods significantly increase core reactivity through improved neutron moderation and fuel lumping. Although a high-temperature clad system for the concept remains to be developed, recent success in tube fabrication and preliminary irradiation testing of silicon carbide (SiC) cladding for light water reactor applications offers good potential for this application, and for future development of other carbide clad designs. A high-temperature ceramic fuel, together with a high-temperature clad material, could also lead to higher thermal safety margins during both normal and transient reactor conditions relative to TRISO fuel. The calculated neutronic results show that the lowenrichment, small diameter fuel rods and low thermal neutron absorbing clad retain the strong negative Doppler fuel temperature coefficient of reactivity that ensures inherent safe operation of the VHTR, and depletion studies demonstrate that an 18-month power cycle can be achieved with the lower enrichment fuel.

Sterbentz, James W

2007-05-01T23:59:59.000Z

200

A Monte Carlo method of evaluating heterogeneous effects in plate-fueled reactors  

SciTech Connect

Few-group nuclear cross sections for small plate-fueled, light and heavy water test reactors are frequently generated with unit cell models that contain a homogeneous mixture of fuel, cladding, and water. The heterogeneous unit cells do not need to be represented explicitly for neutronics calculations when the plate and coolant channel thicknesses are small compared with the mean-free-path of neutrons. However, neutron and photon heating calculations were performed with heterogeneous fuel models to predict accurately the heat deposited in the fuel meat, cladding, and coolant. Heat deposited in the coolant channels and outside the fuel elements does not have a direct impact on the peak fuel meat temperature but must be included in the total coolant system heat balance. The results of a heterogeneous Monte Carlo calculation that estimates the heat loads in different fuel regions are presented and the fact that similar homogeneous fuel models can be used for many calculations. The calculations presented here were performed on models of the Advanced Neutron Source (ANS) and the Massachusetts Institute of Technology Reactor 2 (MITR-2). The ANS is a small, 362-MW (fission), plate-fueled, heavy water reactor designed to produce an intense steady-state source of neutrons.

Thayer, R.C.; Redmond, E.L. II; Ryskamp, J.M. (Idaho National Engineering Lab., Idaho Falls (United States))

1991-01-01T23:59:59.000Z

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201

LWRS Fuels Pathway: Engineering Design and Fuels Pathway Initial Testing of the Hot Water Corrosion System  

Science Conference Proceedings (OSTI)

The Advanced LWR Nuclear Fuel Development R&D pathway performs strategic research focused on cladding designs leading to improved reactor core economics and safety margins. The research performed is to demonstrate the nuclear fuel technology advancements while satisfying safety and regulatory limits. These goals are met through rigorous testing and analysis. The nuclear fuel technology developed will assist in moving existing nuclear fuel technology to an improved level that would not be practical by industry acting independently. Strategic mission goals are to improve the scientific knowledge basis for understanding and predicting fundamental nuclear fuel and cladding performance in nuclear power plants, and to apply this information in the development of high-performance, high burn-up fuels. These will result in improved safety, cladding, integrity, and nuclear fuel cycle economics. To achieve these goals various methods for non-irradiated characterization testing of advanced cladding systems are needed. One such new test system is the Hot Water Corrosion System (HWCS) designed to develop new data for cladding performance assessment and material behavior under simulated off-normal reactor conditions. The HWCS is capable of exposing prototype rodlets to heated, high velocity water at elevated pressure for long periods of time (days, weeks, months). Water chemistry (dissolved oxygen, conductivity and pH) is continuously monitored. In addition, internal rodlet heaters inserted into cladding tubes are used to evaluate repeated thermal stressing and heat transfer characteristics of the prototype rodlets. In summary, the HWCS provides rapid ex-reactor evaluation of cladding designs in normal (flowing hot water) and off-normal (induced cladding stress), enabling engineering and manufacturing improvements to cladding designs before initiation of the more expensive and time consuming in-reactor irradiation testing.

Dr. John Garnier; Dr. Kevin McHugh

2012-09-01T23:59:59.000Z

202

Characterization of spent fuel approved testing material---ATM-105  

Science Conference Proceedings (OSTI)

The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1991-12-01T23:59:59.000Z

203

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

204

Novel Reactor Design for Solid Fuel Chemical Looping Combustion  

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

Novel Reactor Design for Solid Fuel Novel Reactor Design for Solid Fuel Chemical Looping Combustion Opportunity Research is active on the patent pending technology, titled "Apparatus and Method for Solid Fuel Chemical Looping Combustion." This technology is available for licensing and/or further collaborative research from the U.S. Department of Energy's National Energy Technology Laboratory. Overview The removal of CO2 from power plants is challenging because existing methods to separate CO2 from the gas mixture requires a significant fraction of the power plant output. Chemical-looping combustion (CLC) is a novel technology that utilizes a metal oxide oxygen carrier to transport oxygen to the fuel thereby avoiding direct contact between fuel and air. The use of CLC has the advantages of reducing the energy penalty while

205

The Stirred Tank Reactor Polymer Electrolyte Membrane Fuel Cell  

E-Print Network (OSTI)

The design and operation of a differential Polymer Electrolyte Membrane (PEM) fuel cell is described. The fuel cell design is based on coupled Stirred Tank Reactors (STR); the gas phase in each reactor compartment was well mixed. The characteristic times for reactant flow, gas phase diffusion and reaction were chosen so that the gas compositions at both the anode and cathode are uniform. The STR PEM fuel cell is one-dimensional; the only spatial gradients are transverse to the membrane. The STR PEM fuel cell was employed to examine fuel cell start- up, and its dynamic responses to changes in load, temperature and reactant flow rates. Multiple time scales in systems response are found to correspond to water absorption by the membrane, water transport through the membrane and stress-related mechanical changes of the membrane.

Benziger, J; Karnas, E; Moxley, J; Teuscher, C; Kevrekidis, Yu G; Benziger, Jay

2003-01-01T23:59:59.000Z

206

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.

207

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

208

Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems  

Science Conference Proceedings (OSTI)

The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

D. E. Shropshire

2009-01-01T23:59:59.000Z

209

Performance and fuel cycle cost study of the R2 reactor with HEU and LEU fuels  

SciTech Connect

A systematic study of the experiment performance and fuel cycle costs of the 50 MW R2 reactor operated by Studsvik Energiteknik AB has been performed using the current R2 HEU fuel, a variety of LEU fuel element designs, and two core-box/reflector configurations. The results include the relative performance of both in-core and ex-core experiments, control rod worths, and relative annual fuel cycle costs.

Pond, R.B.; Freese, K.E.; Matos, J.E.

1984-01-01T23:59:59.000Z

210

Reactor physics behavior of transuranic-bearing TRISO-particle fuel in a pressurized water reactor  

SciTech Connect

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU) - only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is retained. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint. (authors)

Pope, M. A.; Sen, R. S.; Ougouag, A. M.; Youinou, G. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Boer, B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); SCK-CEN, Boertang 200, BE-2400 Mol (Belgium)

2012-07-01T23:59:59.000Z

211

Reactor Physics Behavior of Transuranic-Bearing TRISO-Particle Fuel in a Pressurized Water Reactor  

SciTech Connect

Calculations have been performed to assess the neutronic behavior of pins of Fully-Ceramic Micro-encapsulated (FCM) fuel in otherwise-conventional Pressurized Water Reactor (PWR) fuel pins. The FCM fuel contains transuranic (TRU)-only oxide fuel in tri-isotropic (TRISO) particles with the TRU loading coming from the spent fuel of a conventional LWR after 5 years of cooling. Use of the TRISO particle fuel would provide an additional barrier to fission product release in the event of cladding failure. Depletion calculations were performed to evaluate reactivity-limited burnup of the TRU-only FCM fuel. These calculations showed that due to relatively little space available for fuel, the achievable burnup with these pins alone is quite small. Various reactivity parameters were also evaluated at each burnup step including moderator temperature coefficient (MTC), Doppler, and soluble boron worth. These were compared to reference UO{sub 2} and MOX unit cells. The TRU-only FCM fuel exhibits degraded MTC and Doppler coefficients relative to UO{sub 2} and MOX. Also, the reactivity effects of coolant voiding suggest that the behavior of this fuel would be similar to a MOX fuel of very high plutonium fraction, which are known to have positive void reactivity. In general, loading of TRU-only FCM fuel into an assembly without significant quantities of uranium presents challenges to the reactor design. However, if such FCM fuel pins are included in a heterogeneous assembly alongside LEU fuel pins, the overall reactivity behavior would be dominated by the uranium pins while attractive TRU destruction performance levels in the TRU-only FCM fuel pins is. From this work, it is concluded that use of heterogeneous assemblies such as these appears feasible from a preliminary reactor physics standpoint.

Michael A. Pope; R. Sonat Sen; Abderrafi M. Ougouag; Gilles Youinou; Brian Boer

2012-04-01T23:59:59.000Z

212

New Tool for Proliferation Resistance Evaluation Applied to Uranium and Thorium Fueled Fast Reactor Fuel Cycles  

E-Print Network (OSTI)

The comparison of nuclear facilities based on their barriers to nuclear material proliferation has remained a difficult endeavor, often requiring expert elicitation for each system under consideration. However, objectively comparing systems using a set of computable metrics to derive a single number representing a system is not, in essence, a nuclear nonproliferation specific problem and significant research has been performed for business models. For instance, Multi-Attribute Utility Analysis (MAUA) methods have been used previously to provide an objective insight of the barriers to proliferation. In this paper, the Proliferation Resistance Analysis and Evaluation Tool for Observed Risk (PRAETOR), a multi-tiered analysis tool based on the multiplicative MAUA method, is presented. It folds sixty three mostly independent metrics over three levels of detail to give an ultimate metric for nonproliferation performance comparison. In order to reduce analysts' bias, the weighting between the various metrics was obtained by surveying a total of thirty three nonproliferation specialists and nonspecialists from fields such as particle physics, international policy, and industrial engineering. The PRAETOR was used to evaluate the Fast Breeder Reactor Fuel Cycle (FBRFC). The results obtained using these weights are compared against a uniform weight approach. Results are presented for five nuclear material diversion scenarios: four examples include a diversion attempt on various components of a PUREX fast reactor cycle and one scenario involves theft from a PUREX facility in a LWR cycle. The FBRFC was evaluated with uranium-plutonium fuel and a second time using thorium-uranium fuel. These diversion scenarios were tested with both uniform and expert weights, with and without safeguards in place. The numerical results corroborate nonproliferation truths and provide insight regarding fast reactor facilities' proliferation resistance in relation to known standards.

Metcalf, Richard R.

2009-05-01T23:59:59.000Z

213

Electrometallurgical treatment of degraded N-reactor fuel  

Science Conference Proceedings (OSTI)

N-Reactor fuel constitutes almost 80% of the entire mass of the US Department of Energy's (DOE's) spent fuel inventory. The current plan for disposition of this fuel calls for interim dry storage, followed by direct repository disposal. However, this approach may not be viable for the entire inventory of N-Reactor fuel. The physical condition and chemical composition of much of the fuel have changed during the period that it has been in storage. The cladding of many of the fuel elements has been breached, allowing the metallic uranium fuel to react with water in the storage pools producing uranium oxides (U{sub x}O{sub y}) and uranium hydride (UH{sub 3}). Even if the breached fuel is placed in dry storage, it may continue to undergo significant changes caused by the reaction of exposed uranium with any remaining water in the container. Uranium oxides, uranium hydride, and hydrogen gas are expected to form as a result of this reaction. The presence of potentially explosive hydrogen and uranium hydride, which under certain conditions is pyrophoric, raises technical concerns that will need to be addressed. The electrometallurgical treatment process developed by Argonne National Laboratory (ANL) has potential for conditioning degraded N-Reactor fuel for long-term storage or disposal. The first step in evaluating the applicability of this process is the preparation of degraded fuel that is similar to the actual degraded N-Reactor fuel. Subsequently, the simulated degraded fuel can be introduced into an electrorefiner to examine the effect of corrosion products on the electrorefining process. Some of the technical issues to be resolved include the viability of direct electrorefining without a head-end reduction step, the effect of adherent corrosion products on the electrorefining kinetics, and the recovery and treatment of loose corrosion products that pull away from the degraded fuel. This paper presents results from an experimental study of the preparation, characterization, and subsequent electrometallurgical treatment of samples of simulated degraded N-Reactor fuel.

Gourishankar, K. V.; Karell, E. J.; Everhart, R. E.; Indacochea, E.

2000-03-03T23:59:59.000Z

214

System for fuel rod removal from a reactor module  

DOE Patents (OSTI)

A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

Matchett, Richard L. (Bethel Park, PA); Roof, David R. (North Huntingdon, PA); Kikta, Thomas J. (Pittsburgh, PA); Wilczynski, Rosemarie (McKees Rocks, PA); Nilsen, Roy J. (Pittsburgh, PA); Bacvinskas, William S. (Bethel Park, PA); Fodor, George (Pittsburgh, PA)

1990-01-01T23:59:59.000Z

215

System for fuel rod removal from a reactor module  

DOE Patents (OSTI)

A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

1988-07-28T23:59:59.000Z

216

High Density Fuel Development for Research Reactors  

SciTech Connect

An international effort to develop, qualify, and license high and very high density fuels has been underway for several years within the framework of multi-national RERTR programs. The current development status is the result of significant contributions from many laboratories, specifically CNEA in Argentina, AECL in Canada, CEA in France, TUM in Germany, KAERI in Korea, VNIIM, RDIPE, IPPE, NCCP and RIARR in Russia, INL, ANL and Y-12 in USA. These programs are mainly engaged with UMo dispersion fuels with densities from 6 to 8 gU/cm3 (high density fuel) and UMo monolithic fuel with density as high as 16 gU/cm3 (very high density fuel). This paper, mainly focused on the French & US programs, gives the status of high density UMo fuel development and perspectives on their qualification.

Daniel Wachs; Dennis Keiser; Mitchell Meyer; Douglas Burkes; Curtis Clark; Glenn Moore; Jan-Fong Jue; Totju Totev; Gerard Hofman; Tom Wiencek; Yeon So Kim; Jim Snelgrove

2007-09-01T23:59:59.000Z

217

Reactor Physics Assessment of the Inclusion of Unseparated Neptunium in MOX Reactor Fuel  

Science Conference Proceedings (OSTI)

Reducing the number of actinide separation streams in a spent fuel recovery process would reduce the cost and complexity of the process, and lower the quantity and numbers of solvents needed. It is more difficult and costly to separate Np and recombine it with Am-Cm prior to co-conversion than to simply co-strip it with the U-Pu-Np. Inclusion of the Np in mixed oxide (MOX) fuel for light water reactor (LWR) applications should not seriously affect the operating behavior of the reactor, nor should it pose insurmountable fuel design issues. In this work, the U, Pu, and Np from typical discharged and cooled PWR spent nuclear fuel are assumed to be used together in the preparation of MOX fuel for use in a pressurized water reactor (PWR). The reactor grade Pu isotopic vector is used in the model and the relative mass ratio of the Pu and Np content (Np/Pu mass is 0.061) from the cooled spent fuel is maintained but the overall Pu-Np MOX wt% is adjusted with respect to the U content (assumed to be at 0.25 wt% 235U enrichment) to offset reactivity and cycle length effects. The SCALE 5.1 scientific package (especially modules TRITON, NEWT, ORIGEN-S, ORIGEN-ARP) was used for the calculations presented in this paper. A typical Westinghouse 17x17 fuel assembly design was modeled at nominal PWR operating conditions. It was seen that U-Pu-Np MOX fuel with NpO2 and PuO2 representing 11.5wt% of the total MOX fuel would be similar to standard MOX fuel in which PuO2 is 9wt% of the fuel. The reactivity, isotopic composition, and neutron and ? sources, and the decay heat details for the discharged MOX fuel are presented and discussed in this paper.

Ellis, Ronald James [ORNL

2009-01-01T23:59:59.000Z

218

Spent fuel drying system test results (second dry-run)  

DOE Green Energy (OSTI)

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks have been detected in the basins and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the second dry-run test, which was conducted without a fuel element. With the concurrence of project management, the test protocol for this run, and subsequent drying test runs, was modified. These modifications were made to allow for improved data correlation with drying procedures proposed under the IPS. Details of these modifications are discussed in Section 3.0.

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01T23:59:59.000Z

219

Fuel pins with both target and fuel pellets in an isotope-production reactor  

DOE Patents (OSTI)

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

220

CHEMICAL PROBLEMS OF NON-AQUEOUS FLUID-FUEL REACTORS  

SciTech Connect

The three main chemical problems of non-aqueous fluidfuel reactors are selection of a fuel system which meets nuclear and thermal requirements, control of corrosion of structural materials, and development of an efficient and economical separation process. The problems of the fastneutron reactor with a core of fused NaCl, PbCl/sub 2/, UCl/sub 4/ mixture and a blanket of fused UCl/ sub 4/ are discussed, except for the separation process. Brief treatment is given the Bi- U - Pu fuel system for thermal reactors, including tentative flowsheets for the separation process. A detailed discussion of the experiments of Bareis at Brookhaven and related experiments on the distribution of various metals between liquid Bi and fused-salt solutions is given, the experimental results correlated, and application made to reactor problems. A general discussion is given of nonaqueous high-temperature separation processes (L.M.T.)

Scatchard, G.; Clark, H.M.; Golden, S.; Boltax, A.; Schuhmann, R. Jr.

1952-10-15T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

LOFT contribution to PWR fuel analysis and testing  

SciTech Connect

The paper provides a brief overview of the Loss-of-Fluid Test (LOFT) Facility, a review of the key results from LOCA tests performed to date, and plans for the future of LOFT, as an introduction to the more specific information about LOFT fuel that is to follow. The purpose of the LOFT tests is to provide data for assessing the accuracy of the analytical models used in evaluating the safety of commercial nuclear power plants. While the main purpose of LOFT is to investigate phenomena important to the loss-of-coolant accident (LOCA), the 55 MWt LOFT reactor system has been scaled to commercial power reactors and heavily instrumented so that information on fuel and system behavior over a range of normal and off-normal conditions can be obtained.

Leach, L.P.

1978-01-01T23:59:59.000Z

222

Characterization of spent fuel approved testing material: ATM-106  

Science Conference Proceedings (OSTI)

The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well- characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs.

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thornhill, C.K.

1988-10-01T23:59:59.000Z

223

Characterization of spent fuel approved testing material--ATM-104  

SciTech Connect

The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1991-12-01T23:59:59.000Z

224

Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel  

SciTech Connect

This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

1994-10-01T23:59:59.000Z

225

Updated FY12 Ceramic Fuels Irradiation Test Plan  

SciTech Connect

The Fuel Cycle Research and Development program is currently devoting resources to study of numerous fuel types with the aim of furthering understanding applicable to a range of reactors and fuel cycles. In FY11, effort within the ceramic fuels campaign focused on planning and preparation for a series of rabbit irradiations to be conducted at the High Flux Isotope Reactor located at Oak Ridge National Laboratory. The emphasis of these planned tests was to study the evolution of thermal conductivity in uranium dioxide and derivative compositions as a function of damage induced by neutron damage. Current fiscal realities have resulted in a scenario where completion of the planned rabbit irradiations is unlikely. Possibilities for execution of irradiation testing within the ceramic fuels campaign in the next several years will thus likely be restricted to avenues where strong synergies exist both within and outside the Fuel Cycle Research and Development program. Opportunities to augment the interests and needs of modeling, advanced characterization, and other campaigns present the most likely avenues for further work. These possibilities will be pursued with the hope of securing future funding. Utilization of synthetic microstructures prepared to better understand the most relevant actors encountered during irradiation of ceramic fuels thus represents the ceramic fuel campaign's most efficient means to enhance understanding of fuel response to burnup. This approach offers many of the favorable attributes embraced by the Separate Effects Testing paradigm, namely production of samples suitable to study specific, isolated phenomena. The recent success of xenon-imbedded thick films is representative of this approach. In the coming years, this strategy will be expanded to address a wider range of problems in conjunction with use of national user facilities novel characterization techniques to best utilize programmatic resources to support a science-based research program.

Nelson, Andrew T. [Los Alamos National Laboratory

2012-05-24T23:59:59.000Z

226

Reactor-specific spent fuel discharge projections, 1984 to 2020  

Science Conference Proceedings (OSTI)

The original spent fuel utility data base (SFDB) has been adjusted to produce agreement with the EIA nuclear energy generation forecast. The procedure developed allows the detail of the utility data base to remain intact, while the overall nuclear generation is changed to match any uniform nuclear generation forecast. This procedure adjusts the weight of the reactor discharges as reported on the SFDB and makes a minimal (less than 10%) change in the original discharge exposures in order to preserve discharges of an integral number of fuel assemblies. The procedure used in developing the reactor-specific spent fuel discharge projections, as well as the resulting data bases themselves, are described in detail in this report. Discussions of the procedure cover the following topics: a description of the data base; data base adjustment procedures; addition of generic power reactors; and accuracy of the data base adjustments. Reactor-specific discharge and storage requirements are presented. Annual and cumulative discharge projections are provided. Annual and cumulative requirements for additional storage are shown for the maximum at-reactor (AR) storage assumption, and for the maximum AR with transshipment assumption. These compare directly to the storage requirements from the utility-supplied data, as reported in the Spent Fuel Storage Requirements Report. The results presented in this report include: the disaggregated spent fuel discharge projections; and disaggregated projections of requirements for additional spent fuel storage capacity prior to 1998. Descriptions of the methodology and the results are included in this report. Details supporting the discussions in the main body of the report, including descriptions of the capacity and fuel discharge projections, are included. 3 refs., 6 figs., 12 tabs.

Heeb, C.M.; Libby, R.A.; Holter, G.M.

1985-04-01T23:59:59.000Z

227

Spent nuclear fuel discharges from US reactors 1992  

SciTech Connect

This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

Not Available

1994-05-05T23:59:59.000Z

228

Large-diameter, high-plutonium metallic fuel testing in EBR-II  

Science Conference Proceedings (OSTI)

The integral fast reactor (IFR) concept makes use of U-Pu-Zr metallic fuel alloys because of favorable core behavior characteristics and compatibility with a pyrometallurgical reprocessing scheme. Advanced reactor designs use U-x Pu-10 Zr (where 20 {le} x {le} 28). Most of the IFR fuel performance database was obtained with fuel containing {le}20 wt% Pu having a 0.439-cm diameter irradiated in cladding tubes having a 0.584-cm o.d. However, commercial implementations of the IFR concept (e.g., the advanced liquid-metal reactor by General Electric) will likely use fuel of larger diameter and plutonium contents {ge}20 wt%. The HT9 advanced driver fuel test, irradiated in Experimental Breeder Reactor II (EBR-II), was initiated to obtain irradiation performance data from larger diameter fuel cast with a range of plutonium compositions and clad in the reference cladding alloy HT9.

Crawford, D.C.; Hayes, S.L.; Pahl, R.G. [Argonne National Lab., Idaho Falls, ID (United States)

1994-12-31T23:59:59.000Z

229

PLUTONIUM FUEL PROCESSING AND FABRICATION FOR FAST CERAMIC REACTORS  

SciTech Connect

>A study was made of the processes available for fabrication of plutonium-containing fuel from a fast ceramic reacter, and for chemical reprocessing of irradiated fuel. Radiations from recycled plutonium are evaluated. Adaptation of conventional glove-box handling procedures to the fabrication of recycle plutonium appears practical. It is concluded that acceptable costs are obtainable using moderate extensions of conventional glove- box fabrication methods and wet processing techniques, provided a significant volume of production is available. The minimum economic scale for the preferred chemical reprocessing method, anion exchange, is about 500 Mw(e) of reactor capacity. The minimum scale of economic operation for the fuel refabrication facility corresponds to three 500 Mw(e) reactors, if only steady-state refueling provides the fabrication load. The minimum volume required falls to one 500 Mw(e) reactor, if the continued growth of capacity provides fabrication volume equal to that for refueling. The chemical reprocessing costs obtained range from 0.27 mills/kwh for 1500 Mw(e) of reactor capacity, to 0.10 mills/kwh for 3000 Mw(e) of capacity. The estimated fuel fabrication cost is l/kg of uranium and plutonium in the core region (excluding axial and radial blankets) or .06/ g of plutonium content, When axial blankets, fabricated in the same rods, are included; the combined average is 34/kg of uranium and plutonium. Radial blanket fabrication cost is /kg of uranium. The overall average of all fuel and blankets is /kg of uranium and plutonium. The fabrication cost is 0.29 mills/kwh for a production rate corresponding to 3000 Mw(e) of capacity (or 1500 Mw(e) of capacity plus growth equivalent to one additional reactor core per year). For one 525 Mw(e) reactor, (plus equivalent growth volume) the fabrication cost becomes 0.42 mills/ kwh. (All fuel throughputs are based on fuel life of 100,000 MWD/T.) Using the estimates developed, the total fuel cycle cost for a typical fast reactor design using PuO/sub 2/UO/sub 2/ fuel is estimated to be about 0.9 mills/kwh. (auth)

Zebroski, E.L.; Alter, H.W.; Collins, G.D.

1962-02-01T23:59:59.000Z

230

Retrievable fuel pin end member for a nuclear reactor  

DOE Patents (OSTI)

A bottom end member (17b) on a retrievable fuel pin (13b) secures the pin (13b) within a nuclear reactor (12) by engaging on a transverse attachment rail (18) with a spring clip type of action. Removal and reinstallation if facilitated as only axial movement of the fuel pin (13b) is required for either operation. A pair of resilient axially extending blades (31) are spaced apart to define a slot (24) having a seat region (34) which receives the rail (18) and having a land region (37), closer to the tips (39) of the blades (31) which is normally of less width than the rail (18). Thus an axially directed force sufficient to wedge the resilient blades (31) apart is required to emplace or release the fuel pin (13b) such force being greater than the axial forces on the fuel pins (13b) which occur during operation of the reactor (12).

Rosa, Jerry M. (Los Gatos, CA)

1982-01-01T23:59:59.000Z

231

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

232

Proliferation resistant fuel for pebble bed modular reactors  

SciTech Connect

We show that it is possible to denature the Plutonium produced in Pebble Bed Modular Reactors (PBMR) by doping the nuclear fuel with either 3050 ppm of {sup 237}Np or 2100 ppm of Am vector. A correct choice of these isotopes concentration yields denatured Plutonium with isotopic ratio {sup 238}Pu/Pu {>=} 6%, for the entire fuel burnup cycle. The penalty for introducing these isotopes into the nuclear fuel is a subsequent shortening of the fuel burnup cycle, with respect to a non-doped reference fuel, by 41.2 Full Power Days (FPDs) and 19.9 FPDs, respectively, which correspond to 4070 MWd/ton and 1965 MWd/ton reduction in fuel discharge burnup. (authors)

Ronen, Y.; Aboudy, M.; Regev, D.; Gilad, E. [Dept. of Nuclear Engineering, Ben-Gurion Univ. of the Negev, Beer-Sheva 84105 (Israel)

2012-07-01T23:59:59.000Z

233

Means for supporting fuel elements in a nuclear reactor  

DOE Patents (OSTI)

A grid structure for a nuclear reactor fuel assembly comprising a plurality of connecting members forming at least one longitudinally extending opening peripheral and inner fuel element openings through each of which openings at least one nuclear fuel element extends, said connecting members forming wall means surrounding said each peripheral and inner fuel element opening, a pair of rigid projections longitudinally spaced from one another extending from a portion of said wall means into said each peripheral and inner opening for rigidly engaging said each fuel element, respectively, yet permit individual longitudinal slippage thereof, and resilient means formed integrally on and from said wall means and positioned in said each peripheral and inner opening in opposed relationship with said projections and located to engage said fuel element to bias the latter into engagement with said rigid projections, respectively

Andrews, Harry N. (Murrysville, PA); Keller, Herbert W. (Monroeville, PA)

1980-01-01T23:59:59.000Z

234

Residential Fuel Cell Performance Test Facility  

Science Conference Proceedings (OSTI)

... Currently, the test facility is setup to deliver natural gas as the fuel, but ... A turbine and magnetic flow meter measure the flow of water for the domestic ...

2011-11-15T23:59:59.000Z

235

FUEL ELEMENTS FOR THERMAL-FISSION NUCLEAR REACTORS  

DOE Patents (OSTI)

Fuel elements for thermal-fission nuclear reactors are described. The fuel element is comprised of a core of alumina, a film of a metal of the class consisting of copper, silver, and nickel on the outer face of the core, and a coating of an oxide of a metal isotope of the class consisting of Un/sup 235/, U/ sup 233/, and Pu/sup 239/ on the metal f ilm.

Flint, O.

1961-01-10T23:59:59.000Z

236

Fuel leak detection apparatus for gas cooled nuclear reactors  

SciTech Connect

Apparatus is disclosed for detecting nuclear fuel leaks within nuclear power system reactors, such as high temperature gas cooled reactors. The apparatus includes a probe assembly that is inserted into the high temperature reactor coolant gaseous stream. The probe has an aperture adapted to communicate gaseous fluid between its inside and outside surfaces and also contains an inner tube for sampling gaseous fluid present near the aperture. A high pressure supply of noncontaminated gas is provided to selectively balance the pressure of the stream being sampled to prevent gas from entering the probe through the aperture. The apparatus includes valves that are operable to cause various directional flows and pressures, which valves are located outside of the reactor walls to permit maintenance work and the like to be performed without shutting down the reactor.

Burnette, Richard D. (San Diego, CA)

1977-01-01T23:59:59.000Z

237

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

238

Performance and emissions of a catalytic reactor with propane, diesel, and Jet A fuels  

DOE Green Energy (OSTI)

As part of the ERDA-funded Gas Turbine Highway Vehicle Systems project, tests were made to determine the performance and emissions of a catalytic reactor operated with propane, No. 2 diesel, and Jet A fuels. A 12-cm diameter and 16-cm long catalytic reactor using a proprietary noble metal catalyst was operated at an inlet temperature of 800 K, a pressure of 3 x 10/sup 5/ Pa and reference velocities of 10 to 15 m/s. No significant differences between the performance of the three fuels were observed when 98.5% purity propane was used. The combustion efficiency for 99.8% purity propane tested later was significantly lower, however. The diesel fuel contained 135 ppM of bound nitrogen and consequently produced the highest NO/sub x/ emissions of the three fuels. As much as 85% of the bound nitrogen was converted to NO/sub x/. Steady-state emissions goals based on half the most stringent proposed automotive standards were met when the reactor was operated at an adiabatic combustion temperature higher than 1350 K with all fuels except the 99.8% purity propane. With that fuel, a minimum temperature of 1480 K was required.

Anderson, D.N.

1977-01-01T23:59:59.000Z

239

Fuels  

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

Goals > Fuels Goals > Fuels XMAT for nuclear fuels XMAT is ideally suited to explore all of the radiation processes experienced by nuclear fuels.The high energy, heavy ion accleration capability (e.g., 250 MeV U) can produce bulk damage deep in the sample, achieving neutron type depths (~10 microns), beyond the range of surface sputtering effects. The APS X-rays are well matched to the ion beams, and are able to probe individual grains at similar penetrations depths. Damage rates to 25 displacements per atom per hour (DPA/hr), and doses >2500 DPA can be achieved. MORE» Fuels in LWRs are subjected to ~1 DPA per day High burn-up fuel can experience >2000 DPA. Traditional reactor tests by neutron irradiation require 3 years in a reactor and 1 year cool down. Conventional accelerators (>1 MeV/ion) are limited to <200-400 DPAs, and

240

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

Note: This page contains sample records for the topic "testing reactor fuels" 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

Mechanical modeling of porous oxide fuel pellet A Test Problem  

Science Conference Proceedings (OSTI)

A poro-elasto-plastic material model has been developed to capture the response of oxide fuels inside the nuclear reactors under operating conditions. Behavior of the oxide fuel and variation in void volume fraction under mechanical loading as predicted by the developed model has been reported in this article. The significant effect of void volume fraction on the overall stress distribution of the fuel pellet has also been described. An important oxide fuel issue that can have significant impact on the fuel performance is the mechanical response of oxide fuel pellet and clad system. Specifically, modeling the thermo-mechanical response of the fuel pellet in terms of its thermal expansion, mechanical deformation, swelling due to void formation and evolution, and the eventual contact of the fuel with the clad is of significant interest in understanding the fuel-clad mechanical interaction (FCMI). These phenomena are nonlinear and coupled since reduction in the fuel-clad gap affects thermal conductivity of the gap, which in turn affects temperature distribution within the fuel and the material properties of the fuel. Consequently, in order to accurately capture fuel-clad gap closure, we need to account for fuel swelling due to generation, retention, and evolution of fission gas in addition to the usual thermal expansion and mechanical deformation. Both fuel chemistry and microstructure also have a significant effect on the nucleation and growth of fission gas bubbles. Fuel-clad gap closure leading to eventual contact of the fuel with the clad introduces significant stresses in the clad, which makes thermo-mechanical response of the clad even more relevant. The overall aim of this test problem is to incorporate the above features in order to accurately capture fuel-clad mechanical interaction. Because of the complex nature of the problem, a series of test problems with increasing multi-physics coupling features, modeling accuracy, and complexity are defined with the objective of accurate simulation of fuel-clad mechanical interaction subjected to a wide-range of thermomechanical stimuli.

Nukala, Phani K [ORNL; Barai, Pallab [ORNL; Simunovic, Srdjan [ORNL; Ott, Larry J [ORNL

2009-10-01T23:59:59.000Z

242

Method of locating a leaking fuel element in a fast breeder power reactor  

DOE Patents (OSTI)

Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

Honekamp, John R. (Downers Grove, IL); Fryer, Richard M. (Idaho Falls, ID)

1978-01-01T23:59:59.000Z

243

Introduction of Thorium-Based Fuels in High Conversion Pressurized Water Reactors  

Science Conference Proceedings (OSTI)

Regular Technical Paper / Special Issue on the Symposium on Radiation Effects in Ceramic Oxide and Novel LWR Fuels / Fission Reactors

V. Vallet; B. Gastaldi; J. Politello; A. Santamarina; L. Van Den Durpel

244

Fuel damage during off-normal transients in metal-fueled fast reactors  

SciTech Connect

Fuel damage during off-normal transients is a key issue in the safety of fast reactors because the fuel pin cladding provides the primary barrier to the release of radioactive materials. Part of the Safety Task of the Integral Fast Reactor Program is to provide assessments of the damage and margins to failure for metallic fuels over the wide range of transients that must be considered in safety analyses. This paper reviews the current status of the analytical and experimental programs that are providing the bases for these assessments. 13 refs., 2 figs.

Kramer, J.M.; Bauer, T.H.

1990-01-01T23:59:59.000Z

245

Method of controlling crystallite size in nuclear-reactor fuels  

DOE Patents (OSTI)

Improved spherules for making enhanced forms of nuclear-reactor fuels are prepared by internal gelation procedures within a sol-gel operation and are accomplished by first boiling the concentrated HMTA-urea feed solution before engaging in the spherule-forming operation thereby effectively controlling crystallite size in the product spherules.

Lloyd, Milton H. (Oak Ridge, TN); Collins, Jack L. (Knoxville, TN); Shell, Sam E. (Oak Ridge, TN)

1985-01-01T23:59:59.000Z

246

METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR  

DOE Patents (OSTI)

A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

Layer, E.H. Jr.; Peet, C.S.

1962-01-23T23:59:59.000Z

247

Nuclear breeder reactor fuel element with silicon carbide getter  

DOE Patents (OSTI)

An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA)

1987-01-01T23:59:59.000Z

248

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

249

Electrolysis cell for reprocessing plutonium reactor fuel  

DOE Patents (OSTI)

An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals is claimed. The cell includes a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket. The anode basket is extendable into the lower pool to dissolve at least some metallic contaminants; the anode basket contains the spent fuel acting as a second anode when in the electrolyte.

Miller, W.E.; Steindler, M.J.; Burris, L.

1985-01-04T23:59:59.000Z

250

Electrolysis cell for reprocessing plutonium reactor fuel  

DOE Patents (OSTI)

An electrolytic cell for refining a mixture of metals including spent fuel containing U and Pu contaminated with other metals, the cell including a metallic pot containing a metallic pool as one anode at a lower level, a fused salt as the electrolyte at an intermediate level and a cathode and an anode basket in spaced-apart positions in the electrolyte with the cathode and anode being retractable to positions above the electrolyte during which spent fuel may be added to the anode basket and the anode basket being extendable into the lower pool to dissolve at least some metallic contaminants, the anode basket containing the spent fuel acting as a second anode when in the electrolyte.

Miller, William E. (Naperville, IL); Steindler, Martin J. (Park Forest, IL); Burris, Leslie (Naperville, IL)

1986-01-01T23:59:59.000Z

251

Interim results from UO/sub 2/ fuel oxidation tests in air  

Science Conference Proceedings (OSTI)

An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to extend the characterization of spent fuel oxidation in air. To characterize oxidation behavior of irradiated UO/sub 2/, fuel oxidation tests were performed on declad light-water reactor spent fuel and nonirradited UO/sub 2/ pellets in the temperature range of 135 to 250/sup 0/C. These tests were designed to determine the important independent variables that might affect spent fuel oxidation behavior. The data from this program, when combined with the test results from other programs, will be used to develop recommended spent fuel dry-storage temperature limits in air. This report describes interim test results. The initial PNL investigations of nonirradiated and spent fuels identified the important testing variables as temperature, fuel burnup, radiolysis of the air, fuel microstructure, and moisture in the air. Based on these initial results, a more extensive statistically designed test matrix was developed to study the effects of temperature, burnup, and moisture on the oxidation behavior of spent fuel. Oxidation tests were initiated using both boiling-water reactor and pressurized-water reactor fuels from several different reactors with burnups from 8 to 34 GWd/MTU. A 10/sup 5/ R/h gamma field was applied to the test ovens to simulate dry storage cask conditions. Nonirradiated fuel was included as a control. This report describes experimental results from the initial tests on both the spent and nonirradiated fuels and results to date on the tests in a 10/sup 5/ R/h gamma field. 33 refs., 51 figs., 6 tabs.

Campbell, T.K.; Gilbert, E.R.; Thornhill, C.K.; White, G.D.; Piepel, G.F.; Griffin, C.W.j

1987-08-01T23:59:59.000Z

252

Sodium fast reactor fuels and materials : research needs.  

SciTech Connect

An expert panel was assembled to identify gaps in fuels and materials research prior to licensing sodium cooled fast reactor (SFR) design. The expert panel considered both metal and oxide fuels, various cladding and duct materials, structural materials, fuel performance codes, fabrication capability and records, and transient behavior of fuel types. A methodology was developed to rate the relative importance of phenomena and properties both as to importance to a regulatory body and the maturity of the technology base. The technology base for fuels and cladding was divided into three regimes: information of high maturity under conservative operating conditions, information of low maturity under more aggressive operating conditions, and future design expectations where meager data exist.

Denman, Matthew R.; Porter, Douglas (Idaho National Laboratory, Idaho Falls, ID); Wright, Art (Argonne National Laboratory Argonne, IL); Lambert, John (Argonne National Laboratory Argonne, IL); Hayes, Steven (Idaho National Laboratory, Idaho Falls, ID); Natesan, Ken (Argonne National Laboratory Argonne, IL); Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Garner, Frank (Radiation Effects Consulting. Richland, WA); Walters, Leon (Advanced Reactor Concepts, Idaho Falls, ID); Yacout, Abdellatif (Argonne National Laboratory Argonne, IL)

2011-09-01T23:59:59.000Z

253

Use of Sensitivity and Uncertainty Analysis in the Design of Reactor Physics and Criticality Benchmark Experiments for Advanced Nuclear Fuel  

Science Conference Proceedings (OSTI)

Technical Paper / Advances in Nuclear Fuel Management - Increased Enrichment/High Burnup and Light Water Reactor Fuel Cycle Optimization

B. T. Rearden; W. J. Anderson; G. A. Harms

254

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

255

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

256

Power Burst Facility (PBF) severe fuel damage test 1-4 test results report  

DOE Green Energy (OSTI)

A comprehensive evaluation of the Severe Fuel Damage (SFD) Test 1-4 performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory is presented. Test SFD 1-4 was the fourth and final test in an internationally sponsored light water reactor severe accident research program, initiated by the US Nuclear Regulatory Commission. The overall technical objective of the test was to contribute to the understanding of fuel and control rod behavior, aerosol and hydrogen generation, and fission product release and transport during a high-temperature, severe fuel damage transient. A test bundle, comprised of 26 previously irradiated (36,000 MWd/MtU) pressurized water-reactor-type fuel rods, 2 fresh instrumented fuel rods, and 4 silver-indium-cadmium control rods, was surrounded by an insulating shroud and contained in a pressurized in-pile tube. The experiment consisted of a 1.3-h transient at a coolant pressure of 6.95 MPa in which the inlet coolant flow to the bundle was reduced to 0.6 g/s while the bundle fission power was gradually increased until dryout, heatup, cladding rupture, and oxidation occurred. With sustained fission power and heat from oxidation, temperatures continued to rise rapidly, resulting in zircaloy and control rod absorber alloy melting, fuel liquefaction, material relocation, and the release of hydrogen, aerosols, and fission products. The transient was terminated over a 2100-s time span by slowly reducing the reactor power and cooling the damaged bundle with argon gas. A description and evaluation of the major phenomena, based upon the response of on-line instrumentation, analysis of fission product and aerosol data, postirradiation examination of the fuel bundle, and calculations using the SCDAP/RELAP5 computer code, are presented. 40 refs., 160 figs., 31 tabs.

Petti, D.A.; Martinson, Z.R.; Hobbins, R.R.; Allison, C.M.; Carlson, E.R.; Hagrman, D.L.; Cheng, T.C.; Hartwell, J.K.; Vinjamuri, K.; Seifken, L.J.

1989-04-01T23:59:59.000Z

257

PBF (Power Burst Facility) severe fuel damage test 1--3 test results report  

Science Conference Proceedings (OSTI)

A comprehensive evaluation of the Severe Fuel Damage (SFD) Test 1--3 performed in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory is presented. Test SFD 1--3 was the third test in an internationally sponsored light water reactor severe accident research program, initiated by the US Nuclear Regulatory Commission. The overall technical objective of the test was to contribute to the understanding of fuel rod behavior, hydrogen generation, and fission product release and transport during a high-temperature, severe fuel damage transient. A test bundle, comprised of 26 previously irradiated (38,000 MWd/tU) pressurized water reactor-type fuel rods, 2 fresh instrumented fuel rods, and 4 empty zircaloy guide tubes, was surrounded by an insulating shroud and contained in a pressurized in-pile tube. The experiment consisted of a 1-h transient at a nominal coolant pressure of 6.85 MPa in which the inlet coolant flow to the bundle was reduced to 0.6 g/s while the bundle fission power was gradually increased until dryout, heatup, cladding rupture, and oxidation occurred. With sustained fission power and heat from oxidation, temperatures continued to rise rapidly, resulting in zircaloy melting, fuel liquefaction, material relocation, and the release of hydrogen, aerosols, and fission products. The transient was terminated over a 1340-s time span by slowly reducing the reactor power and cooling the damaged bundle with argon gas. A description and evaluation of the major phenomena, based upon the response of online instrumentation, analysis of fission product data, postirradiation examination of the fuel bundle, and calculations using the SCDAP/RELAP5 computer code, are presented. 34 refs., 241 figs., 51 tabs.

Martinson, Z.R.; Gasparini, M.; Hobbins, R.R.; Petti, D.A.; Allison, C.M.; Hohorst, J.K.; Hagrman, D.L.; Vinjamuri, K. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1989-10-01T23:59:59.000Z

258

Storage of LWR (light-water-reactor) spent fuel in air  

Science Conference Proceedings (OSTI)

An experimental program is being conducted at Pacific Northwest Laboratory (PNL) to determine the oxidation response of light-water-reactor (LWR) spent fuels under conditions appropriate to fuel storage in air. The program is designed to investigate several independent variables that might affect the oxidation behavior of spent fuel. Included are temperature (135 to 230{degree}C), fuel burnup (to about 34 MWd/kgM), reactor type (pressurized and boiling water reactors), moisture level in the air, and the presence of a high gamma field. In continuing tests with declad spent fuel and nonirradiated UO{sub 2} specimens, oxidation rates were monitored by weight-gain measurements and the microstructures of subsamples taken during the weighing intervals were characterized by several analytical methods. The oxidation behavior indicated by weight gain and time to form powder will be reported in Volume III of this series. The characterization results obtained from x-ray diffractometry, transmission electron microscopy, scanning electron microscopy, and Auger electron spectrometry of oxidized fuel samples are presented in this report. 28 refs., 21 figs., 3 tabs.

Thomas, L.E.; Charlot, L.A.; Coleman, J.E. (Pacific Northwest Lab., Richland, WA (USA)); Knoll, R.W. (Johnson Controls, Inc., Madison, WI (USA))

1989-12-01T23:59:59.000Z

259

Optimization of Boiling Water Reactor Fuel Crud Characteristics for Reducing Radiation Fields: Evaluation of BWR Fuel Crud Properties  

Science Conference Proceedings (OSTI)

Fuel crud formation and its properties are the combined result of many factors, including corrosion product input, zinc addition rates, reactor coolant chemistry, and fuel and core design. Crud deposition may impact fuel performance as well as radiation field generation. Many projects have evaluated changes in fuel crud properties resulting from changing reactor coolant chemistry. However, the desired crud properties for both good fuel performance and mitigation of radiation field source term are ...

2013-11-26T23:59:59.000Z

260

Overview of the fast reactors fuels program. [LMFBR  

SciTech Connect

Each nation involved in LMFBR development has its unique energy strategies which consider energy growth projections, uranium resources, capital costs, and plant operational requirements. Common to all of these strategies is a history of fast reactor experience which dates back to the days of the Manhatten Project and includes the CLEMENTINE Reactor, which generated a few watts, LAMPRE, EBR-I, EBR-II, FERMI, SEFOR, FFTF, BR-1, -2, -5, -10, BOR-60, BN-350, BN-600, JOYO, RAPSODIE, Phenix, KNK-II, DFR, and PFR. Fast reactors under design or construction include PEC, CRBR, SuperPhenix, SNR-300, MONJU, and Madras (India). The parallel fuels and materials evolution has fully supported this reactor development. It has involved cermets, molten plutonium alloy, plutonium oxide, uranium metal or alloy, uranium oxide, and mixed uranium-plutonium oxides and carbides.

Evans, E.A.; Cox, C.M.; Hayward, B.R.; Rice, L.H.; Yoshikawa, H.H.

1980-04-01T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

Spent fuel drying system test results (first dry-run)  

DOE Green Energy (OSTI)

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site. Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL) on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of the first dry-run test, which was conducted without a fuel element. The empty test apparatus was subjected to a combination of low- and high-temperature vacuum drying treatments that were intended to mimic, wherever possible, the fuel treatment strategies of the IPS. The data from this dry-run test can serve as a baseline for the first two fuel element tests, 1990 (Run 1) and 3128W (Run 2). The purpose of this dry-run was to establish the background levels of hydrogen in the system, and the hydrogen generation and release characteristics attributable to the test system without a fuel element present. This test also serves to establish the background levels of water in the system and the water release characteristics. The system used for the drying test series was the Whole Element Furnace Testing System, described in Section 2.0, which is located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodology are given in section 3.0, and the experimental results provided in Section 4.0. These results are further discussed in Section 5.0.

Klinger, G.S.; Oliver, B.M.; Abrefah, J.; Marschman, S.C.; MacFarlan, P.J.; Ritter, G.A.

1998-07-01T23:59:59.000Z

262

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

263

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

264

A Parametric Study of the DUPIC Fuel Cycle to Reflect Pressurized Water Reactor Fuel Management Strategy  

SciTech Connect

For both pressurized water reactor (PWR) and Canada deuterium uranium (CANDU) tandem analysis, the Direct Use of spent PWR fuel In CANDU reactor (DUPIC) fuel cycle in a CANDU 6 reactor is studied using the DRAGON/DONJON chain of codes with the ENDF/B-V and ENDF/B-VI libraries. The reference feed material is a 17 x 17 French standard 900-MW(electric) PWR fuel. The PWR spent-fuel composition is obtained from two-dimensional DRAGON assembly transport and depletion calculations. After a number of years of cooling, this defines the initial fuel nuclide field in the CANDU unit cell calculations in DRAGON, where it is further depleted with the same neutron group structure. The resulting macroscopic cross sections are condensed and tabulated to be used in a full-core model of a CANDU 6 reactor to find an optimized channel fueling rate distribution on a time-average basis. Assuming equilibrium refueling conditions and a particular refueling sequence, instantaneous full-core diffusion calculations are finally performed with the DONJON code, from which both the channel power peaking factors and local parameter effects are estimated. A generic study of the DUPIC fuel cycle is carried out using the linear reactivity model for initial enrichments ranging from 3.2 to 4.5 wt% in a PWR. Because of the uneven power histories of the spent PWR assemblies, the spent PWR fuel composition is expected to differ from one assembly to the next. Uneven mixing of the powder during DUPIC fuel fabrication may lead to uncertainties in the composition of the fuel bundle and larger peaking factors in CANDU. A mixing method for reducing composition uncertainties is discussed.

Rozon, Daniel; Shen Wei [Institut de Genie Nucleaire (Canada)

2001-05-15T23:59:59.000Z

265

Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - III: Spent DUPIC Fuel Disposal Cost  

Science Conference Proceedings (OSTI)

The disposal costs of spent pressurized water reactor (PWR), Canada deuterium uranium (CANDU) reactor, and DUPIC fuels have been estimated based on available literature data and the engineering design of a spent CANDU fuel disposal facility by the Atomic Energy of Canada Limited. The cost estimation was carried out by the normalization concept of total electricity generation. Therefore, the future electricity generation scale was analyzed to evaluate the appropriate capacity of the high-level waste disposal facility in Korea, which is a key parameter of the disposal cost estimation. Based on the total electricity generation scale, it is concluded that the disposal unit costs for spent CANDU natural uranium, CANDU-DUPIC, and PWR fuels are 192.3, 388.5, and 696.5 $/kg heavy element, respectively.

Ko, Won Il; Choi, Hangbok; Roh, Gyuhong; Yang, Myung Seung [Korea Atomic Energy Research Institute (Korea, Republic of)

2001-05-15T23:59:59.000Z

266

Electrolytic recovery of reactor metal fuel  

DOE Patents (OSTI)

A new electrolytic process and apparatus are provided using sodium, cerium or a similar metal in alloy or within a sodium beta or beta[double prime]-alumina sodium ion conductor to electrolytically displace each of the spent fuel metals except for cesium and strontium on a selective basis from the electrolyte to an inert metal cathode. Each of the metals can be deposited separately. An electrolytic transfer of spent fuel into the electrolyte includes a sodium or cerium salt in the electrolyte with sodium or cerium alloy being deposited on the cathode during the transfer of the metals from the spent fuel. The cathode with the deposit of sodium or cerium alloy is then shunted to an anode and the reverse transfer is carried out on a selective basis with each metal being deposited separately at the cathode. The result is that the sodium or cerium needed for the process is regenerated in the first step and no additional source of these reactants is required. 2 figs.

Miller, W.E.; Tomczuk, Z.

1994-09-20T23:59:59.000Z

267

Behavior of actinides in the Integral Fast Reactor fuel cycle  

SciTech Connect

The Integral Fast Reactor (IFR) under development by Argonne National Laboratory uses metallic fuels instead of ceramics. This allows electrorefining of spent fuels and presents opportunities for recycling minor actinide elements. Four minor actinides ({sup 237}Np, {sup 240}Pu, {sup 241}Am, and {sup 243}Am) determine the waste storage requirements of spent fuel from all types of fission reactors. These nuclides behave the same as uranium and other plutonium isotopes in electrorefining, so they can be recycled back to the reactor without elaborate chemical processing. An experiment has been designed to demonstrate the effectiveness of the high-energy neutron spectra of the IFR in consuming these four nuclides and plutonium. Eighteen sets of seven actinide and five light metal targets have been selected for ten day exposure in the Experimental Breeder Reactor-2 which serves as a prototype of the IFR. Post-irradiation analyses of the exposed targets by gamma, alpha, and mass spectroscopy are used to determine nuclear reaction-rates and neutron spectra. These experimental data increase the authors` confidence in their ability to predict reaction rates in candidate IFR designs using a variety of neutron transport and diffusion programs.

Courtney, J.C. [Louisiana State Univ., Baton Rouge, LA (United States). Nuclear Science Center; Lineberry, M.J. [Argonne National Lab., Idaho Falls, ID (United States). Technology Development Div.

1994-06-01T23:59:59.000Z

268

Feasibility of breeding in hard spectrum boiling water reactors with oxide and nitride fuels  

E-Print Network (OSTI)

This study assesses the neutronic, thermal-hydraulic, and fuel performance aspects of using nitride fuel in place of oxides in Pu-based high conversion light water reactor designs. Using the higher density nitride fuel ...

Feng, Bo, Ph. D. Massachusetts Institute of Technology

2011-01-01T23:59:59.000Z

269

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel  

SciTech Connect

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

2013-03-01T23:59:59.000Z

270

Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel  

DOE Green Energy (OSTI)

The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

2011-03-01T23:59:59.000Z

271

Use of silicide fuel in the Ford Nuclear Reactor - to lengthen fuel element lifetimes  

SciTech Connect

Based on economic considerations, it has been proposed to increase the lifetime of LEU fuel elements in the Ford Nuclear Reactor by raising the {sup 235}U plate loading from 9.3 grams in aluminide (UAl{sub x}) fuel to 12.5 grams in silicide (U{sub 3}Si{sub 2}) fuel. For a representative core configuration, preliminary neutronic depletion and steady state thermal hydraulic calculations have been performed to investigate core characteristics during the transition from an all-aluminide to an all-silicide core. This paper discusses motivations for this fuel element upgrade, results from the calculations, and conclusions.

Bretscher, M.M.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Burn, R.R.; Lee, J.C. [Univ. of Michigan, Ann Arbor, MI (United States). Phoenix Memorial Lab.

1995-12-31T23:59:59.000Z

272

Prototype spent-fuel canister design, analysis, and test  

SciTech Connect

Sandia National Laboratories was asked by the US Energy Research and Development Administration (now US Department of Energy) to design the spent fuel shipping cask system for the Clinch River Breeder Reactor Plant (CRBRP). As a part of this task, a canister which holds liquid sodium and the spent fuel assembly was designed, analyzed, and tested. The canister body survived the regulatory Type-B 9.1-m (30-ft) drop test with no apparent leakage. However, the commercially available metal seal used in this design leaked after the tests. This report describes the design approach, analysis, and prototype canister testing. Recommended work for completing the design, when funding is available, is included.

Leisher, W.B.; Eakes, R.G.; Duffey, T.A.

1982-03-01T23:59:59.000Z

273

A FUEL REPROCESSING PLANT FOR FAST CERAMIC REACTORS  

SciTech Connect

A study was made of the adaptation of the HAPO anion exchange process to the reprocessing of Fast Ceramic Reactor (FCR) fuel using the Idaho Small Plant Concept. It is shown that the anion exchange flowsheet meets the reprocessing objectives of the FCR case and can be adequately accommodated in the Small Plant Concept. Capacities of up to 1550 Md(e) are feasible in the Small Plant and unit reprocessing costs range from 0.14 to 0.28 mills/kwh depending on the number of reactors to be processed. (auth)

Alter, H.W.

1962-02-01T23:59:59.000Z

274

FUEL ELEMENTS FOR THE ARGONNE ADVANCED RESEARCH REACTOR  

SciTech Connect

The core design and the fuel element concept for the high-flux Argonne Advanced Research Reactor are presented. The core is cooled and moderated by light water and utilizes beryllium as a reflector. The fuel element assembly is rhomboidal in cross section and consists of 27 plates fastened together at their edges by dovetailed locking keys, and at each end by end fittings. Each fuel plate is 40 mils thick and contains a uniform dispersion of highly enriched UO/ sub 2/ particles, up to a maximum of 37 wt%, in a matrix of sintered stainless steel powder. A 5 mil thick stainless steel cladding is metallurgically bonded to each side of the fueled matrix. (N.W.R.)

Adolph, N.R.; Silberstein, M.S.; Weinstein, A.

1962-01-01T23:59:59.000Z

275

Development and transfer of fuel fabrication and utilization technology for research reactors  

SciTech Connect

Approximately 300 research reactors supplied with US-enriched uranium are currently in operation in about 40 countries, with a variety of types, sizes, experiment capabilities and applications. Despite the usefulness and popularity of research reactors, relatively few innovations in their core design have been made in the last fifteen years. The main reason can be better understood by reviewing briefly the history of research reactor fuel technology and enrichment levels. Stringent requirements on the enrichment of the uranium to be used in research reactors were considered and a program was launched to assist research reactors in continuing their operation with the new requirements and with minimum penalties. The goal of the new program, the Reduced Enrichment Research and Test Reactor (RERTR) Program, is to develop the technical means to utilize LEU instead of HEU in research reactors without significant penalties in experiment performance, operating costs, reactor modifications, and safety characteristics. This paper reviews briefly the RERTR Program activities with special emphasis on the technology transfer aspects of interest to this conference.

Travelli, A.; Domagala, R.F.; Matos, J.E.; Snelgrove, J.L.

1982-01-01T23:59:59.000Z

276

Development of Technical Nuclear Forensics for Spent Research Reactor Fuel  

E-Print Network (OSTI)

Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct reactor parameters from a spent fuel sample using results from a radiochemical analysis. In the inverse analysis, a reactor physics code is used as a forward model. Verification and validation of different reactor physics codes was performed for usage in the inverse analysis. The verification and validation process consisted of two parts. The first is a variance analysis of Monte Carlo reactor physics burnup simulation results. The codes used in this work are MONTEBURNS and MCNPX/CINDER. Both utilize Monte Carlo transport calculations for reaction rate and flux results. Neither code has a variance analysis that will propagate through depletion steps, so a method to quantify and understand the variance propagation through these depletion calculations was developed. The second verification and validation process consisted of comparing reactor physics code output isotopic compositions to radiochemical analysis results. A sample from an Oak Ridge Research Reactor spent fuel assembly was acquired through a drilling process. This sample was then dissolved in nitric acid and diluted in three different quantities, creating three separate samples. A radiochemical analysis was completed and the results were compared to simulation outputs at different levels ofdetail. After establishing a forward model, an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict burnup, initial 235U, and 236U enrichments. The analytic calculation results may also be used stand alone or in a database search algorithm. In this work, a reactor physics code is used as a for- ward model with the analytic results as initial conditions in a numerical optimization algorithm. In the numerical analysis, the burnup and initial uranium isotopic com- positions are reconstructed until the iterative spent fuel characteristics converge with the measured data. Upon convergence of the sample’s burnup and initial uranium isotopic composition, the cooling time can be reconstructed. To reconstruct cooling time, the standard decay equation is inverted and solved for time. Two methods were developed. One method uses the converged burnup and initial uranium isotopic compositions along in a reactor depletion simulation. The second method uses an isotopic signature that does not decay out of its mass bin and has a simple production chain. An example would be 137Cs which decays into the stable 137Ba. Similar results are achieved with both methods, but extended shutdown time or time away from power results in over prediction of the cooling time. The over prediction of cooling time and comparison of different burnup reconstruction isotope results are indicator signatures of extended shutdown or time away from power. Due to dynamic operation in time and function, detailed power history reconstruction for research reactors is very challenging. Frequent variations in power, repeated variable shutdown time length, and experimentation history affect the spectrum an individual assembly is burned with such that full reactor parameter reconstruction is difficult. The results from this technical nuclear forensic analysis may be used with law enforcement, intelligence data, macroscopic and microscopic sample characteristics in a process called attribution to suggest or exclude possible sources of origin for a sample.

Sternat, Matthew 1982-

2012-12-01T23:59:59.000Z

277

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

278

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS  

DOE Patents (OSTI)

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

Loeb, E.; Nicklas, J.H.

1959-02-01T23:59:59.000Z

279

Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels  

E-Print Network (OSTI)

Minor actinides represent the long-term radiotoxicity of nuclear wastes. As one of their potential incineration options, partitioning and transmutation in fission reactors are seriously considered worldwide. If implemented, these technologies could also be a source of nuclear fuel materials required for sustainability of nuclear energy. The objective of this research was to evaluate performance characteristics of Very High Temperature Reactors (VHTRs) and their variations due to configuration adjustments targeting achievability of spectral variations. The development of realistic whole-core 3D VHTR models and their benchmarking against experimental data was an inherent part of the research effort. Although the performance analysis was primarily focused on prismatic core configurations, 3D pebble-bed core models were also created and analyzed. The whole-core 3D models representing the prismatic block and pebble-bed cores were created for use with the SCALE 5.0 code system. Each of the models required the Dancoff correction factor to be externally calculated. The code system DANCOFF-MCThe whole-core/system 3D models with multi-heterogeneity treatments were validated by the benchmark problems. Obtained results are in agreement with the available High Temperature Test Reactor data. Preliminary analyses of actinide-fueled VHTR configurations have indicated promising performance characteristics. Utilization of minor actinides as a fuel component would facilitate development of new fuel cycles and support sustainability of a fuel source for nuclear energy assuring future operation of Generation IV nuclear energy systems. was utilized to perform the Dancoff factor calculations.

Ames, David E, II

2006-08-01T23:59:59.000Z

280

Effect of reduced enrichment on the fuel cycle for research reactors  

SciTech Connect

The new fuels developed by the RERTR Program and by other international programs for application in research reactors with reduced uranium enrichment (<20% EU) are discussed. It is shown that these fuels, combined with proper fuel-element design and fuel-management strategies, can provide at least the same core residence time as high-enrichment fuels in current use, and can frequently significantly extend it. The effect of enrichment reduction on other components of the research reactor fuel cycle, such as uranium and enrichment requirements, fuel fabrication, fuel shipment, and reprocessing are also briefly discussed with their economic implications. From a systematic comparison of HEU and LEU cores for the same reference research reactor, it is concluded that the new fuels have a potential for reducing the research reactor fuel cycle costs while reducing, at the same time, the uranium enrichment of the fuel.

Travelli, A.

1982-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

Safeguards Guidance for Prismatic Fueled High Temperature Gas Reactors (HTGR)  

National Nuclear Security Administration (NNSA)

5) 5) August 2012 Guidance for High Temperature Gas Reactors (HTGRs) with Prismatic Fuel INL/CON-12-26130 Revision 0 Safeguards-by-Design: Guidance for High Temperature Gas Reactors (HTGRs) With Prismatic Fuel Philip Casey Durst (INL Consultant) August 2012 DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness, of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product,

282

Detachable connection for a nuclear reactor fuel assembly  

DOE Patents (OSTI)

A locking connection for releasably attaching a handling socket to the duct tube of a fuel assembly for a nuclear reactor. The connection comprises a load pad housing mechanically attached to the duct tube and a handling socket threadably secured within the housing. A retaining ring is interposed between the housing and the handling socket and is formed with a projection and depression engagable within a cavity and groove of the housing and handling socket, respectively, to form a detachable interlocked connection assembly.

Christiansen, D.W.; Karnesky, R.A.

1983-08-29T23:59:59.000Z

283

M-1 N-Reactor fuel development. Excerpts from N-Reactor Department Research and Development Budget for FY-1967 and Revision of Budget for FY-1966  

SciTech Connect

The use of metallic uranium fuel elements on a large scale in a pressurized water cooled graphite moderated power reactor environment in N-Reactor not only represents an extrapolation from previous Hanford fuel technology but it represents the only production-scale application of this concept in this country or abroad. Provision of a supporting technology is and will continue to be a major activity at Hanford as only a limited amount of directly applicable R&D will be available from other sites. The major benefit to be achieved through an extensive and continuing fuel program is attainment of the full capabilities of the N-Reactor at much reduced fuel cycle costs. The current N-Reactor fuel design represents the best engineering judgment of what is required for adequate performance. While the design is intentionally conservative, some features may not fully provide the level of performance required to sustain efficient reactor operation. Results of production-scale irradiation experience and special test irradiations will provide direction to a continuing program to correct any excesses or deficiencies in the initial fuel design. Reduction of unwarranted conservatism in the design will lower fuel fabrication costs, and correction of deficiencies will lower irradiation costs through increased time operated efficiency. Costs of this program are summarized for 1965, 66, and 67. The paper describes the scope of the program, its relationship to other programs, and technical progress in FY-1965 which included improved performance of the fuel elements even with the large number of rapid shutdowns of the reactor.

1965-04-19T23:59:59.000Z

284

A strategy for transition from a uranium fueled, open cycle SFR to a transuranic fueled, closed cycle sodium cooled fast reactor  

E-Print Network (OSTI)

Reactors utilizing a highly energetic neutron spectrum, often termed fast reactors, offer large fuel utilization improvements over the thermal reactors currently used for nuclear energy generation. Conventional fast reactor ...

Richard, Joshua (Joshua Glenn)

2012-01-01T23:59:59.000Z

285

External attachment of titanium sheathed thermocouples to zirconium nuclear fuel rods for the LOFT reactor  

SciTech Connect

The Exxon Nuclear Company, Inc., acting as a Subcontractor to EG and G Idaho Inc., Idaho National Engineering Laboratory, Idaho Falls, Idaho, has developed a welding process to attach titanium sheathed thermocouples to the outside of the zircaloy clad fuel rods. The fuel rods and thermocouples are used to test simulated loss-of-coolant accident (LOCA) conditions in a pressurized water reactor (LOFT Reactor, Idaho National Laboratory). A laser beam was selected as the optimum welding process because of the extremely high energy input per unit volume that can be achieved allowing local fusion of a small area irrespective of the difference in material thickness to be joined. A commercial pulsed laser and energy control system was installed along with specialized welding fixtures. Laser room facility requirements and tolerances were established. Performance qualifications, and detailed welding procedures were also developed. Product performance tests were conducted to assure that engineering design requirements could be met on a production basis.

Welty, R.K.

1980-01-01T23:59:59.000Z

286

NREL: Fleet Test and Evaluation - Alternative Fuels  

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

Alternative Fuels Alternative Fuels NREL's Fleet Test and Evaluation Team works with industry partners to evaluate the use of alternative fuels in delivery, transit, and freight vehicles. Although biodiesel is the most commonly used alternative fuel in medium- and heavy-duty diesel vehicles, compressed and liquefied natural gas and Fischer-Tropsch diesel are also viable options for trucking companies. Learn more about the team's evaluations of alternative fuels in fleet operations: Biodiesel Compressed Natural Gas Fischer-Tropsch Diesel Liquefied Natural Gas Printable Version Fleet Test and Evaluation Home Research & Development Vehicle Drive Cycle Analysis Hybrid Electric Drive Systems Electric & Plug-in Hybrid Electric Drive Systems Hydraulic Hybrid Drive Systems Truck Stop Electrification

287

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network (OSTI)

The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

Hayes, A C

2012-01-01T23:59:59.000Z

288

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network (OSTI)

The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

A. C. Hayes; Gerard Jungman

2012-05-30T23:59:59.000Z

289

SUPPLEMENT ANALYSIS OF FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL  

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

FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL TRANSPORTATION ALONG OTHER THAN~. PRESENTATIVE ROUTE FROM CONCORD NAVAL WEAPO~~ STATION TO IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LADORA TORY Introduction The Department of Energy is planning to transport foreign research reactor spent nuclear fuel by rail from the Concord Naval Weapons Station (CNWS), Concord, California, to the Idaho National Engineering and Environmental Laboratory (INEEL). The environmental analysis supporting the decision to transport, by rail or truck, foreign research reactor spent nuclear fuel from CNWS to the INEEL is contained in +he Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliftration Policy Concerning Foreign Research Reactor

290

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.

291

Effect of High Reactor Water Zinc on Fuel Performance in Quad Cities 2  

Science Conference Proceedings (OSTI)

Due to reduction in feedwater Fe, reactor water Zn concentrations have been increasing in U.S. boiling water reactors (BWRs). The fuel performance experience base is limited to 8 to 10 ppb, and no fuel surveillance was performed in a plant operated with greater than 12 ppb reactor water Zn. The impact of high reactor water Zn on fuel performance is unknown. However, the change in the trends is large enough to raise a concern, and it requires a confirmation of the fuel performance with fuel ...

2013-07-02T23:59:59.000Z

292

EA-1954: Resumption of Transient Testing of Nuclear Fuels and...  

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

Testing of Nuclear Fuels and Materials at the Idaho National Laboratory, Idaho EA-1954: Resumption of Transient Testing of Nuclear Fuels and Materials at the Idaho National...

293

Hydrogen Fuel Pilot Plant and Hydrogen ICE Vehicle Testing  

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

Fuel Pilot Plant and Hydrogen ICE Vehicle Testing Jim Francfort (INEEL) Don Karner (ETA) 2004 Fuel Cell Seminar - San Antonio Session 5B - Hydrogen DOE - Advanced Vehicle Testing...

294

A fuel for sub-critical fast reactor  

Science Conference Proceedings (OSTI)

Along with the problem of the nuclear waste transmutation, the problem of minimization of waste production is of current interest. It is not possible to eliminate production of waste at a nuclear power plant, but, as is shown in this report, it is in principle possible to arrange a fuel composition with no net production of transuranic elements. The idea is to find the transuranic elements composition to which the depleted uranium is continuously supplied during frequent reprocessing, and amount of each other transuranic fuel component remains unchanged in time. For each transuranic component, the balance is achieved by equating burnup and production rates. The production is due to neutron capture by the neighboring lighter isotope and subsequent beta-decay. The burnup includes fission, neutron capture and decays. For the calculations a simplified burnup model which accounts for 9 isotopes of uranium, neptunium, plutonium and americium is used. The calculated fuel composition consists mainly of uranium with minority of plutonium isotopes. Such a fuel, after usage in a sub-critical fast reactor, should be reprocessed. The fission product content increases during burnup, representing a net production of waste, while the transuranic elements and {sup 238}U should be recycled into a new fuel. For such a fuel cycle, the net consumption is only for 238U, and the net waste production is just fission products.

Moiseenko, V. E.; Chernitskiy, S. V.; Agren, O.; Noack, K. [Institute of Plasma Physics, National Science Center 'Kharkiv Institute of Physics and Technology', Akademichna St. 1, 61108 Kharkiv (Ukraine); Institute of Nuclear Physics, National Science Center 'Kharkiv Institute of Physics and Technology', Akademichna St. 1, 61108 Kharkiv (Ukraine); Uppsala University, Angstroem Laboratory, Division of Electricity, Box 534, SE-75121 Uppsala (Sweden)

2012-06-19T23:59:59.000Z

295

MIXED-OXIDE FUEL USE IN COMMERCIAL LIGHT WATER REACTORS  

E-Print Network (OSTI)

In a Commission briefing on high-bumup fuel on March 25, 1997, the staff said that they would prepare a white paper on mixed-oxide (MOX) fuel in anticipation of a DOE program to bum excess weapons plutonium in commercial reactors. This memorandum and its attachment comprise that paper and are provided to inform the Commissioners of technical issues associated with such a program. More recently, on February 5, 1999, I was contacted by the Nuclear Control Institute regarding a paper they have written on this subject. They presented that paper to the staff in a public meeting on April 7, 1999. The Nuclear Control Institute's written paper had been provided to the staff earlier, and we have taken the paper into consideration in preparing this memorandum. Back-ground In January 1997, the U.S. Department of Energy released a record of decision for the storage and disposition of weapons-usable fissile materials. In this record, DOE recommended that excess weapons-grade plutonium be disposed of by two methods: (1) reconstituting the plutonium into mixed-oxide (MOX) fuel rods and burning it in current light water reactors, and (2) immobilizing the plutonium in glass logs with appropriate radioactive isotopes to deter theft prior to geologic disposal. Based on current information, it now appears that, if the MOX fuel method is utilized, fuel fabrication will take place at the Savannah River site in South Carolina with burning in nearby Westinghouse-type PWRs. Although DOE will probably not receive funding in FY 2000 for developing a license application, Congress has already given its approval for NRC licensing authority over a MOX fuel fabrication facility operated under

United States; William D. Travers

1999-01-01T23:59:59.000Z

296

Nuclear breeder reactor fuel element with axial tandem stacking and getter  

DOE Patents (OSTI)

A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

Gibby, Ronald L. (Richland, WA); Lawrence, Leo A. (Kennewick, WA); Woodley, Robert E. (Richland, WA); Wilson, Charles N. (Richland, WA); Weber, Edward T. (Kennewick, WA); Johnson, Carl E. (Elk Grove, IL)

1981-01-01T23:59:59.000Z

297

Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities  

E-Print Network (OSTI)

Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with nuclear weapons programs, they are capable of being used in civilian power programs. The simpler design constraints associated with this type of reactor would make them ideal for developing nations to bolster their electricity generation and help promote a greater standard of living in those nations. However, the same benefits that make this type of reactor desirable also make it suspicious to the international community as a possible means to shorten that state?s nuclear latency. If a safeguards approach could be developed for a fuel cycle featuring one of these reactors, it would ease the tension surrounding their existence and possibly lead to an increased latency through engineered barriers. The development of this safeguards approach follows a six step procedure. First, the fuel cycle was analyzed for the types of facilities found in it and how nuclear material flows between facilities. The goals of the safeguards system were established next, using the normal IAEA standards for the non-detection and false alarm probabilities. The 5 MWe Reactor was modeled for both plutonium production and maximum power capacity. Each facility was analyzed for material throughput and the processes that occur in each facility were researched. Through those processes, diversion pathways were developed to test the proposed safeguards system. Finally, each facility was divided into material balance areas and a traditional nuclear material accountancy system was set up to meet the established safeguards goals for the facility. The DPRK weapons program is a great example of the type of fuel cycle that is the problem. The three major facilities in the fuel cycle, the Fuel Fabrication Facility, the 5 MWe Reactor, and the Radiochemical Laboratory, can achieve the two goals of safeguards using traditional methods. Each facility can be adequately safeguarded using methods and practices that are relatively inexpensive and can obtain material balance periods close to the timeliness limits set forth by the IAEA. The Fuel Fabrication Facility can be safeguarded at both its current needed capacity and its full design capacity using inexpensive measurements. The material balance period needed for both capacities are reasonable. For the 5 MWe reactor, plutonium production is simulated to be 6.7 kg per year and is on the high side of estimates. The Radiochemical Laboratory can also be safeguarded at its current capacity. In fact, the timeliness goal for the facility dictates what the material balance period must be for the chosen set of detectors which make it very reasonable.

Rauch, Eric B.

2009-05-01T23:59:59.000Z

298

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

299

Full-length high-temperature severe fuel damage test No. 1  

SciTech Connect

This report describes the first full-length high-temperature test (FLHT-1) performed by Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. The test is part of a series of experiments being performed for the NRC as a part of their Severe Fuel Damage Program and is one of several planned for PNL`s Coolant Boilaway and Damage Progression Program. The report summarizes the test design and test plan. it also provides a summary and discussion of the data collected during the test and of the photos taken during the post-test examination. All objectives for the test were met. The key objective was to demonstrate that severe fuel damage tests on full-length fuel bundles can be safely conducted in the NRU reactor.

Rausch, W.N.; Hesson, G.M.; Pilger, J.P.; King, L.L.; Goodman, R.L.; Panisko, F.E.

1993-08-01T23:59:59.000Z

300

The second and third NGNP advanced gas reactor fuel irradiation experiments  

SciTech Connect

The United States Dept. of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is currently scheduled to irradiate a total of five low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The irradiations are being accomplished to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas cooled reactors. The experiments will each consist of at least six separate capsules, and will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The effluent sweep gas will also have on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The second experiment (AGR-2) started irradiation in June 2010, and the third and fourth experiments have been combined into a single larger irradiation (AGR-3/4) that is currently being assembled. The design and status of the second through fourth experiments as well as the irradiation results of the second experiment to date are discussed. (authors)

Grover, S. B.; Petti, D. A. [Idaho National Laboratory, 2525 N. Fremont Ave., Idaho Falls, ID 83415 (United States)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

SciTech Connect

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-01-01T23:59:59.000Z

302

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

Science Conference Proceedings (OSTI)

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-12-31T23:59:59.000Z

303

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells  

DOE Green Energy (OSTI)

This topical report summarizes work accomplished for the Program from November 1, 2001 to December 31, 2002 in the following task areas: Task 1: Materials Development; Task 2: Composite Development; Task 4: Reactor Design and Process Optimization; Task 8: Fuels and Engine Testing; 8.1 International Diesel Engine Program; 8.2 Nuvera Fuel Cell Program; and Task 10: Program Management. Major progress has been made towards developing high temperature, high performance, robust, oxygen transport elements. In addition, a novel reactor design has been proposed that co-produces hydrogen, lowers cost and improves system operability. Fuel and engine testing is progressing well, but was delayed somewhat due to the hiatus in program funding in 2002. The Nuvera fuel cell portion of the program was completed on schedule and delivered promising results regarding low emission fuels for transportation fuel cells. The evaluation of ultra-clean diesel fuels continues in single cylinder (SCTE) and multiple cylinder (MCTE) test rigs at International Truck and Engine. FT diesel and a BP oxygenate showed significant emissions reductions in comparison to baseline petroleum diesel fuels. Overall through the end of 2002 the program remains under budget, but behind schedule in some areas.

E.T. (Skip) Robinson; James P. Meagher; Prasad Apte; Xingun Gui; Tytus R. Bulicz; Siv Aasland; Charles Besecker; Jack Chen Bart A. van Hassel; Olga Polevaya; Rafey Khan; Piyush Pilaniwalla

2002-12-31T23:59:59.000Z

304

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011  

SciTech Connect

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

2012-03-01T23:59:59.000Z

305

EA-1954: Resumption of Transient Testing of Nuclear Fuels and Materials at  

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

4: Resumption of Transient Testing of Nuclear Fuels and 4: Resumption of Transient Testing of Nuclear Fuels and Materials at the Idaho National Laboratory, Idaho EA-1954: Resumption of Transient Testing of Nuclear Fuels and Materials at the Idaho National Laboratory, Idaho SUMMARY This Environmental Assessment (EA) evaluates U.S. Department of Energy (DOE) activities associated with its proposal to resume testing of nuclear fuels and materials under transient high-power test conditions at the Transient Reactor Test (TREAT) Facility at the Idaho National Laboratory. The State of Idaho and Shoshone-Bannock Tribes are cooperating agencies. PUBLIC COMMENT OPPORTUNITIES DOE invites the public to read and comment on a draft environmental assessment it has prepared for a proposal to resume transient testing of nuclear fuels and materials at either Idaho National Laboratory or Sandia

306

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 ...

307

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

308

Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel  

Science Conference Proceedings (OSTI)

Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration /Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

Primm, Trent [ORNL; Guida, Tracey [University of Pittsburgh

2010-02-01T23:59:59.000Z

309

Nuclear processes in magnetic fusion reactors with polarized fuel  

E-Print Network (OSTI)

We consider the processes $d +d \\to n +{^3He}$, $d +{^3He} \\to p +{^4He}$, $d +{^3H} \\to n +{^4He}$, ${^3He} +{^3He}\\to p+p +{^4He}$, ${^3H} +{^3He}\\to d +{^4He}$, with particular attention for applications in fusion reactors. After a model independent parametrization of the spin structure of the matrix elements for these processes at thermal colliding energies, in terms of partial amplitudes, we study polarization phenomena in the framework of a formalism of helicity amplitudes. The strong angular dependence of the final nuclei and of the polarization observables on the polarizations of the fuel components can be helpful in the design of the reactor shielding, blanket arrangement etc..We analyze also the angular dependence of the neutron polarization for the processes $\\vec d +\\vec d \\to n +{^3He}$ and $\\vec d +\\vec {^3H} \\to n +{^4He}$.

Rekalo, M P; Rekalo, Michail P.; Tomasi-Gustafsson, Egle

2000-01-01T23:59:59.000Z

310

Dr. Hussein Khalil at Reactor and Fuel Cycle Technologies Subcommittee  

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

Blue Blue ribbon presentation by Dr. Hussein Khalil Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library About Nuclear Energy Nuclear Reactors Designed by Argonne Argonne's Nuclear Science and Technology Legacy Opportunities within NE Division Visit Argonne Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Celebrating the 70th Anniversary of Chicago Pile 1 (CP-1) Argonne OutLoud on Nuclear Energy Argonne Energy Showcase 2012 Highlights Bookmark and Share Blue ribbon presentation by Hussein Khalil Hussein Khalil Dr. Hussein Khalil during the panel discussion Oct. 21, 2010 On October 12 Hussein Khalil, director of Argonne's Nuclear Engineering Division, participated in a Reactor and Fuel Cycle Technologies

311

Nuclear processes in magnetic fusion reactors with polarized fuel  

E-Print Network (OSTI)

We consider the processes $d +d \\to n +{^3He}$, $d +{^3He} \\to p +{^4He}$, $d +{^3H} \\to n +{^4He}$, ${^3He} +{^3He}\\to p+p +{^4He}$, ${^3H} +{^3He}\\to d +{^4He}$, with particular attention for applications in fusion reactors. After a model independent parametrization of the spin structure of the matrix elements for these processes at thermal colliding energies, in terms of partial amplitudes, we study polarization phenomena in the framework of a formalism of helicity amplitudes. The strong angular dependence of the final nuclei and of the polarization observables on the polarizations of the fuel components can be helpful in the design of the reactor shielding, blanket arrangement etc..We analyze also the angular dependence of the neutron polarization for the processes $\\vec d +\\vec d \\to n +{^3He}$ and $\\vec d +\\vec {^3H} \\to n +{^4He}$.

Michail P. Rekalo; Egle Tomasi-Gustafsson

2000-10-16T23:59:59.000Z

312

Investigation of Freeze-Cast Scaffolds as an Advanced Reactor Fuel ...  

Science Conference Proceedings (OSTI)

Symposium, Materials and Fuels for the Current and Advanced Nuclear Reactors II ... A Rate-Theory Approach to Irradiation Damage Modeling with Random ...

313

The OSU Hydro-Mechanical Fuel Test Facility: Standard Fuel Element Testing  

Science Conference Proceedings (OSTI)

Oregon State University (OSU) and the Idaho National Laboratory (INL) are currently collaborating on a test program which entails hydro-mechanical testing of a generic plate type fuel element, or standard fuel element (SFE), for the purpose of qualitatively demonstrating mechanical integrity of uranium-molybdenum monolithic plates as compared to that of uranium aluminum dispersion, and aluminum fuel plates due to hydraulic forces. This test program supports ongoing work conducted for/by the fuel development program and will take place at OSU in the Hydro-Mechanical Fuel Test Facility (HMFTF). Discussion of a preliminary test matrix, SFE design, measurement and instrumentation techniques, and facility description are detailed in this paper.

Wade R. Marcum; Brian G. Woods; Ann Marie Phillips; Richard G. Ambrosek; James D. Wiest; Daniel M. Wachs

2001-10-01T23:59:59.000Z

314

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

315

Fueling Robot Automates Hydrogen Hose Reliability Testing (Fact Sheet)  

SciTech Connect

Automated robot mimics fueling action to test hydrogen hoses for durability in real-world conditions.

Harrison, K.

2014-01-01T23:59:59.000Z

316

Future of Hydrogen Fuel Flows Through New NIST Test ...  

Science Conference Proceedings (OSTI)

Future of Hydrogen Fuel Flows Through New NIST Test Facility. For Immediate Release: February 16, 2010. ...

2012-10-15T23:59:59.000Z

317

THORIUM BREEDER REACTOR EVALUATION. PART 1. FUEL YIELD AND FUEL CYCLE COSTS IN FIVE THERMAL BREEDERS  

SciTech Connect

The performances of aqueous-homogeneous (AHBR), molten-salt (MSBR), liquid-bismuth (LBBR), gas-cooled graphite-moderated (GGBR), and deuterium- moderated gas-cooled (DGBR) breeder reactors were evaluated in respect to fuel yield, fuel cycle costs, and development status. A net electrical plant capability of 1000 Mwe was selected, and the fuel and fertile streams were processed continuously on-site. The maximum annual fuel yields were 1.5 mills/ kwhr. The minimum estimated fuel cycle costs were 0.9, 0.6, 1.0, 1.2, and 1.3 mills/kwhr at fuel yields of were 0.9, 0.9, 1.5, 1.5, and 1.3 mills/kwhr. Only the AHBR and the MSBR are capable of achieving fuel yields substantially in excess of 4%/yr, and therefore, in view of the uncertainties in nuclear data and efficiencies of processing methods, only these two can be listed with confidence as being able to satisfy the main criterion of the AEC longrange thorium breeder program, viz. a doubling time of 25 years or less. The development effort required to bring the various concepts to the stage where a prototype station could be designed was estimated to be least for the AHBR, somewhat more for the MSBR, and several times as much for the other systems. The AHBR was judged to rank first in regard to nuclear capability, fuel cycle potential, and status of development. (auth)

Alexander, L.G.; Carter, W.L.; Chapman, R.H.; Kinyon, R.W.; Miller, J.W.; Van Winkle, R.

1961-05-24T23:59:59.000Z

318

INVENTORY AND DESCRIPTION OF COMMERCIAL REACTOR FUELS WITHIN THE UNITED STATES  

Science Conference Proceedings (OSTI)

There are currently 104 nuclear reactors in 31 states, operated by 51 different utilities. Operation of these reactors generates used fuel assemblies that require storage prior to final disposition. The regulatory framework within the United States (U.S.) allows for the licensing of used nuclear fuel storage facilities for an initial licensing period of up to 40 years with potential for license extensions in 40 years increments. Extended storage, for periods of up to 300 years, is being considered within the U.S. Therefore, there is an emerging need to develop the technical bases to support the licensing for long-term storage. In support of the Research and Development (R&D) activities required to support the technical bases, a comprehensive assessment of the current inventory of used nuclear fuel based upon publicly available resources has been completed that includes the most current projections of used fuel discharges from operating reactors. Negotiations with the nuclear power industry are ongoing concerning the willingness of individual utilities to provide information and material needed to complete the R&D activities required to develop the technical bases for used fuel storage for up to 300 years. This report includes a status of negotiations between DOE and industry in these regards. These negotiations are expected to result in a framework for cooperation between the Department and industry in which industry will provide and specific information on used fuel inventory and the Department will compensate industry for the material required for Research and Development and Testing and Evaluation Facility activities.

Vinson, D.

2011-03-31T23:59:59.000Z

319

Reactor fuel conversion assistance request. Technical progress report, August 15, 1992--May 14, 1993  

SciTech Connect

This report is a summary of the progress that has been made on the preparations required to convert the WSU TRIGA reactor from High Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel.

Tripard, G.E.

1993-06-01T23:59:59.000Z

320

Reactor fuel conversion assistance request: Technical progress report, August 15, 1992-December 31, 1994  

SciTech Connect

This report is a summary of the progress that has been made on the preparations required to convert the WSU TRIGA reactor from High Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel.

Tripard, G.E.

1994-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

An integrated performance model for high temperature gas cooled reactor coated particle fuel  

E-Print Network (OSTI)

The performance of coated fuel particles is essential for the development and deployment of High Temperature Gas Reactor (HTGR) systems for future power generation. Fuel performance modeling is indispensable for understanding ...

Wang, Jing, 1976-

2004-01-01T23:59:59.000Z

322

Conceptual design of an annular-fueled superheat boiling water reactor  

E-Print Network (OSTI)

The conceptual design of an annular-fueled superheat boiling water reactor (ASBWR) is outlined. The proposed design, ASBWR, combines the boiler and superheater regions into one fuel assembly. This ensures good neutron ...

Ko, Yu-Chih, Ph. D. Massachusetts Institute of Technology

2011-01-01T23:59:59.000Z

323

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

324

Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests  

SciTech Connect

The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85{degree}C and 25{degree}C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs.

Wilson, C.N.

1990-06-01T23:59:59.000Z

325

The effect of fuel type in unsaturated spent fuel tests  

Science Conference Proceedings (OSTI)

Two well-characterized types of spent nuclear fuel (ATM-103 and ATM-106) were tested under simulated unsaturated conditions with simulated groundwater at 90{degree}C. The actinides present in the leachate were measured after periods of approximately 60, 120, and 275 days. The vessels were acid stripped after 120 and 275 days. Both colloidal and soluble actinide species were detected in the leachates which had pHs ranging from 4 to 7. Alpha spectroscopy studies of filtered and unfiltered leachates showed that large amounts of actinides may be bound in colloids. The uranium phases identified in the colloids were schoepite and soddyite. The actinide release behavior of the two fuels appears to be different. The ATM-106 fuel began to release actinides later than the ATM-103 fuel, but after 275 days, it had released more. The amount of americium released from the two fuels was a higher percentage of the maximum amount of americium present than was the percentage of the simultaneous amount of uranium released.

Finn, P.A.; Gong, M.; Bates, J.K.; Emery, J.W.; Hoh, J.C.

1994-04-01T23:59:59.000Z

326

Effects of fabrication and irradiation on the dissolution of (U,Pu)O$sub 2$ reactor fuels  

SciTech Connect

From American Ceramics Society nuclear division meeting; San Francisco, California, USA (29 Oct 1973). LMFBR-type reactors will be fueled with stainless- steel-clad MFBR fuel cycle are the recovery of uranium and plutonium and the refabrication of the fuel elements in the minimum practicable time at lowest cost. Effect of fabrication method and irradiation conditions on recovery of the fuel is discussed. The Purex process is used to prepare the feed solutions. Test specimens contained fuels derived from sol-gel, coprecipitated, and mechanically blended oxides. Irradiation levels varied from unirradiated to 100,000 MWd/ton. Solubility of the fuels in terms of the fabrication method is coprecipitated> sol- gel > mechanically blended. Irradiation tends to increase the fuel solubility. (LK)

Goode, J.H.; Fitzgerald, C.L.; Vaughen, V.C.A.

1973-01-01T23:59:59.000Z

327

Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report  

SciTech Connect

The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

William Anderson; James Tulenko; Bradley Rearden; Gary Harms

2008-09-11T23:59:59.000Z

328

An integrated approach for the verification of fresh mixed oxide fuel (MOX) assemblies at light water reactor MOX recycle reactors  

Science Conference Proceedings (OSTI)

This paper presents an integrated approach for the verification of mixed oxide (MOX) fuel assemblies prior to their being loaded into the reactor. There is a coupling of the verification approach that starts at the fuel fabrication plant and stops with the transfer of the assemblies into the thermal reactor. The key measurement points are at the output of the fuel fabrication plant, the receipt at the reactor site, and the storage in the water pool as fresh fuel. The IAEA currently has the capability to measure the MOX fuel assemblies at the output of the fuel fabrication plants using a passive neutron coincidence counting systems of the passive neutron collar (PNCL) type. Also. at the MOX reactor pool, the underwater coincidence counter (UWCC) has been developed to measure the MOX assemblies in the water. The UWCC measurement requires that the fuel assembly be lifted about two meters up in the storage rack to avoid interference from the fuel that is stored in the rack. This paper presents a new method to verify the MOX fuel assemblies that are in the storage rack without the necessity of moving the fuel. The detector system is called the Underwater MOX Verification System (UMVS). The integration and relationship of the three measurements systems is described.

Menlove, Howard O [Los Alamos National Laboratory; Lee, Sang - Yoon [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

329

Axially staggered seed-blanket reactor fuel module construction  

DOE Patents (OSTI)

A heterogeneous nuclear reactor of the seed-blanket type is provided wher the fissile (seed) and fertile (blanket) nuclear fuels are segregated axially within each fuel element such that fissile and fertile regions occur in an alternating pattern along the length of the fuel element. Further, different axial stacking patterns are used for the fuel elements of at least two module types such that when modules of different types are positioned adjacent to one another, the fertile regions of the modules are offset or staggered. Thus, when a module of one type is surrounded by modules of the second type the fertile regions thereof will be surrounded on all sides by fissile material. This provides enhanced neutron communication both radially and axially, thereby resulting in greater power oscillation stability than other axial arrangements. The arrangements of the fissile and fertile regions in an alternating axial manner minimizes the radial power peaking factors and provides a more optional thermal-hydraulic design than is afforded by radial arrangements.

Cowell, Gary K. (Monroeville, PA); DiGuiseppe, Carl P. (West Mifflin, PA)

1985-01-01T23:59:59.000Z

330

Economic Analyiss of "Symbiotic" Light Water Reactor/Fast Burner Reactor Fuel Cycles Proposed as Part of the U.S. Advanced Fuel Cycle Initiative (AFCI)  

Science Conference Proceedings (OSTI)

A spreadsheet-based 'static equilibrium' economic analysis was performed for three nuclear fuel cycle scenarios, each designed for 100 GWe-years of electrical generation annually: (1) a 'once-through' fuel cycle based on 100% LWRs fueled by standard UO2 fuel assemblies with all used fuel destined for geologic repository emplacement, (2) a 'single-tier recycle' scenario involving multiple fast burner reactors (37% of generation) accepting actinides (Pu,Np,Am,Cm) from the reprocessing of used fuel from the uranium-fueled LWR fleet (63% of generation), and (3) a 'two-tier' 'thermal+fast' recycle scenario where co-extracted U,Pu from the reprocessing of used fuel from the uranium-fueled part of the LWR fleet (66% of generation) is recycled once as full-core LWR MOX fuel (8% of generation), with the LWR MOX used fuel being reprocessed and all actinide products from both UO2 and MOX used fuel reprocessing being introduced into the closed fast burner reactor (26% of generation) fuel cycle. The latter two 'closed' fuel cycles, which involve symbiotic use of both thermal and fast reactors, have the advantages of lower natural uranium requirements per kilowatt-hour generated and less geologic repository space per kilowatt-hour as compared to the 'once-through' cycle. The overall fuel cycle cost in terms of $ per megawatt-hr of generation, however, for the closed cycles is 15% (single tier) to 29% (two-tier) higher than for the once-through cycle, based on 'expected values' from an uncertainty analysis using triangular distributions for the unit costs for each required step of the fuel cycle. (The fuel cycle cost does not include the levelized reactor life cycle costs.) Since fuel cycle costs are a relatively small percentage (10 to 20%) of the overall busbar cost (LUEC or 'levelized unit electricity cost') of nuclear power generation, this fuel cycle cost increase should not have a highly deleterious effect on the competitiveness of nuclear power. If the reactor life cycle costs are included in the analysis, with the fast reactors having a higher $/kw(e) capital cost than the LWRs, the overall busbar generation cost ($/MWh) for the closed cycles is approximately 12% higher than for the all-LWR once-through fuel cycle case, again based on the expected values from an uncertainty analysis. It should be noted that such a percentage increase in the cost of nuclear power is much smaller than that expected for fossil fuel electricity generation if CO2 is costed via a carbon tax, cap and trade regimes, or carbon capture and sequestration (CCS).

Williams, Kent Alan [ORNL; Shropshire, David E. [Idaho National Laboratory (INL)

2009-01-01T23:59:59.000Z

331

Tests of prototype salt stripper system for IFR fuel cycle  

Science Conference Proceedings (OSTI)

One of the waste treatment steps for the on-site reprocessing of spent fuel from the Integral Fast Reactor fuel cycles is stripping of the electrolyte salt used in the electrorefining process. This involves the chemical reduction of the actinides and rare earth chlorides forming metals which then dissolve in a cadmium pool. To develop the equipment for this step, a prototype salt stripper system has been installed in an engineering scale argon-filled glovebox. Pumping trails were successful in transferring 90 kg of LiCl-KCl salt containing uranium and rare earth metal chlorides at 500{degree}C from an electrorefiner to the stripper vessel at a pumping rate of about 5 L/min. The freeze seal solder connectors which were used to join sections of the pump and transfer line performed well. Stripping tests have commenced employing an inverted cup charging device to introduce a Cd-15 wt % Li alloy reductant to the stripper vessel.

Carls, E.L.; Blaskovitz, R.J.; Johnson, T.R. [Argonne National Lab., IL (United States); Ogata, T. [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

1993-09-01T23:59:59.000Z

332

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

333

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

334

A compact breed and burn fast reactor using spent nuclear fuel blanket  

Science Conference Proceedings (OSTI)

A long-life breed-and-burn (B and B) type fast reactor has been investigated from the neutronics points of view. The B and B reactor has the capability to breed the fissile fuels and use the bred fuel in situ in the same reactor. In this work, feasibility of a compact sodium-cooled B and B fast reactor using spent nuclear fuel as blanket material has been studied. In order to derive a compact B and B fast reactor, a tight fuel lattice and relatively large fuel pin are used to achieve high fuel volume fraction. The core is initially loaded with an LEU (Low Enriched Uranium) fuel and a metallic fuel is used in the core. The Monte Carlo depletion has been performed for the core to see the long-term behavior of the B and B reactor. Several important parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, fission power, and fast neutron fluence, are analyzed through Monte Carlo reactor analysis. Evolution of the core fuel composition is also analyzed as a function of burnup. Although the long-life small B and B fast reactor is found to be feasible from the neutronics point of view, it is characterized to have several challenging technical issues including a very high fast neutron fluence of the structural materials. (authors)

Hartanto, D.; Kim, Y. [Korea Advanced Inst. of Science and Technology KAIST, 291 Daehak-ro, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2012-07-01T23:59:59.000Z

335

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

336

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

337

Fuel Cell Development and Test Laboratory (Fact Sheet), NREL...  

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

NREL's state-of-the-art Fuel Cell Development and Test Laboratory in the Energy Systems Integration Facility (ESIF) supports NREL's fuel cell research and development...

338

Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation  

E-Print Network (OSTI)

Wood, Completely automated nuclear reactors for long- termhandbook: Nuclear physics and reactor theory. 1993. (121 p).The analysis of any nuclear reactor system is closely

Heidet, Florent

2010-01-01T23:59:59.000Z

339

Evaluation of the advanced mixed-oxide fuel test FO-2 irradiated in the FFTF (Fast Flux Test Facility)  

SciTech Connect

The advanced mixed-oxide (UO{sub 2}-PuO{sub 2}) test assembly, FO-2, irradiated in the Fast Flux Test Facility (FFTF) is undergoing postirradiation examination. This is one of the first FFTF tests examined that used the advanced ferrite-martensite alloy, HT9, which is highly resistant to irradiation swelling. The FO-2 includes the first annular fueled pins irradiated in FFTF to undergo destructive examination. The FO-2 is a lead assembly for the ongoing FFTF Core Demonstration Experiment (CDE) and was designed to evaluate the effects of fuel design variables, such as pellet density, smeared density, and fuel form (annular or solid fuel), on advanced pin performance. The assembly contains a total of 169 fuel pins of 12 different types. Two L (annular) fuel pins, GF02L04 (FFTF and transient tested) and GF02L09 (FFTF only), were destructively examined. Evaluation of the FO-2 fuel pins and assembly shows the excellent and predictable performance of the mixed-oxide fuels with HT9 structural material. This, combined with the robust behavior of the pins in transient tests, and the continued excellent performance of the CDE indicate this is a superior fuel system for liquid-metal reactors. It offers greatly reduced deformation during irradiation, while maintaining good operating characteristics.

Burley Gilpin, L.L.; Chastain, S.A.; Baker, R.B.

1989-01-01T23:59:59.000Z

340

Development of OTM Syngas Process and Testing of Syngas Derived Ultra-clean Fuels in Diesel Engines and Fuel Cells  

DOE Green Energy (OSTI)

This final report summarizes work accomplished in the Program from January 1, 2001 through December 31, 2004. Most of the key technical objectives for this program were achieved. A breakthrough material system has lead to the development of an OTM (oxygen transport membrane) compact planar reactor design capable of producing either syngas or hydrogen. The planar reactor shows significant advantages in thermal efficiency and a step change reduction in costs compared to either autothermal reforming or steam methane reforming with CO{sub 2} recovery. Syngas derived ultra-clean transportation fuels were tested in the Nuvera fuel cell modular pressurized reactor and in International Truck and Engine single cylinder test engines. The studies compared emission and engine performance of conventional base fuels to various formulations of ultra-clean gasoline or diesel fuels. A proprietary BP oxygenate showed significant advantage in both applications for reducing emissions with minimal impact on performance. In addition, a study to evaluate new fuel formulations for an HCCI engine was completed.

E.T. Robinson; John Sirman; Prasad Apte; Xingun Gui; Tytus R. Bulicz; Dan Corgard; John Hemmings

2005-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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.


341

Cadmium Depletion Impacts on Hardening Neutron6 Spectrum for Advanced Fuel Testing in ATR  

SciTech Connect

For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim Effect in the test region.

Gray S. Chang

2011-05-01T23:59:59.000Z

342

Fuel qualification issues and strategies for reactor-based surplus plutonium disposition  

SciTech Connect

The Department of Energy (DOE) has proposed irradiation of mixed-oxide (MOX) fuel in existing commercial reactors as a disposition method for surplus plutonium from the weapons program. The burning of MOX fuel in reactors is supported by an extensive technology base; however, the infrastructure required to implement reactor-based plutonium disposition does not exist domestically. This report identifies and examines the actions required to qualify and license weapons-grade (WG) plutonium-based MOX fuels for use in domestic commercial light-water reactors (LWRs).

Cowell, B.S.; Copeland, G.L.; Moses, D.L.

1997-08-01T23:59:59.000Z

343

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

344

Conversion and standardization of US university reactor fuels using LEU, status 1989  

SciTech Connect

In 1986, the US Department of Energy initiated a program to change the fuel used in most of the US university research reactors using HEU (93%) to LEU({lt}20{percent}) in order to minimize the risk of theft or diversion of this weapons-useable material. An important consideration in the LEU conversion planning process has been the desire to standardize the fuels that are used and to enhance the performance and utilization of the reactors. This paper describes the current status of this conversion process and the plans and schedules to complete an orderly transition from HEU to LEU fuel in most of these reactors. To date, three university reactors have been converted to LEU fuel, completed safety documentation for three reactors is being evaluated by the USNRC, and work on the safety documentation for six reactors is in progress. 13 refs., 9 tabs.

Brown, K.R.; Matos, J.E. (EG and G Idaho, Inc., Idaho Falls, ID (USA); Argonne National Lab., IL (USA))

1989-01-01T23:59:59.000Z

345

Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under  

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

Assembly Shaker Test for Determining Loads on a PWR Assembly Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 Fuel Assembly Shaker Test for Determining Loads on a PWR Assembly under Surrogate Normal Conditions of Truck Transport R0.1 The United States current approach of long-term storage at its nuclear power plants and independent spent fuel storage installation, and deferred transportation of used nuclear fuel (UNF), along with the trend of nuclear power plants using reactor fuel for a longer time, creates questions concerning the ability of this aged, high-burnup fuel to withstand stresses and strains seen during normal conditions of transport from its current location to a future consolidated storage facility or permanent repository. UNFD R&D conducted testing employing surrogate instrumented

346

Failed fuel monitoring and surveillance techniques for liquid metal cooled fast reactors  

Science Conference Proceedings (OSTI)

The Experimental Breeder Reactor II (EBR-II) has been used as a facility for irradiation of LMR fuels and components for thirty years. During this time many tests of experimental fuel were continued to cladding breach in order to study modes of element failure; the methods used to identify such failures are described in a parallel paper. This paper summarizes experience of monitoring the delayed-neutron (DN) and fission-gas (FG) release behavior of a smaller number of elements that continued operation in the run-beyond-cladding-breach (RBCB) mode. The scope of RBCB testing, the methods developed to characterize failures on-line, and examples of DN/FG behavior are described.

Lambert, J.D.B.; Mikaili, R.; Gross, K.C.; Strain, R.V. [Argonne National Lab., IL (United States); Aoyama, T.; Ukai, S.; Nomura, S.; Nakae, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan)

1995-05-01T23:59:59.000Z

347

Radiation Shielding Analysis for Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors (DUPIC)  

Science Conference Proceedings (OSTI)

As a part of the compatibility analysis of DUPIC fuel in Canada deuterium uranium (CANDU) reactors, the radiation physics calculations have been performed for the CANDU primary shielding system, which was originally designed for natural uranium core. At first, the conventional CANDU primary shield analysis method was validated using the Monte Carlo code MCNP-4B in order to assess the current analysis code system and the cross-section data. The computational benchmark calculation was performed for the CANDU end shield system, which has shown that the conventional method produces results consistent with the reference calculations as far as the total dose rate and total heat deposition rate are concerned. Second, the primary shield system analysis was performed for the DUPIC fuel core based on the power distribution obtained from the time-average core model, and the results have shown that the dose rates and heat deposition rates through the primary shield of the DUPIC fuel core are not much different from those of the natural uranium core because the power levels on the core periphery are similar for both cores. This study has shown that the current primary shield system is adaptable for the DUPIC fuel CANDU core without design modification.

Roh, Gyuhong; Choi, Hangbok [Korea Atomic Energy Research Institute (Korea, Republic of)

2004-06-15T23:59:59.000Z

348

DISPERSIONS OF URANIUM CARBIDES IN ALUMINUM PLATE-TYPE RESEARCH REACTOR FUEL ELEMENTS  

DOE Green Energy (OSTI)

The technical feasibility of employing uranium carbide aluminun dispersions in aluminum-base research reactor fuel elements was investigated This study was motivated by the need to obtain higher uranium loadings in these fuel elements. Although toe MTR-type unit, containing a 13 18 wt% U-Al alloy is a proven reactor component, fabrication problems of considerable magnitude arise when attempts are made to increase the uranium investment in the alloy to more than 25 wt.%. Au approach to these fabrication difficulties is to select a compound with significantly higher density tban UAl/sub 4/ or UAl/sub 3/ compounds of the alloy system which when dispersed in aluminum powder, will reduce the volume occupied by the brittle, fissile phase. The uranium carbides, with densities ranging from 11.68 to 13.63 g/cm/sup 3/), appear to be suited for this application and were selected for development as a fuel material for aluminum-base dispersions. Studies were conducted at 580 to 620 deg C to determine the chemical compatibility of carbides with aluminum in sub-size cold- pressed comparts as well as in full-size fabricated fuel plates. Procedures were also developed to prepare uranium carbides, homogernously disperse the compounds in aluminum, roll clad the dispersions to form composite plates, and braze the plates into fuel assemblies. Corrosion tests of the fuel material were conducted in 20 and 60 deg C water to determine the integrity of the fuel material in the event of sin inadventent cladding failure. In addition, specimens were prepared to evaluate penformance under extensive irradiation Prior to studying the uranium carbide-aluminum system, methods for preparing the carbides were investigated. Are melting uranium and carnon was satisfactory for obtaining small quantities of various carbides. Later, reaction of graphite with UO/sub 2/ was successfully employed in the preparation of large quantities of UC/sub 2/, Studies of the chemical compatibility of cold-pressed compacts containing 50 wt% uranium carbide dispersed in aluminum revealed a marked trend toward stebifity as the carbon content of the uranium carbide increased from 446 to 9.20% C. Severe volume increases occurred in monocarbide dispersions with attendant formation of large quantities of the uranium-allumnim inter-metallic compounds. Dicarbide dispersions, on the other band, exhibited negligible reaction with aluminum after extended periods at 580 and 620 deg C. However, it was demonstrated that hydrogen can promote a reaction in UC/sub 2/-Al compacts. The hydrogen appears to reduce the UC/sub 2/ to UC which can subsequently react with aluminum producing the previously noted deleterious effects. A growth study at 605 deg C of composite fuel plates containing 59 wt.% UC/sub 2/ revealed insignificant changes within processing periods envisioned for fuel element processing. However, plate elongations as high as 2.5% were observed after 100 hr at this temperature. Severe blistering which occurred on fuel plates fabricated in the initial stages of the investigation was attributed to gaseous hydrocarbons, and the condition was ellminated by vacuum degasification of cold-pressed compacts. With the exception of the degasification requirement, procedures for manufacturing UC- bearing fuel elements were identical to those specified for the Geneva Conference Reactor fuel elements. Dispersions of uranium dicarbide corroded catastrophically in 20 and 60 deg C water, thus limiting the application of this material However, spocimens were prepared and insented in the MTR to evaluate the irradiation behavior of this fuel because of its potential application in onganic- cooled reactors. (auth)

Thurber, W.C.; Beaver, R.J.

1959-11-19T23:59:59.000Z

349

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

350

Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration  

SciTech Connect

The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

Benedict, R.W.; Goff, K.M.

1993-01-01T23:59:59.000Z

351

Advanced Vehicle Testing Activity: Alternative Fuel Vehicles  

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

Alternative Fuel Vehicles SuperShuttle CNG Van Alternative fuel vehicles (AFVs) are vehicles designed to operate on alternative fuels such as compressed and liquefied natural gas,...

352

The burnup dependence of light water reactor spent fuel oxidation  

SciTech Connect

Over the temperature range of interest for dry storage or for placement of spent fuel in a permanent repository under the conditions now being considered, UO{sub 2} is thermodynamically unstable with respect to oxidation to higher oxides. The multiple valence states of uranium allow for the accommodation of interstitial oxygen atoms in the fuel matrix. A variety of stoichiometric and nonstoichiometric phases is therefore possible as the fuel oxidizers from UO{sub 2} to higher oxides. The oxidation of UO{sub 2} has been studied extensively for over 40 years. It has been shown that spent fuel and unirradiated UO{sub 2} oxidize via different mechanisms and at different rates. The oxidation of LWR spent fuel from UO{sub 2} to UO{sub 2.4} was studied previously and is reasonably well understood. The study presented here was initiated to determine the mechanism and rate of oxidation from UO{sub 2.4} to higher oxides. During the early stages of this work, a large variability in the oxidation behavior of samples oxidized under nearly identical conditions was found. Based on previous work on the effect of dopants on UO{sub 2} oxidation and this initial variability, it was hypothesized that the substitution of fission product and actinide impurities for uranium atoms in the spent fuel matrix was the cause of the variable oxidation behavior. Since the impurity concentration is roughly proportional to the burnup of a specimen, the oxidation behavior of spent fuel was expected to be a function of both temperature and burnup. This report (1) summarizes the previous oxidation work for both unirradiated UO{sub 2} and spent fuel (Section 2.2) and presents the theoretical basis for the burnup (i.e., impurity concentration) dependence of the rate of oxidation (Sections 2.3, 2.4, and 2.5), (2) describes the experimental approach (Section 3) and results (Section 4) for the current oxidation tests on spent fuel, and (3) establishes a simple model to determine the activation energies associated with spent fuel oxidation (Section 5).

Hanson, B.D.

1998-07-01T23:59:59.000Z

353

On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks  

SciTech Connect

This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

Samuel Bays; Ayodeji Alajo

2010-05-01T23:59:59.000Z

354

Development of the JAERI (Japan Atomic Energy Research Institute) fuel cleanup system for tests at the Tritium Systems Test Assembly  

Science Conference Proceedings (OSTI)

Tritium Process Laboratory (TPL) at the Japan Atomic Energy Research Institute (JAERI) has developed the Fuel Cleanup System (FCU) which accepts simulated fusion reactor exhaust and produces pure hydrogen isotopes and tritium-free waste. The major components are: a palladium diffuser, a catalytic reactor, cold traps, a ceramic electrolysis cell, and zirconium-cobalt beds. In 1988, an integrated loop of the FCU process was installed in the TPL and a number of hot'' runs were performed to study the system characteristics and improve system performance. Under the US-Japan collaboration program, the JAERI Fuel Cleanup System'' (JFCU) was designed and fabricated by JAERI/TPL for testing at the Tritium Systems Test Assembly (TSTA) in Los Alamos National Laboratory as a major subsystem of the simulated fusion fuel cycle. The JFCU was installed in the TSTA in early 1990.

Konishi, S.; Inoue, M.; Hayashi, T.; Okuno, K.; Naruse, Y. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)); Barnes, J.W.; Anderson, J.L. (Los Alamos National Lab., NM (USA))

1990-01-01T23:59:59.000Z

355

Fast Reactor Spent Fuel Processing: Experience and Criticality Safety  

SciTech Connect

This paper discusses operational and criticality safety experience associated with the Idaho National Laboratory Fuel Conditioning Facility which uses a pyrometallurgical process to treat spent fast reactor metallic fuel. The process is conducted in an inert atmosphere hot cell. The process starts with chopping metallic fuel elements into a basket. The basket is lowered into molten salt (LiCl-KCl) along with a steel mandrel. Active metal fission products, transuranic metals and sodium metal in the spent fuel undergo chemical oxidation and form chlorides. Voltage is applied between the basket, which serves as an anode, and the mandrel, which serves as a cathode, causing metallic uranium in the spent fuel to undergo electro-chemical oxidation thereby forming uranium chloride. Simultaneously at the cathode, uranium chloride undergoes electro-chemical reduction and deposits uranium metal onto the mandrel. The uranium metal and accompanying entrained salt are placed in a distillation furnace where the uranium melts forming an ingot and the entrained salt boils and subsequently condenses in a separate crucible. The uranium ingots are placed in long term storage. During the ten year operating history, over one hundred criticality safety evaluations were prepared. All criticality safety related limits and controls for the entire process are contained in a single document which required over thirty revisions to accommodate the process changes. Operational implementation of the limits and controls includes use of a near real-time computerized tracking system. The tracking system uses an Oracle database coupled with numerous software applications. The computerized tracking system includes direct fuel handler interaction with every movement of material. Improvements to this system during the ten year history include introduction of web based operator interaction, tracking of moderator materials and the development of a plethora database queries to assist in day to day operations as well as obtaining historical information. Over 12,000 driver fuel elements have been processed resulting in the production of 2500 kg of 20% enriched uranium. Also, over one thousand blanket fuel elements have been processed resulting in the production of 2400 kg of depleted uranium. These operations required over 35,000 fissile material transfers between zones and over 6000 transfers between containers. Throughout all of these movements, no mass limit violations occurred. Numerous lessons were learned over the ten year operating history. From a criticality safety perspective, the most important lesson learned was the involvement of a criticality safety practitioner in daily operations. A criticality safety engineer was assigned directly to facility operations, and was responsible for implementation of limits and controls including upkeep of the associated computerized tracking files. The criticality safety engineer was also responsible for conducting fuel handler training activities including serving on fuel handler qualification oral boards, and continually assessing operations from a criticality control perspective. The criticality safety engineer also attended bimonthly project planning meetings to identify upcoming process changes that would require criticality safety evaluation. Finally, the excellent criticality safety record was due in no small part to the continual support, involvement, trust, and confidence of project and operations mana

Chad Pope

2007-05-01T23:59:59.000Z

356

Foreign Research Reactor Spent Nuclear Fuel Acceptance Program  

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

Global Threat Reduction Initiative: Global Threat Reduction Initiative: U.S. Nuclear Remove Program Foreign Research Reactor Spent Nuclear Fuel (FRR SNF) Acceptance 2007 DOE TEC Meeting Chuck Messick DOE/NNSA/SRS 2 Contents * Program Objective and Policy * Program implementation status * Shipment Information * Operational Logistics * Lessons Learned * Conclusion 3 U.S. Nuclear Remove Program Objective * To play a key role in the Global Threat Reduction Remove Program supporting permanent threat reduction by accepting program eligible material. * Works in conjunction with the Global Threat Reduction Convert Program to accept program eligible material as an incentive to core conversion providing a disposition path for HEU and LEU during the life of the Acceptance Program. 4 Reasons for the Policy

357

FUEL CYCLE COSTS IN A GRAPHITE MODERATED SLIGHTLY ENRICHED FUSED SALT REACTOR  

SciTech Connect

A fuel cycle economic study has been made for a 315Mwe graphite- moderated slightly enriched fused-salt reactor. Fuel cycle costs of less than 1.5 mills may be possible for such reactors operating on a ten-year cycle even when the fuel is discarded at the end of the cycle. Recovery of the uranium and plutonium at the end of the cycle reduces the fuel cycle costs to approximates 1 mill/kwh. Changes in the waste storage cost, reprocessing cost or salt inventory have a relatively minor effect on fuel cycle costs. (auth)

Guthrie, C.E.

1959-01-01T23:59:59.000Z

358

Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel  

SciTech Connect

One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

1994-03-25T23:59:59.000Z

359

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

360

The First Decommissioning of a Fusion Reactor Fueled by Deuterium-Tritium  

SciTech Connect

The Tokamak Fusion Test Reactor (TFTR) at the Plasma Physics Laboratory of Princeton University (PPPL) was the first fusion reactor fueled by a mixture of deuterium and tritium (D-T) to be decommissioned in the world. The decommissioning was performed over a period of three years and was completed safely, on schedule, and under budget. Provided is an overview of the project and detail of various factors which led to the success of the project. Discussion will cover management of the project, engineering planning before the project started and during the field work as it was being performed, training of workers in the field, the novel adaptation of tools from other industry, and the development of an innovative process for the use of diamond wire to segment the activated/contaminated vacuum vessel. The success of the TFTR decommissioning provides a viable model for the decommissioning of D-T burning fusion devices in the future.

Charles A. Gentile; Erik Perry; Keith Rule; Michael Williams; Robert Parsells; Michael Viola; James Chrzanowski

2003-10-28T23:59:59.000Z

Note: This page contains sample records for the topic "testing reactor fuels" 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

Pressurized solid oxide fuel cell testing  

DOE Green Energy (OSTI)

The Pressurized SOFC Test Program is an integral part of the Cooperative Agreement between Westinghouse and DOE and was put into place to evaluate the effects of pressurization on SOFC performance. The goals of the SOFC pressurized test program are to obtain cell voltage versus current (VI) performance data as a function of pressure; to evaluate the effects of operating parameters such as temperature, air stoichiometry, and fuel utilization on cell performance, and to demonstrate long term stability of the SOFC materials at elevated pressures.

Ray, E.R.; Basel, R.A.; Pierre, J.F.

1995-12-31T23:59:59.000Z

362

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

experience in the nuclear fuels field. I am also extremelyreactor core components, nuclear fuel-element design hasreactors, commercial nuclear fuel still consists of uranium

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

363

Validation and application of a physics database for fast reactor fuel cycle analysis  

SciTech Connect

An effort has been made to automate the execution of fast reactor fuel cycle analysis, using EBR-II as a demonstration vehicle, and to validate the analysis results for application to the IFR closed fuel cycle demonstration at EBR-II and its fuel cycle facility. This effort has included: (1) the application of the standard ANL depletion codes to perform core-follow analyses for an extensive series of EBR-II runs, (2) incorporation of the EBR-II data into a physics database, (3) development and verification of software to update, maintain and verify the database files, (4) development and validation of fuel cycle models and methodology, (5) development and verification of software which utilizes this physics database to automate the application of the ANL depletion codes, methods and models to perform the core-follow analysis, and (6) validation studies of the ANL depletion codes and of their application in support of anticipated near-term operations in EBR-II and the Fuel Cycle Facility. Results of the validation tests indicate the physics database and associated analysis codes and procedures are adequate to predict required quantities in support of early phases of FCF operations.

McKnight, R.D.; Stillman, J.A.; Toppel, B.J.; Khalil, H.S.

1994-03-01T23:59:59.000Z

364

Legacy Vehicle Fuel System Testing with Intermediate Ethanol Blends  

DOE Green Energy (OSTI)

The effects of E10 and E17 on legacy fuel system components from three common mid-1990s vintage vehicle models (Ford, GM, and Toyota) were studied. The fuel systems comprised a fuel sending unit with pump, a fuel rail and integrated pressure regulator, and the fuel injectors. The fuel system components were characterized and then installed and tested in sample aging test rigs to simulate the exposure and operation of the fuel system components in an operating vehicle. The fuel injectors were cycled with varying pulse widths during pump operation. Operational performance, such as fuel flow and pressure, was monitored during the aging tests. Both of the Toyota fuel pumps demonstrated some degradation in performance during testing. Six injectors were tested in each aging rig. The Ford and GM injectors showed little change over the aging tests. Overall, based on the results of both the fuel pump testing and the fuel injector testing, no major failures were observed that could be attributed to E17 exposure. The unknown fuel component histories add a large uncertainty to the aging tests. Acquiring fuel system components from operational legacy vehicles would reduce the uncertainty.

Davis, G. W.; Hoff, C. J.; Borton, Z.; Ratcliff, M. A.

2012-03-01T23:59:59.000Z

365

Mechanical Characteristics of Fresh and Irradiated U-10Mo Fuel  

Science Conference Proceedings (OSTI)

Abstract Scope, The Reduced Enrichment for Research and Test Reactors ... test reactors to use low-enriched uranium instead of high-enriched uranium fuels ...

366

Storage of spent fuel from the nation`s nuclear reactors: Status, technology, and policy options  

SciTech Connect

Since the beginning of the commercial nuclear electric power industry, it has been recognized that spent nuclear reactor fuel must be able to be readily removed from the reactor vessel in the plant and safely stored on-site. The need for adjacent ready storage is first for safety. In the event of an emergency, or necessary maintenance that requires the removal of irradiated fuel from the reactor vessel, cooled reserve storage capacity for the full amount of fuel from the reactor core must be available. Also, the uranium fuel in the reactor eventually reaches the point where its heat generation is below the planned efficiency for steam production which drives the turbines and generators. It then must be replaced by fresh uranium fuel, with the ``spent fuel`` elements being removed to a safe and convenient storage location near the reactor vessel. The federal nuclear waste repository program, even without delays in the current schedule of disposal becoming available in 2003, will result in a large percentage of the 111 existing operable commercial reactors requiring expansion of their spent fuel storage capacity. How that need can and will be met raises issues of both technology and policy that will be reviewed in this report.

1989-10-01T23:59:59.000Z

367

Fuel Reliability Program: Post-Irradiation Examination and Performance Assessment of ATRIUM-10 BWR Fuel from Browns Ferry-3 Reactor  

Science Conference Proceedings (OSTI)

ATRIUM-10 design (10x10 lattice) fuel was irradiated for one 24-month period during Cycle 12 to 25 MWd/kgU rod-average exposure at Tennessee Valley Authority's Browns Ferry Unit 3 reactor. The project goal was to characterize the behavior of modern boiling water reactor (BWR) fuel at low exposures to assess early-life performance in a well-documented reactor environment. This report includes results from hot cell post-irradiation examinations. In a future Electric Power Research Institute (EPRI) report, ...

2011-06-09T23:59:59.000Z

368

Fuel Cell Development and Test Laboratory (Fact Sheet)  

DOE Green Energy (OSTI)

This fact sheet describes the purpose, lab specifications, applications scenarios, and information on how to partner with NREL's Fuel Cell Development and Test Laboratory at the Energy Systems Integration Facility. NREL's state-of-the-art Fuel Cell Development and Test Laboratory in the Energy Systems Integration Facility (ESIF) supports NREL's fuel cell research and development projects through in-situ fuel cell testing. Current projects include various catalyst development projects, a system contaminant project, and the manufacturing project. Testing capabilities include but are not limited to single cell fuel cells and fuel cell stacks.

Not Available

2011-10-01T23:59:59.000Z

369

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

370

Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor  

Science Conference Proceedings (OSTI)

The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge burnup level, while retaining its inherent safety characteristics. Using generic pebble bed reactor cores, this task will perform physics calculations to evaluate the capabilities of the pebble bed reactor to perform utilization and destruction of LWR used-fuel transuranics. The task will use established benchmarked models, and will introduce modeling advancements appropriate to the nature of the fuel considered (high TRU content and high burn-up).

B. Boer; A. M. Ougouag

2010-09-01T23:59:59.000Z

371

Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors  

E-Print Network (OSTI)

of plutonium attainable with MOX fuel [24, 23]. In theof recycles feasible with MOX fuel is limited because the

Terrani, Kurt Amir

2010-01-01T23:59:59.000Z

372

Metal-fueled HWR (heavy water reactors) severe accident issues: Differences and similarities to commercial LWRs (light water reactors)  

DOE Green Energy (OSTI)

Differences and similarities in severe accident progression and phenomena between commercial Light Water Reactors (LWR) and metal-fueled isotopic production Heavy Water Reactors (HWR) are described. It is very important to distinguish between accident progression in the two systems because each reactor type behaves in a unique manner to a fuel melting accident. Some of the lessons learned as a result of the extensive commercial severe accident research are not applicable to metal-fueled heavy water reactors. A direct application of severe accident phenomena developed from oxide-fueled LWRs to metal-fueled HWRs may lead to large errors or substantial uncertainties. In general, the application of severe accident LWR concepts to HWRs should be done with the intent to define the relevant issues, define differences, and determine areas of overlap. This paper describes the relevant differences between LWR and metal-fueled HWR severe accident phenomena. Also included in the paper is a description of the phenomena that govern the source term in HWRs, the areas where research is needed to resolve major uncertainties, and areas in which LWR technology can be directly applied with few modifications.

Ellison, P.G.; Hyder, M.L.; Monson, P.R. (Westinghouse Savannah River Co., Aiken, SC (USA)); Coryell, E.W. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1990-01-01T23:59:59.000Z

373

Nuclear mass inventory, photon dose rate and thermal decay heat of spent research reactor fuel assemblies  

SciTech Connect

This document has been prepared to assist research reactor operators possessing spent fuel containing enriched uranium of United States origin to prepare part of the documentation necessary to ship this fuel to the United States. Data are included on the nuclear mass inventory, photon dose rate, and thermal decay heat of spent research reactor fuel assemblies. Isotopic masses of U, Np, Pu and Am that are present in spent research reactor fuel are estimated for MTR, TRIGA and DIDO-type fuel assembly types. The isotopic masses of each fuel assembly type are given as functions of U-235 burnup in the spent fuel, and of initial U-235 enrichment and U-235 mass in the fuel assembly. Photon dose rates of spent MTR, TRIGA and DIDO-type fuel assemblies are estimated for fuel assemblies with up to 80% U-235 burnup and specific power densities between 0.089 and 2.857 MW/kg[sup 235]U, and for fission product decay times of up to 20 years. Thermal decay heat loads are estimated for spent fuel based upon the fuel assembly irradiation history (average assembly power vs. elapsed time) and the spent fuel cooling time.

Pond, R.B.; Matos, J.E.

1996-12-31T23:59:59.000Z

374

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

375

Core Designs and Economic Analyses of Homogeneous Thoria-Urania Fuel in Light Water Reactors  

SciTech Connect

The objective is to develop equilibrium fuel cycle designs for a typical pressurized water reactor (PWR) loaded with homogeneously mixed uranium-thorium dioxide (ThO{sub 2}-UO{sub 2}) fuel and compare those designs with more conventional UO{sub 2} designs.The fuel cycle analyses indicate that ThO{sub 2}-UO{sub 2} fuel cycles are technically feasible in modern PWRs. Both power peaking and soluble boron concentrations tend to be lower than in conventional UO{sub 2} fuel cycles, and the burnable poison requirements are less.However, the additional costs associated with the use of homogeneous ThO{sub 2}-UO{sub 2} fuel in a PWR are significant, and extrapolation of the results gives no indication that further increases in burnup will make thoria-urania fuel economically competitive with the current UO{sub 2} fuel used in light water reactors.

Saglam, Mehmet; Sapyta, Joe J.; Spetz, Stewart W.; Hassler, Lawrence A. [Framatome ANP, Inc. (France)

2004-07-15T23:59:59.000Z

376

Used Fuel Disposition Campaign Phase I Ring Compression Testing...  

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

Phase I Ring Compression Testing of High Burnup Cladding Used Fuel Disposition Campaign Phase I Ring Compression Testing of High Burnup Cladding The purpose of ring compression...

377

Status of the Integral Fast Reactor fuel cycle demonstration and waste management practices  

SciTech Connect

Over the past few years, Argonne National Laboratory has been preparing for the demonstration of the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety and operations, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle, which will be demonstrated at Argonne-West in Idaho, employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The required facility modifications and process equipment for the demonstration are nearing completion. Their status and the results from initial fuel fabrication work, including the waste management aspects, are presented. Additionally, estimated compositions of the various process waste streams have been made, and characterization and treatment methods are being developed. The status of advanced waste processing equipment being designed and fabricated is described.

Benedict, R.W.; Goff, K.M.; McFarlane, H.F.

1994-07-01T23:59:59.000Z

378

Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin  

SciTech Connect

This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy`s Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006.

Matos, J.E.

1994-12-01T23:59:59.000Z

379

The evaluation of the use of metal alloy fuels in pressurized water reactors. Final report  

Science Conference Proceedings (OSTI)

The use of metal alloy fuels in a PWR was investigated. It was found that it would be feasible and competitive to design PWRs with metal alloy fuels but that there seemed to be no significant benefits. The new technology would carry with it added economic uncertainty and since no large benefits were found it was determined that metal alloy fuels are not recommended. Initially, a benefit was found for metal alloy fuels but when the oxide core was equally optimized the benefit faded. On review of the optimization of the current generation of ``advanced reactors,`` it became clear that reactor design optimization has been under emphasized. Current ``advanced reactors`` are severely constrained. The AP-600 required the use of a fuel design from the 1970`s. In order to find the best metal alloy fuel design, core optimization became a central effort. This work is ongoing.

Lancaster, D.

1992-10-26T23:59:59.000Z

380

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

Note: This page contains sample records for the topic "testing reactor fuels" 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

The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels  

DOE Green Energy (OSTI)

Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases and thus reduce three temperature losses in the system associated with (1) heat transfer from the fuel to the reactor coolant, (2) temperature rise across the reactor core, and (3) heat transfer across the heat exchangers between the reactor and H2 production plant. Lowering the peak reactor temperatures and thus reducing the high-temperature materials requirements may make the AHTR the enabling technology for low-cost nuclear hydrogen production.

Forsberg, C.W.; Peterson, P.F.; Ott, L.

2004-10-06T23:59:59.000Z

382

Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report  

SciTech Connect

Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºC to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the beginning of the materials processing setup. Also included within this section is a thesis proposal by Jeff Hausaman. Appendix C contains the public papers and presentations introduced at the 2010 American Nuclear Society Winter Meeting. Appendix A—MSNE theses of David Garnetti and Grant Helmreich and proposal by Jeff Hausaman A.1 December 2009 Thesis by David Garnetti entitled “Uranium Powder Production Via Hydride Formation and Alpha Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.2 September 2009 Presentation by David Garnetti (same title as document in Appendix B.1) A.3 December 2010 Thesis by Grant Helmreich entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys for Advanced Nuclear Fuel Applications” A.4 October 2010 Presentation by Grant Helmreich (same title as document in Appendix B.3) A.5 Thesis Proposal by Jeffrey Hausaman entitled “Hot Extrusion of Alpha Phase Uranium-Zirconium Alloys for TRU Burning Fast Reactors” Appendix B—External presentations introduced at the 2010 ANS Winter Meeting B.1 J.S. Hausaman, D.J. Garnetti, and S.M. McDeavitt, “Powder Metallurgy of Alpha Phase Uranium Alloys for TRU Burning Fast Reactors,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.2 PowerPoint Presentation Slides from C.1 B.3 G.W. Helmreich, W.J. Sames, D.J. Garnetti, and S.M. McDeavitt, “Uranium Powder Production Using a Hydride-Dehydride Process,” Proceedings of 2010 ANS Winter Meeting, Las Vegas, Nevada, USA, November 7-10, 2010 B.4. PowerPoint Presentation Slides from C.3 B.5 Poster Presentation from C.3 Appendix C—Fuel cycle research and development undergraduate materials and poster presentation C.1 Poster entitled “Characterization of Alpha-Phase Sintering of Uranium and Uranium-Zirconium Alloys” presented at the Fuel Cycle Technologies Program Annual Meeting C.2 April 2011 Honors Undergraduate Thesis by William Sames, Research Fellow, entitled “Uranium Metal Powder Production, Particle Dis

Sean M. McDeavitt

2011-04-29T23:59:59.000Z

383

Fuel Grading Study on a Low-Enriched Uranium Fuel Design for the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2009-11-01T23:59:59.000Z

384

Assemblies with both target and fuel pins in an isotope-production reactor  

DOE Patents (OSTI)

A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

Cawley, W.E.; Omberg, R.P.

1982-08-19T23:59:59.000Z

385

Development of a High Temperature Gas-Cooled Reactor TRISO-coated particle fuel chemistry model  

E-Print Network (OSTI)

The first portion of this work is a comprehensive analysis of the chemical environment in a High Temperature Gas-Cooled Reactor TRISO fuel particle. Fission product inventory versus burnup is calculated. Based on those ...

Diecker, Jane T

2005-01-01T23:59:59.000Z

386

General analysis of breed-and-burn reactors and limited-separations fuel cycles  

E-Print Network (OSTI)

A new theoretical framework is introduced, the "neutron excess" concept, which is useful for analyzing breed-and-burn (B&B) reactors and their fuel cycles. Based on this concept, a set of methods has been developed which ...

Petroski, Robert C

2011-01-01T23:59:59.000Z

387

Implementation of vented fuel assemblies in the supercritical CO?-cooled fast reactor  

E-Print Network (OSTI)

Analysis has been undertaken to investigate the utilization of fuel assembly venting in the reference design of the gas-cooled fast reactor under study as part of the larger research effort at MIT under Gen-IV NERI Project ...

McKee, Stephanie A

2008-01-01T23:59:59.000Z

388

Design strategies for optimizing high burnup fuel in pressurized water reactors  

E-Print Network (OSTI)

This work is focused on the strategy for utilizing high-burnup fuel in pressurized water reactors (PWR) with special emphasis on the full array of neutronic considerations. The historical increase in batch-averaged discharge ...

Xu, Zhiwen, 1975-

2003-01-01T23:59:59.000Z

389

Forced-convection boiling tests performed in parallel simulated LMR fuel assemblies  

SciTech Connect

Forced-convection tests have been carried out using parallel simulated Liquid Metal Reactor fuel assemblies in an engineering-scale sodium loop, the Thermal-Hydraulic Out-of-Reactor Safety facility. The tests, performed under single- and two-phase conditions, have shown that for low forced-convection flow there is significant flow augmentation by thermal convection, an important phenomenon under degraded shutdown heat removal conditions in an LMR. The power and flows required for boiling and dryout to occur are much higher than decay heat levels. The experimental evidence supports analytical results that heat removal from an LMR is possible with a degraded shutdown heat removal system.

Rose, S.D.; Carbajo, J.J.; Levin, A.E.; Lloyd, D.B.; Montgomery, B.H.; Wantland, J.L.

1985-04-21T23:59:59.000Z

390

Alternative Fuels Data Center: Biodiesel Quality Testing Procedures  

Alternative Fuels and Advanced Vehicles Data Center (EERE)

Biodiesel Quality Biodiesel Quality Testing Procedures to someone by E-mail Share Alternative Fuels Data Center: Biodiesel Quality Testing Procedures on Facebook Tweet about Alternative Fuels Data Center: Biodiesel Quality Testing Procedures on Twitter Bookmark Alternative Fuels Data Center: Biodiesel Quality Testing Procedures on Google Bookmark Alternative Fuels Data Center: Biodiesel Quality Testing Procedures on Delicious Rank Alternative Fuels Data Center: Biodiesel Quality Testing Procedures on Digg Find More places to share Alternative Fuels Data Center: Biodiesel Quality Testing Procedures on AddThis.com... More in this section... Federal State Advanced Search All Laws & Incentives Sorted by Type Biodiesel Quality Testing Procedures Each biodiesel or other renewable diesel producer, distributor, or importer

391

Introduction to Nuclear Reactors, Fuels, and Materials: Heather ...  

Science Conference Proceedings (OSTI)

Feb 27, 2012 ... What goes on in a nuclear power plant. • Challenges in nuclear fuels and materials. Key lessons: • Fuels and materials change during ...

392

Novel Reactor Design for Solid Fuel Chemical Looping Combustion  

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

for Solid Fuel Chemical Looping Combustion Opportunity Research is active on the patent pending technology, titled "Apparatus and Method for Solid Fuel Chemical Looping...

393

Conversion and standardization of university reactor fuels using low-enrichment uranium - options and costs  

SciTech Connect

The highly-enriched uranium (HEU) fuel used in twenty United States university reactors can be viewed as contributing to the risk of theft or diversion of weapons-useable material. The US Nuclear Regulatory Commission has issued a policy statement expressing its concern and has published a proposed rule on limiting the use of HEU in NRC-licensed non-power reactors. The fuel options, functional impacts, licensing, and scheduling of conversion and standardization of these reactor fuels to use of low-enrichment uranium (LEU) have been assessed. The university reactors span a wide range in form and function, from medium-power intense neutron sources where HEU fuel may be required, to low-power training and research facilities where HEU fuel is unnecessary. Conversion provides an opportunity to standardize university reactor fuels and improve reactor utilization in some cases. The entire program is estimated to cost about $10 million and to last about five years. Planning for conversion and standardization is facilitated by the US Department of Energy. 20 refs., 1 tab.

Harris, D.R.; Matos, J.E.; Young, H.H.

1985-01-01T23:59:59.000Z

394

Spent Fuel Drying System Test Results (Dry-Run in Preparation for Run 8)  

Science Conference Proceedings (OSTI)

The water-filled K-Basins in the Hanford 100 Area have been used to store N-Reactor spent nuclear fuel (SNF) since the 1970s. Because some leaks in the basin have been detected and some of the fuel is breached due to handling damage and corrosion, efforts are underway to remove the fuel elements from wet storage. An Integrated Process Strategy (IPS) has been developed to package, dry, transport, and store these metallic uranium fuel elements in an interim storage facility on the Hanford Site (WHC 1995). Information required to support the development of the drying processes, and the required safety analyses, is being obtained from characterization tests conducted on fuel elements removed from the K-Basins. A series of whole element drying tests (reported in separate documents, see Section 7.0) have been conducted by Pacific Northwest National Laboratory (PNNL)(a)on several intact and damaged fuel elements recovered from both the K-East and K-West Basins. This report documents the results of a test ''dry-run'' conducted prior to the eighth and last of those tests, which was conducted on an N-Reactor outer fuel element removed from K-West canister6513U. The system used for the dry-run test was the Whole Element Furnace Testing System, described in Section 2.0, located in the Postirradiation Testing Laboratory (PTL, 327 Building). The test conditions and methodologies are given in Section 3.0. The experimental results are provided in Section 4.0 and discussed Section 5.0.

BM Oliver; GS Klinger; J Abrefah; SC Marschman; PJ MacFarlan; GA Ritter

1999-08-11T23:59:59.000Z

395

Optimum Discharge Burnup for Nuclear Fuel: A Comprehensive Study of Duke Power's Reactors  

Science Conference Proceedings (OSTI)

Economic analysis of two pressurized water reactors (PWRs) shows that increasing the discharge burnup of light water reactor (LWR) fuel above current values can result in significant cost benefits. Optimum discharge burnup levels, however, may not be achievable without exceeding the current limit on enrichment.

1999-06-01T23:59:59.000Z

396

The behaviour of transuranic mixed oxide fuel in a Candu-900 reactor  

SciTech Connect

The production of transuranic actinide fuels for use in current thermal reactors provides a useful intermediary step in closing the nuclear fuel cycle. Extraction of actinides reduces the longevity, radiation and heat loads of spent material. The burning of transuranic fuels in current reactors for a limited amount of cycles reduces the infrastructure demand for fast reactors and provides an effective synergy that can result in a reduction of as much as 95% of spent fuel waste while reducing the fast reactor infrastructure needed by a factor of almost 13.5 [1]. This paper examines the features of actinide mixed oxide fuel, TRUMOX, in a CANDU{sup R}* nuclear reactor. The actinide concentrations used were based on extraction from 30 year cooled spent fuel and mixed with natural uranium in 3.1 wt% actinide MOX fuel. Full lattice cell modeling was performed using the WIMS-AECL code, super-cell calculations were analyzed in DRAGON and full core analysis was executed in the RFSP 2-group diffusion code. A time-average full core model was produced and analyzed for reactor coefficients, reactivity device worth and online fuelling impacts. The standard CANDU operational limits were maintained throughout operations. The TRUMOX fuel design achieved a burnup of 27.36 MWd/kg HE. A full TRUMOX fuelled CANDU was shown to operate within acceptable limits and provided a viable intermediary step for burning actinides. The recycling, reprocessing and reuse of spent fuels produces a much more sustainable and efficient nuclear fuel cycle. (authors)

Morreale, A. C.; Ball, M. R.; Novog, D. R.; Luxat, J. C. [Dept. of Engineering Physics, McMaster Univ., 1280 Main St. W, Hamilton, ON (Canada)

2012-07-01T23:59:59.000Z

397