WorldWideScience

Sample records for rertr fuel developmemt

  1. Solvent extraction studies of RERTR silicide fuels

    International Nuclear Information System (INIS)

    Gouge, Anthony P.

    1983-01-01

    Uranium silicide fuels, which are candidate RERTR fuel compositions, may require special considerations in solvent extraction reprocessing. Since Savannah River Plant may be reprocessing RERTR fuels as early as 1985, studies have been conducted at Savannah River Laboratory to demonstrate the solvent extraction behavior of this fuel. Results of solvent extraction studies with both unirradiated and irradiated fuel are presented along with the preliminary RERTR solvent extraction reprocessing flow sheet for Savannah River Plant. (author)

  2. Metallographic analysis of irradiated RERTR-3 fuel test specimens

    International Nuclear Information System (INIS)

    Meyer, M. K.; Hofman, G. L.; Strain, R. V.; Clark, C. R.; Stuart, J. R.

    2000-01-01

    The RERTR-3 irradiation test was designed to investigate the irradiation behavior of aluminum matrix U-MO alloy dispersion fuels under high-temperature, high-fission-rate conditions. Initial postirradiation examination of RERTR-3 fuel specimens has concentrated on binary U-MO atomized fuels. The rate of matrix aluminum depletion was found to be higher than predictions based on low temperature irradiation data. Wavelength Dispersive X-ray Spectroscopy (WDS) indicates that aluminum is present in the interior of the fuel particles. WDS data is supported by a mass and volume balance calculation performed on the basis of image analysis results. The depletion of matrix aluminum seems to have no detrimental effects on fuel performance under the conditions tested to date

  3. Irradiation behavior of uranium-molybdenum dispersion fuel: Fuel performance data from RERTR-1 and RERTR-2

    International Nuclear Information System (INIS)

    Meyer, M.K.; Clark, C.R.; Hayes, S.L.; Strain, R.V.; Hofman, G.L.; Snelgrove, J.L.; Park, J.M.; Kim, K.H.

    1999-01-01

    This paper presents quantitative data on the irradiation behavior of uranium-molybdenum fuels from the low temperature RERTR-1 and -2 experiments. Fuel swelling measurements of U-Mo fuels at ∼40% and ∼70% burnup are presented. The rate of fuel-matrix interaction layer growth is estimated. Microstructures of fuel in the pre- and postirradiation condition were compared. Based on these data, a qualitative picture of the evolution of the U-Mo fuel microstructure during irradiation has been developed. Estimates of uranium-molybdenum fuel swelling and fuel-matrix interaction under high-power research reactor operating conditions are presented. (author)

  4. Modeling RERTR experimental fuel plates using the PLATE code

    International Nuclear Information System (INIS)

    Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Snelgrove, J.L.; Brazener, R.A.

    2003-01-01

    Modeling results using the PLATE dispersion fuel performance code are presented for the U-Mo/Al experimental fuel plates from the RERTR-1, -2, -3 and -5 irradiation tests. Agreement of the calculations with experimental data obtained in post-irradiation examinations of these fuels, where available, is shown to be good. Use of the code to perform a series of parametric evaluations highlights the sensitivity of U-Mo dispersion fuel performance to fabrication variables, especially fuel particle shape and size distributions. (author)

  5. RERTR program progress in qualifying reduced-enrichment fuels

    International Nuclear Information System (INIS)

    Snelgrove, James L.

    1983-01-01

    In order to provide the technical means for reducing the enrichment of uranium used to fuel research and test reactors, the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program has been engaged in the development and testing of higher-uranium-density fuels than had been used previously. This fuel development effort included work to increase the density of fuels which were being used at the time the Program began and work on a fuel with the potential for much higher density. The ultimate goal of the fuel development and testing phase of the Program is to 'qualify' the fuel for use. A fuel is considered qualified when a sufficient data base for the fuel exists that it can be approved by regulating bodies for use in reactors. To convert a core to the use of reduced-enrichment fuel it is necessary to show that the core will behave properly during normal and off-normal operating conditions and to show that the fuel will behave properly to a reasonable margin beyond the conditions expected during normal operation. It is this latter area that this paper will address. The main characteristics to be considered in evaluating the performance of a fuel are its swelling, its blister-threshold temperature, and its metallurgical appearance. Data for the qualification of the reduced-enrichment fuels being developed by the RERTR Program are obtained from examination of miniature fuel plates (miniplates) which successfully pass the irradiation screening tests and from examinations of full-sized fuel elements. This paper will summarize the miniplate data reported in other papers presented during this meeting and will give the status of full-sized element irradiations. Finally, the current status of qualification of the various fuel types will be discussed and some projections of the future will be given

  6. Stability Study of the RERTR Fuel Microstructure

    Energy Technology Data Exchange (ETDEWEB)

    Jian Gan; Dennis Keiser; Brandon Miller; Daniel Wachs

    2014-04-01

    The irradiation stability of the interaction phases at the interface of fuel and Al alloy matrix as well as the stability of the fission gas bubble superlattice is believed to be very important to the U-Mo fuel performance. In this paper the recent result from TEM characterization of Kr ion irradiated U-10Mo-5Zr alloy will be discussed. The focus will be on the phase stability of Mo2-Zr, a dominated second phase developed at the interface of U-10Mo and the Zr barrier in a monolithic fuel plate from fuel fabrication. The Kr ion irradiations were conducted at a temperature of 200 degrees C to an ion fluence of 2.0E+16 ions/cm2. To investigate the thermal stability of the fission gas bubble superlattice, a key microstructural feature in both irradiated dispersion U-7Mo fuel and monolithic U-10Mo fuel, a FIB-TEM sample of the irradiated U-10Mo fuel (3.53E+21 fission/cm3) was used for a TEM in-situ heating experiment. The preliminary result showed extraordinary thermal stability of the fission gas bubble superlattice. The implication of the TEM observation from these two experiments on the fuel microstructural evolution under irradiation will be discussed.

  7. Irradiation testing of miniature fuel plates for the RERTR program

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R L; Martin, M M [Oak Ridge National Laboratory, Oak Ridge, TN 37830 (United States)

    1983-08-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. The objective of these tests is to screen various candidate fuel materials as to their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% {sup 235}U in place of highly enriched fuel for these reactors would reduce the potential for {sup 235}U diversion. Fuel materials currently being evaluated in this first phase of these screening tests include aluminum-base dispersion-type fuel plates with fuel cores of 1) high uranium content U{sup 3}){sup 8}-Al being developed by ORNL, 2) high uranium content UAI{sub x}-Al being developed by EG and G Idaho, Inc., and 3) very high uranium content U{sub 3}Si-Al- being developed by ANL. The miniplates are 115-mm long by 50-mm wide with overall plate thicknesses of 1.27 or 1.52 mm. The fuel core dimensions vary according to overall plate thicknesses with a minimal clad thickness requirement of 0.20 mm. Sixty such miniplates (thirty of each thickness) can be irradiated in one test facility. The irradiation test facility, designated as HFED-1 is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The peak neutron flux measured for this experiment is 1.96 x 10{sup 18} neutrons m{sub -2} s{sub -1}. The various types of miniplates will achieve burnups of up to approximately 2.2x10{sup 27} fissions/m{sup 3} of fuel, which will require approximately eight full power months of irradiation. During reactor shutdown periods, the experiment is removed from the reactor, moved to a special poolside station, disassembled, and inspected

  8. RERTR program activities related to the development and application of new LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-01-01

    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 3 Si 2 -Al and U 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 3 each year, from the current 1.7 g U/cm 3 to the 7.0 g U/cm 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

  9. Characterization of an irradiated RERTR-7 fuel plate using transmission electron microscopy

    International Nuclear Information System (INIS)

    Gan, J.; Keiser, D.D. Jr.; Miller, B.D.; Robinson, A.B.; Medvedev, P.

    2010-01-01

    Transmission electron microscopy (TEM) has been used to characterize an irradiated fuel plate with Al-2Si matrix from the Reduced Enrichment Research and Test Reactor RERTR-7 experiment that was irradiated under moderate reactor conditions. The results of this work showed the presence of a bubble superlattice within the U-7Mo grains that accommodated fission gases (e.g., Xe). The presence of this structure helps the U-7Mo exhibit a stable swelling behaviour during irradiation. Furthermore, TEM analysis showed that the Si-rich interaction layers that develop around the fuel particles at the U-7Mo/matrix interface during fuel plate fabrication and irradiation become amorphous during irradiation. An important question that remains to be answered about the irradiation behaviour of U-Mo dispersion fuels is how do more aggressive irradiation conditions affect the behaviour of fission gases within the U-7Mo fuel particles and in the amorphous interaction layers on the microstructural scale that can be characterized using TEM? This paper will discuss the results of TEM analysis that was performed on a sample taken from an irradiated RERTR-7 fuel plate with Al-2Si matrix. This plate was exposed to more aggressive irradiation conditions than the RERTR-6 plate. The microstructural features present within the U-7Mo and the amorphous interaction layers will be discussed. The results of this analysis will be compared to what was observed in the earlier RERTR-6 fuel plate characterization. (author)

  10. Status of REBUS fuel management software development for RERTR applications

    International Nuclear Information System (INIS)

    Olson, Arne P.

    2000-01-01

    The REBUS-5 burnup code has evolved substantially in order to meet the needs of the ANL RERTR Program. This paper presents a summary of the past changes and improvements in the capabilities of this software, and also identifies future plans. (author)

  11. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    International Nuclear Information System (INIS)

    Wiencek, T.C.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched (∼93% 235 U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm 3 ) HEU fuel elements to highly loaded (up to 7 g U/cm 3 ) low-enrichment ( 235 U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing

  12. Summary report on fuel development and miniplate fabrication for the RERTR Program, 1978 to 1990

    Energy Technology Data Exchange (ETDEWEB)

    Wiencek, T.C. [Argonne National Lab., IL (United States). Energy Technology Div.

    1995-08-01

    This report summarizes the efforts of the Fabrication Technology Section at Argonne National Laboratory in the program of Reduced Enrichment Research and Test Reactors (RERTR). The main objective of this program was to reduce the amount of high enriched ({approx}93% {sup 235}U) uranium (HEU) used in nonpower reactors. Conversion from low-density (0.8--1.6 g U/cm{sup 3}) HEU fuel elements to highly loaded (up to 7 g U/cm{sup 3}) low-enrichment (<20% {sup 235}U) uranium (LEU) fuel elements allows the same reactor power levels, core designs and sizes to be retained while greatly reducing the possibility of illicit diversion of HEU nuclear fuel. This document is intended as an overview of the period 1978--1990, during which the Section supported this project by fabricating mainly powder metallurgy uranium-silicide dispersion fuel plates. Most of the subjects covered in detail are fabrication-related studies of uranium silicide fuels and fuel plate properties. Some data are included for out-of-pile experiments such as corrosion and compatibility tests. Also briefly covered are most other aspects of the RERTR program such as irradiation tests, full-core demonstrations, and technology transfer. References included are for further information on most aspects of the entire program. A significant portion of the report is devoted to data that were never published in their entirety. The appendices contain a list of previous RERTR reports, ANL fabrication procedures, calculations for phases present in two-phase fuels, chemical analysis of fuels, miniplate characteristics, and a summary of bonding runs made by hot isostatic pressing.

  13. RERTR-10 Irradiation Summary Report

    International Nuclear Information System (INIS)

    Perez, D.M.

    2011-01-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-10 was designed to further test the effectiveness of modified fuel/clad interfaces in monolithic fuel plates. The experiment was conducted in two campaigns: RERTR-10A and RERTR-10B. The fuel plates tested in RERTR-10A were all fabricated by Hot Isostatic Pressing (HIP) and were designed to evaluate the effect of various Si levels in the interlayer and the thickness of the Zr interlayer (0.001-inch) using 0.010-inch and 0.020-inch nominal foil thicknesses. The fuel plates in RERTR-10B were fabricated by Friction Bonding (FB) with two different thickness Si layers and Nb and Zr diffusion barriers.1 The following report summarizes the life of the RERTR-10A/B experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  14. RERTR-10 Irradiation Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Perez

    2011-05-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-10 was designed to further test the effectiveness of modified fuel/clad interfaces in monolithic fuel plates. The experiment was conducted in two campaigns: RERTR-10A and RERTR-10B. The fuel plates tested in RERTR-10A were all fabricated by Hot Isostatic Pressing (HIP) and were designed to evaluate the effect of various Si levels in the interlayer and the thickness of the Zr interlayer (0.001”) using 0.010” and 0.020” nominal foil thicknesses. The fuel plates in RERTR-10B were fabricated by Friction Bonding (FB) with two different thickness Si layers and Nb and Zr diffusion barriers.1 The following report summarizes the life of the RERTR-10A/B experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  15. Characterization and testing of monolithic RERTR fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D.; Jue, J.F.; Burkes, D.E. [Idaho National Lab., Idaho Falls, ID (United States)

    2007-07-01

    Monolithic fuel plates are being developed as a LEU (low enrichment uranium) fuel for application in research reactors throughout the world. These fuel plates are comprised of a U-Mo alloy foil encased in aluminum alloy cladding. Three different fabrication techniques have been looked at for producing monolithic fuel plates: hot isostatic pressing (HIP), transient liquid phase bonding (TLPB), and friction stir welding (FSW). Of these three techniques, HIP and FSW are currently being emphasized. As part of the development of these fabrication techniques, fuel plates are characterized and tested to determine properties like hardness and the bond strength at the interface between the fuel and cladding. Testing of HIP-made samples indicates that the foil/cladding interaction behavior depends on the Mo content in the UMo foil, the measured hardness values are quite different for the fuel, cladding, and interaction zone phase and Ti, Zr and Nb are the most effective diffusion barriers. For FSW samples, there is a dependence of the bond strength at the foil/cladding interface on the type of tool that is employed for performing the actual FSW process. (authors)

  16. Development of very-high-density fuels by the RERTR program

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

    1996-01-01

    The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

  17. The status of uranium-silicon alloy fuel development for the RERTR program

    International Nuclear Information System (INIS)

    Domagala, R.F.; Wiencek, T.C.; Thresh, H.R.; Stahl, D.

    1983-01-01

    As part of the national Reduced Enrichment Research and Test Reactor (RERTR) Program, Argonne National Laboratory (ANL) is engaged in a fuel-alloy development project. The fuel alloys are dispersed in an aluminum matrix and metallurgically roll-bonded within 6061 Al alloy. To date, 'miniplates' with up to 40 vol. fuel alloy have been successfully fabricated. Thirty-one of these plates have been or are being irradiated in the Oak Ridge Reactor (ORR). Three different fuels have been used in the ANL miniplates: U 3 Si (U + 4 wt.% Si), U 3 Si 2 (U + 7.4 wt.% Si), or ''U 3 SiAl'' (U + 3.5 wt.% Si + 1.5 wt.% Al). All three are candidates for permitting higher fuel loadings and thus lower enrichments of 235 U than would be possible with either UAl x or U 3 O 8 , the current fuels for plate-type elements. The enrichment level employed at ANL is ∼19.8%. Continuing effort involves the production of miniplates with up to ∼60 vol. % fuel, the development of a technology for full-size plate fabrication, and post-irradiation examination of miniplates already removed from the ORR. (author)

  18. RERTR-7 Irradiation Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    D. M. Perez; M. A. Lillo; G. S. Chang; G. A. Roth; N. E. Woolstenhulme; D. M. Wachs

    2011-12-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-7A, was designed to test several modified fuel designs to target fission densities representative of a peak low enriched uranium (LEU) burnup in excess of 90% U-235 at peak experiment power sufficient to generate a peak surface heat flux of approximately 300 W/cm2. The RERTR-7B experiment was designed as a high power test of 'second generation' dispersion fuels at peak experiment power sufficient to generate a surface heat flux on the order of 230 W/cm2.1 The following report summarizes the life of the RERTR-7A and RERTR-7B experiments through end of irradiation, including as-run neutronic analyses, thermal analyses and hydraulic testing results.

  19. JAERI RERTR program

    International Nuclear Information System (INIS)

    Sato, M.

    1983-01-01

    In 1979, Japan Atomic Energy Research Institute (JAERI) has started a five year program for the utilization of reduced enrichment uranium fuel, in place of currently used highly enriched uranium fuel for the JAERI Research and Test Reactors (RERTR), such as the JRR-2 (10 MWt), the JRR-4 35 MWt), the JMTR(50 MWO and the JMTRC (100 Wt, nuclear mockup of the JMTR), for contributing to the reduction of proliferation concerns. In order to assess the feasibility of converting the JAERI reactors to use of fuel with reduced enrichment uranium, ANL and JAERI have embarked on a joint study program since January 1980. This document provides outlines of the JAERI RERTR Program and of the ANL-JAERI Joint Study, and main results of JAERI's Phase A report which was made under the joint study

  20. RERTR activities in Argentina

    International Nuclear Information System (INIS)

    Adelfang, P.; Alvarez, L.; Pasqualini, E.

    2002-01-01

    The Atomic Energy Commission of Argentina has been an active participant of the RERTR program since 1978. An important milestone of the Argentinean RERTR program was the development and manufacturing at industrial scale of U 3 O 8 dispersed fuel and its subsequent application to the conversion of the RA-3 reactor core to LEU fuel. More recently, our activities were focused on the development of U 3 Si 2 fuel with a density of 4.8 gU/cm 3 and the improvement of the manufacturing process of this type of fuel. The program to qualify CNEA as a supplier U 3 Si 2 dispersed fuels is scheduled to finalize by mid 2003. To hasten this program the main research reactor of Argentina, the RA3, raised its power from 5 MW to 8 Mw in October this year. This is an intermediate step in the program to increase RA-3 power to 10 MW Currently, one of the main objectives is to develop and qualify the technology for the production of high-density LEU fuel elements using U-Mo alloy. An original way to produce U-Mo powder (the HMD process) was developed and its being upgraded to plant scale production. Another significant progress was the development of LEU targets for the production of 99 Mo, in the form of miniplates prepared with dispersed LEU U-Al x . The shipment to USA of 207 MTR spent fuels containing US origin highly enriched uranium, successfully carried out in the end of 2000, is another remarkable achievement of the Argentinean RERTR program. This activity was carried out in the framework of the United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuels were fabricated in Argentina and used in the RA-3 reactor from 1968 to 1987. The inventory of the shipped HEU spent fuel consisted in 166 standard assemblies and 41 control assemblies. (author)

  1. RERTR-12 Insertion 1 Irradiation Summary Report

    International Nuclear Information System (INIS)

    Perez, D.M.; Lillo, M.A.; Chang, G.S.; Woolstenhulme, N.E.; Roth, G.A.; Wachs, D.M.

    2012-01-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications. RERTR-12 insertion 1 includes the capsules irradiated during the first two irradiation cycles. These capsules include Z, X1, X2 and X3 capsules. The following report summarizes the life of the RERTR-12 insertion 1 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  2. RERTR-12 Insertion 2 Irradiation Summary Report

    International Nuclear Information System (INIS)

    Perez, D.M.; Chang, G.S.; Wachs, D.M.; Roth, G.A.; Woolstenhulme, N.E.

    2012-01-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-12 was designed to provide comprehensive information on the performance of uranium-molybdenum (U-Mo) based monolithic fuels for research reactor applications.1 RERTR-12 insertion 2 includes the capsules irradiated during the last three irradiation cycles. These capsules include Z, Y1, Y2 and Y3 type capsules. The following report summarizes the life of the RERTR-12 insertion 2 experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.

  3. Proceedings of the international meeting on development, fabrication, and application of Reduced Enrichment fuels for Research and Test Reactors (RERTR). Base technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-08-01

    The international effort to develop new fuel materials and designs which will make it feasible to fuel research and test reactors throughout the world with low-enrichment uranium, instead of high-enrichment uranium, has made significant progress during the past year. This progress has taken place at research centers located in many different countries, and is of crucial interest to reactor operators and licensors whose geographical distribution is even more varied. It is appropriate, therefore, that international meetings be held periodically to foster direct communication among the specialists in this area. To achieve this purpose, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the third of a series which begun in 1978. The papers presented at this meeting were divided into sessions according to relevant subject: status of RERTR program and safety issues; development of new fuel types; testing of new fuel elements; specific reactor applications. These proceedings were edited by various members of the RERTR Program.

  4. Proceedings of the international meeting on development, fabrication, and application of Reduced Enrichment fuels for Research and Test Reactors (RERTR). Base technology

    International Nuclear Information System (INIS)

    1983-08-01

    The international effort to develop new fuel materials and designs which will make it feasible to fuel research and test reactors throughout the world with low-enrichment uranium, instead of high-enrichment uranium, has made significant progress during the past year. This progress has taken place at research centers located in many different countries, and is of crucial interest to reactor operators and licensors whose geographical distribution is even more varied. It is appropriate, therefore, that international meetings be held periodically to foster direct communication among the specialists in this area. To achieve this purpose, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the third of a series which begun in 1978. The papers presented at this meeting were divided into sessions according to relevant subject: status of RERTR program and safety issues; development of new fuel types; testing of new fuel elements; specific reactor applications. These proceedings were edited by various members of the RERTR Program

  5. The RERTR Program

    International Nuclear Information System (INIS)

    Travelli, A.

    1992-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1991 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1991 in collaboration with its many international partners

  6. Abstracts and Papers of the 2003 International RERTR Meeting

    International Nuclear Information System (INIS)

    2003-10-01

    The papers presented at the 25th International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR) were classified according to the following subjects: National programs covering shipment of HEU to countries of origin and national RERTR programs; Molybdenum 99 production by using LEU fuel; Development and fabrication of new type of LEU fuel elements; Reactor core conversion analysis; Safety, spent fuel and regulatory issues; spent fuel storage and management

  7. Abstracts and papers of the 1997 International RERTR Meeting

    International Nuclear Information System (INIS)

    1997-10-01

    The 1997 International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR) was sponsored by the Argonne National Laboratory and was held in Jackson Hole, Wyoming, U.S.A. on 5-10 October 1997. The abstracts and available papers that were presented at this meeting are dealing with: National RERTR Programs; Fuel Development, Testing and Evaluation; Safety Tests and Evaluations; Core Conversion Studies; New LEU Reactors; Mo-99 Production From LEU Fission; Spent Fuel Management; Fuel Cycle issues

  8. Status and progress of the RERTR program

    International Nuclear Information System (INIS)

    Travelli, A.

    1996-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1996 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1995 in collaboration with its many international partners. Significant progress has been made during the past year in each of the three areas that the delegates to last year's RERTR meeting chose to address, in their letter to President Clinton, as requiring special attention. (1) In the area of U.S. acceptance of spent fuel from foreign research reactors, a second shipment of 99 urgent-relief spent fuel elements was completed. The Final Environmental Impact Statement was published in February 1996, and the Record of Decision was published in May 1996. The first shipments under the Record of Decision, containing 280 spent fuel elements, were received at the Savannah River Site in September 1996. (2) In the area of advanced fuel development, adequate funding and guidance were received by the RERTR program in March 1996. Fuel development activities are now in progress, including procurement of equipment, screening of candidate materials, and preparations for the production of a first series of microplates. The first irradiations are planned to begin in the Advanced Test Reactor. in Idaho, during April 1997. (3) In the area of conversion of DOE research reactors, the RERTR program has been tasked by the Department of Energy to assess the feasibility of converting to LEU fuel each of the DOE research reactors which currently use HEU fuel. A preliminary assessment will be presented at this meeting. Significant progress has been made by the Russian RERTR program, which aims to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with excellent results

  9. Abstracts and papers of the 1999 International RERTR Meeting

    International Nuclear Information System (INIS)

    1999-10-01

    The papers presented at the 22nd International RERTR Meeting dealt with the following topics: development and testing of new fuel elements (uranium-molybdenum alloys); research reactors core conversion studies (change from highly to moderately or slightly enriched uranium), including both measurements and calculations: spent fuel storage and transportation; production of 99 Mo from low enriched uranium. A number of papers were devoted to the status and future of national RERTR programs

  10. Review and summary of RERTR '98

    International Nuclear Information System (INIS)

    Gruber, Gerhard J.

    1999-01-01

    The RERTR program started in 1978 and had its last annual international meeting in Sao Paulo, Brazil in October 1998. The target of this program is to eliminate the risk of proliferation and misuse of the HEU by converting research reactors from using HEU to LEU. 36 research reactors are in the process of conversion (8) or fully converted (28) on the basis oft the existing U 3 Si 2 (4.8 gU/cc) fuel. Higher density fuels up to 8-9 gU/cc on a U-Mo basis are being developed to also achieve the conversion of high performance research reactors. The U.S. spent nuclear fuel return program is under good progress. Efforts are being made to establish similar RERTR and spent fuel return programs in Russia. Concerns about missing global or regional back-end solutions mainly for countries without a nuclear program were raised. (author)

  11. Abstracts and papers of the 2002 International RERTR Meeting

    International Nuclear Information System (INIS)

    2002-11-01

    This 24th RERTR meeting has been organized in cooperation with the IAEA. It was attended by almost 200 participants coming from almost 30 different countries. Among them we have representatives from governmental and international organizations, private companies, reactor operators, fuel developers, regulators. The papers presented during following sessions: national and international RERTR programs, fuel development, manufacturing and qualification; fuel performance; HEU and LEU fuel cycle; LEU target development; licensing, safety and core assessment; reactor core conversion; acceptance program and spent fuel transportation; spent fuel management

  12. Status of the US RERTR Program

    International Nuclear Information System (INIS)

    Travelli, A.

    1995-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1994 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1993 in collaboration with its many international partners. The RERTR Program has moved aggressively to support President Clinton's nonproliferation policy and his goal open-quotes to minimize the use of highly-enriched uranium in civil nuclear programsclose quotes. An Environmental Assessment which addresses the urgent-relief acceptance of 409 spent fuel elements was completed, and the first shipment of spent fuel elements is scheduled for this month. An Environmental Impact Statement addressing the acceptance of spent research reactor fuel containing enriched uranium of U.S. origin is scheduled for completion by the end of June 1995. The U.S. administration has decided to resume development of high-density LEU research reactor fuels. DOE funding and guidance are expected to begin soon. A preliminary plan for the resumption of fuel development has been prepared and is ready for implementation. The scope and main technical activities of a plan to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels was agreed upon by the RERTR Program and four Russian institutes lead by RDIPE. Both Secretary O'Leary and Minister Michailov have expressed strong support for this initiative. Joint studies have made significant progress, especially in assessing the technical and economic feasibility of using reduced enrichment fuels in the SAFARI-I reactor in South Africa and in the Advanced Neutron Source reactor under design at ORNL. Significant progress was achieved on several aspects of producing 99 Mo from fission targets utilizing LEU instead of HEU to the achievement of the common goal

  13. Review and summary of RERTR'97

    International Nuclear Information System (INIS)

    Krull, W.

    1998-01-01

    The 1997 RERTR meeting was the 20th annual meeting. Time to review the status of the program and to discuss the planned RERTR program speed. After years of agony there is still no sufficient activity in conversion studies and conversion of research reactors. Flanking measures are absolutely necessary to convince operators who have converted their reactors that they are not sitting alone in the rain. Some of these measures are very satisfying, e.g. the take back program of US origin fuel (free of charge for low income countries). The development of higher density fuel allows reactor operators to believe that they are not the only one to pay the penalties. The development of fuel for Russian reactors must be followed by a similar take back program to reduce the stockpiles. The light is being switched on again on this program. (author)

  14. The RERTR program: A status report

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A [Argonne National Laboratory, Argonne, IL (United States)

    1995-07-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1991 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. Postirradiation fuel data have continued to be analyzed and interpreted to gain a better understanding of the behaviour of research reactor fuels under irradiation. Final reports of the Oak Ridge Research Reactor (ORR) Whole-Core Demonstration and of the Ford Nuclear Reactor (FNR) Whole-Core Demonstration were published, and a final report on RERTR fuel development activities is at the final draft stage. Several computer codes for the analysis of research reactors were successfully converted to run on SUN workstations. Final contributions to the IAEA Safety and Licensing Guidebook for Core Conversions were completed. Analyses, calculations, and safety evaluations were conducted to support both US and foreign research reactors in converting to the use of low enrichment uranium. Six additional reactors that used to need EU exports have become fully converted to LEU fuels, bringing to nine the total of such reactors. Two more U.S. university reactors were also fully converted, bringing their total to five. An approximate quantitative evaluation shows that the midpoint of the road to conversion of all reactors which used to require HEU exports has been passed, and last year's progress is consistent with the projection that most reactors which do not require development of new fuels could be converted within three years. The major current program goal is to work closely with the various reactor and fuel fabrication organizations, so that this projection becomes reality. International

  15. The RERTR program: A status report

    International Nuclear Information System (INIS)

    Travelli, A.

    1995-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1991 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners. The RERTR program, has concentrated its efforts on technology transfer and implementation activities consistent with the guidance received from the Department of Energy at the end of 1990. Postirradiation fuel data have continued to be analyzed and interpreted to gain a better understanding of the behaviour of research reactor fuels under irradiation. Final reports of the Oak Ridge Research Reactor (ORR) Whole-Core Demonstration and of the Ford Nuclear Reactor (FNR) Whole-Core Demonstration were published, and a final report on RERTR fuel development activities is at the final draft stage. Several computer codes for the analysis of research reactors were successfully converted to run on SUN workstations. Final contributions to the IAEA Safety and Licensing Guidebook for Core Conversions were completed. Analyses, calculations, and safety evaluations were conducted to support both US and foreign research reactors in converting to the use of low enrichment uranium. Six additional reactors that used to need EU exports have become fully converted to LEU fuels, bringing to nine the total of such reactors. Two more U.S. university reactors were also fully converted, bringing their total to five. An approximate quantitative evaluation shows that the midpoint of the road to conversion of all reactors which used to require HEU exports has been passed, and last year's progress is consistent with the projection that most reactors which do not require development of new fuels could be converted within three years. The major current program goal is to work closely with the various reactor and fuel fabrication organizations, so that this projection becomes reality. International

  16. Progress of the Russian RERTR program: Development of new-type fuel elements for Russian-built research reactors

    International Nuclear Information System (INIS)

    Vatulin, A. V.; Stetskiy, Y.A.; Mishunin, V.A.; Suprun, V.B.; Dobrikova, I.V.

    2002-01-01

    The new design of pin-type fuel elements and fuel assembly on their basis for Russian research reactors has been developed. The number of following activities has been performed: computational and experimental substantiation of fuel element design; development of fabrication process of fuel elements; manufacturing of experimental assembly for lifetime in-pile tests. The relevant fuel assemblies are considered to be perspective for usage as low-enriched fuel for Russian research reactors. (author)

  17. Progress on RERTR issues in Australia

    International Nuclear Information System (INIS)

    Ripley, M.I.; Horlock, K.W.

    2002-01-01

    Australia has long been involved with and sympathetic to the goals of the RERTR program. This overview paper gives a brief introduction to RERTR-related activities in Australia since RERTR-2000. (author)

  18. RERTR programme. French situation in 1996

    International Nuclear Information System (INIS)

    Guidez, J.; Ballagny, A.

    1996-01-01

    This paper summarizes the status of the RERTR Programme in France in 1996. The reactors which are affected by the RERTR Programme are those that consume a considerable amount of fuel. These are the neutron beam reactors Orphee (at Saclay) and RHF (at Grenoble) and the irradiation reactors Osiris (at Saclay) and Siloe (at Grenoble). Plans to construct a new 100 MW, multi-purpose MTR reactor using LEU fuel at the Atomic Research Center at Cadarache are described. Two main topics of fuel research and development are summarized: (1) improving knowledge on the reference silicide fuel, including optimization of the manufacturing process, thermal properties, behavior with cladding failure, etc., and (2) research into low-enriched uranium fuel with a higher U-235 content in order to limit fuel assembly consumption and to improve performance levels. Tests planned in the research and development programme for the silicide fuel are tabulated. The fuel cycle option adopted by the CEA to get rid of spent fuel elements is reprocessing. The 'Caramel' fuel elements consumed in the Osiris reactor until 1995 are currently being reprocessed in CEA facilities at Marcoule. The UAl and UAlx fuel elements irradiated in Siloe and Orphee are currently being reprocessed in the Cogema facilities at Marcoule (Plant UP1). However, the U3Si2 fuel elements irradiated in the Osiris reactor since 1995 will be progressively sent to interim storage dry facilities located at Cadarache. (author)

  19. Post-irradiation analysis of low enriched U-Mo/Al dispersions fuel miniplate tests, RERTR 4 and 5

    International Nuclear Information System (INIS)

    Hofman, G.L.; Finlay, M.R.; Kim, Y.S.

    2005-01-01

    Interpretation of the post irradiation data of U-Mo/Al dispersion fuel mini plates irradiated in the Advanced Test Reactor to a maximum U-235 burn up of 80% are presented. The analyses addresses fuel swelling and porosity formation as these fuel performance issues relate to fuel fabrication and irradiation parameters. Specifically, mechanisms involved in the formation of porosity observed in the U-Mo/Al interaction phase are discussed and, means of mitigating or eliminating this irradiation phenomenon are offered. (author)

  20. Progress of the RERTR programme in 1999

    International Nuclear Information System (INIS)

    Travelli, A.

    1999-01-01

    This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during 1999 and discusses planned activities for the coming year. The past year was characterized by exceptionally important accomplishments and events for the RERTR program. - Three additional shipments containing 1,006 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,237 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. - Postirradiation examinations of the first two batches of microplates revealed good irradiation behavior of U-6Mo, and excellent irradiation behavior of U-Mo alloys with higher Mo content or with small Ru additions. Irradiation of a new batch of microplates to investigate the behavior of these fuels at high temperatures is scheduled to begin in October 1999. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g /cm 3 range. - Progress on irradiation testing and safety analyses was made in the Russian RERTR program, which aims to develop and demonstrate the technical means needed to convert Russian supplied research reactors to LEU fuels. - The U.S. Government has decided to aggressively pursue, in cooperation with the Russian Government, eventual conversion of three Russian plutonium production reactors to the use of low-enriched UO 2 -Al dispersion fuel. This effort is now proceeding, with assistance from RERTR personnel. - At the request of the German government, the RERTR program has addressed the performance of a new alternative LEU FRM-II core design that could be installed in the same building structure erected for the current 20 MW HEU design with the same 50-day fuel life. The results have been favorable. - Significant improvements were made in the design of an LEU metal-foil target

  1. Recent developments of the US RERTR program

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-01-01

    The status of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief outline of the RERTR Program objectives, goals and past accomplishments, emphasis is placed on the developments which took place during 1983 and on current program plans and schedules. Most program activities have proceeded as planned and a combination of two silicide fuels (U 3 Si 2 -Al and U 3 Si-Al) was found to hold excellent promise for achieving the long-term program goals. A modification of the program plan, including the development and demonstration of those fuels, was prepared and is now being implemented. The uranium density of qualified RERTR fuels for plate-type reactors is forecasted to grow by approximately 1 g U/cm 3 each year, from the current 1.7 g U/cm 3 to the 7.0 g U/cm 3 which will be reached in 1988. The technical needs of research reactors for HEU exports are also forecasted to undergo a gradual and dramatic decline in the coming years

  2. Progress of the RERTR program in 2001

    International Nuclear Information System (INIS)

    Travelli, A.

    2002-01-01

    This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of mini plates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm 3 range. Qualification of the U-Mo dispersion fuels has been declared by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm 3 are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid 2005. U-Mo fuel with uranium density of 8-9 g/cm 3 is expected to be qualified by mid 2007. Final irradiation tests of LEU 99 Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/ AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO 2 dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIJNM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September 30, 2001. Altogether, 4'562 spent fuel assemblies from foreign research

  3. Progress of the RERTR program in 2001

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A. [Technology Development Division Argonne National Laboratory, 9700 South Cass Avenue, Argonne, Illinois 60439-4841 (United States)

    2002-07-01

    This paper describes the 2001 progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners. Postirradiation examinations of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of mini plates of greater sizes was completed in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g/cm{sup 3} range. Qualification of the U-Mo dispersion fuels has been declared by a patent issue involving KAERI. Test fuel elements with uranium density of 6 g/cm{sup 3} are being fabricated by BWXT and are expected to begin undergoing irradiation in the HFR-Petten reactor around March 2003, with a goal of qualifying this fuel by mid 2005. U-Mo fuel with uranium density of 8-9 g/cm{sup 3} is expected to be qualified by mid 2007. Final irradiation tests of LEU {sup 99}Mo targets in the RAS-GAS reactor at BATAN, in Indonesia, had to be postponed because of the 9/11 attacks, but the results collected to date indicate that these targets will soon be ready for commercial production. Excellent cooperation is also in progress with the CNEA in Argentina, MDSN/ AECL in Canada, and ANSTO in Australia. Irradiation testing of five WWR-M2 tube-type fuel assemblies fabricated by the NZChK and containing LEU UO{sub 2} dispersion fuel was successfully completed within the Russian RERTR program. A new LEU U-Mo pin-type fuel that could be used to convert most Russian-designed research reactors has been developed by VNIJNM and is ready for testing. Four additional shipments containing 822 spent fuel assemblies from foreign research reactors were accepted by the U.S. by September 30, 2001. Altogether, 4'562 spent fuel

  4. United States policy initiatives in promoting the RERTR program

    International Nuclear Information System (INIS)

    Huizenga, David G.

    1996-01-01

    The Reduced Enrichment for Research and Test Reactors (RERTR) program has been successful in furthering efforts to reduce and eventually eliminate highly enriched uranium (HEU) from international commerce. Three key policy initiatives are underway to further promote the RERTR program. The first initiative is implementation of a new nuclear weapons nonproliferation policy concerning foreign research reactor spent nuclear fuel. Under this policy, the United States will accept over the next 13 years research reactor spent fuel from 41 countries that have converted or plan to convert to use LEU fuels. The second initiative is to pursue cooperative efforts to expand the RERTR program to new regions of the globe, including Russia and China. The third initiative is to restart the advanced LEU fuels development program at the Argonne National Laboratory in order to increase the number of reactors that can convert to use LEU without significant detriment to their performance

  5. The RERTR program status and progress

    International Nuclear Information System (INIS)

    Travelli, A.

    2004-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1995 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 994 in collaboration with its many international partners. The revelation that Iraq was on the verge of developing a nuclear weapon at the time of the Gulf War, and that it was planning to do so by extracting HEU from the fuel of its research reactors, has given new impetus and urgency to the RERTR commitment of eliminating HEU use in research and test reactors worldwide. The Draft Environmental Impact Statement on the DOE policy for managing spent nuclear fuels from foreign research reactors was published on schedule. A Record of Decision is due in December 1995. After much legal debate, a first shipment of 153 urgent-relief elements took place, and another 151 elements became eligible for shipment. Development of advanced LEU research reactor fuels is scheduled to begin in October 1995, after DOE funding is received. Funding for equipment needed to begin this activity was provided by the US Department of State, and procurement of the equipment is in progress. The Russian RERTR program, which aims to develop and demonstrate within the next five years the technical means needed to convert Russian-supplied research reactors to LEU fuels, is now in operation. A Statement of Intent was signed by high US and Chinese officials, endorsing cooperative activities between the RERTR program and Chinese laboratories involved in similar activities. Joint studies of LEU technical feasibility were completed for the SAFARI-1 reactor in South Africa and for the ANS reactor in the US. The ANS project was later canceled for budgetary and nonproliferation reasons. A new study has been initiated for the FRM-II reactor in Germany. Significant progress was made on several aspects of producing 99 Mo from fission targets utilizing LEU instead of

  6. Major results on the development of high density U-Mo fuel and pin-type fuel elements executed under the Russian RERTR program and in cooperation with ANL (USA)

    International Nuclear Information System (INIS)

    Vatulin, A.; Morozov, A.; Stetsky, Y.; Suprun, V.; Dobrikova, I.; Trifonov, Y.; Mishunin, V.; Sorokin, V.

    2003-01-01

    VNIINM is active participant of 'Russian program on Reduced Enrichment for Research and Test Reactors'. Institute Works in two main directions: 1) development of new high-density fuels (HDF) and 2) development of new design of fuel elements with LEU. The development of the new type fuel element is carried out both for existing reactors, and for developing new advanced reactors. The 'TVEL' concern is coordinator of works of this program. The majority enterprises of branch (NIIAR, PIYaF, RRC KI, NZChK) take part in this work. Since 2000 these works are being conducted in cooperation with Argonne National Laboratory (USA) within the RERTR program under VNIINM with ANL contract. At the present, a large set of pre-pile investigations has been completed. All necessary fabrication procedures have been developed for utilization of U-Mo dispersion fuel in Russian-designed research reactors. For irradiation tests the pin-type mini-fuel elements with HDF dispersion fuel with LEU and the uranium density equaled to 4,0 and 6,0 g/cm 3 (up to 40 vol.%) have been manufactured. Their irradiation began in August 2003 in the MIR reactor (NIIAR, Dimitrovgrad). A large set of works for preparation of lifetime tests (WWR-M reactor in Gatchina) of two full-scale fuel assemblies with new pin-type fuel elements on basis LEU UO 2 -Al and UMo-Al fuels has been completed. The in-pile tests of fuel assemblies began in September 2003. The summary of important results of performed works and their near-term future are presented in paper. (author)

  7. Progress of Indonesia RERTR related programs

    International Nuclear Information System (INIS)

    Soentono, S.; Arbie, B.; Surpto, A.

    2004-01-01

    In Indonesia, there are two main activities covering study, research ad development which can be related to reduced enrichment for research and test reactors (RERTR) program. The first activity is the attempt to improve the G.A. Siwabessy multi-purpose reactor (RSG-GAS) performance following the successful RERTR program on the development of low-enriched uranium (LEU) with high uranium loading density. This activity consists of manufacturing technology development and fabrication of LEU fuel in the form Of U 3 Si 2 -Al with high U loading density being capable to reach high burn-up, study oil the core conversion of the RSG- GAS from using oxide LEU fuel with loading density of ∼3 gU/cc into using LEU silicide fuel of higher U loading density to improve the in-core fuel management, the fuel utilization, and the cycle length, while keeping the neutron flux in the irradiation facilities remaining satisfactory and the power of the reactor remaining the same without resulting any serious penalty on the reactor safety. The second RERTR related activity is the attempt to use LEU, instead of HEU, as the target or fission product 99 Mo production to supply 99 mTc for medical purposes. This second activity is expected to consist of experiments covering irradiation of various forms of LEU targets in the RSG-GAS, dissolution processes of the irradiated targets, separation and purification processes to meet the radiopharmaceutical requirements and recovery of 235 U from the waste as well as treatment of the emerged wastes. (author)

  8. The U.S. RERTR program status and progress

    International Nuclear Information System (INIS)

    Travelli, A.

    1998-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program since its inception in 1978 is described. A brief summary of the results which the RERTR Program had achieved by the end of 1996 in collaboration with its many international partners is followed by a detailed review of the major events, findings, and activities of 1997. Significant progress has been made during the past year. In the area of U.S. acceptance of spent fuel from foreign research reactors, several shipments have taken place and additional are being planned. Intense fuel development activities are in progress, including procurement of equipment, screening of candidate materials, and production of microplates. Irradiation of the first series of microplates began in August 1997 in the Advanced Test Reactor, in Idaho. Progress has been made in the Russian RERTR program, which aims to develop and demonstrate within five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, controversial performance issues which were raised at last year's meeting. Progress was also made on several aspects of producing molybdenum-99 from fission targets utilizing LEU instead of HEU. Various types of targets and processes are being pursued, with FDA approval of an LEU process projected to occur within two years. The feasibility of LEU Fuel conversion for three important DOE research reactors (BMRR, HFBR, and HFIR) has been evaluated by the RERTR program. In spite of the many momentous events which have occurred during the intervening years, and the excellent progress achieved, the most important challenges that the RERTR program faces today are not very different in type from those that were faced during the first RERTR meeting. Now, as then, the most important task is to develop new LEU fuels satisfying requirements which cannot be satisfied by any

  9. The U.S. RERTR Program. Status and progress

    International Nuclear Information System (INIS)

    Travelli, A.

    1997-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program since its inception in 1978 is described. A brief summary of the results which the RERTR Program had achieved by the end of 1996 in collaboration with its many international partners is followed by a detailed review of the major events, findings, and activities of 1997. Significant progress has been made during the past year. In the area of U.S. acceptance of spent fuel from foreign research reactors, several shipments have taken place and additional are being planned. Intense fuel development activities are in progress, including procurement of equipment, screening of candidate materials, and production of microplates. Irradiation of the first series of microplates began in August 1997 in the Advanced Test Reactor, in Idaho. Progress has been made in the Russian RERTR program, which aims to develop and demonstrate within five years the technical means needed to convert Russian-supplied research reactors to LEU fuels. The study of an alternative LEU core for the FRM-II design has been extended to address, with favorable results, controversial performance issues which were raised at last year's meeting. Progress was also made on several aspects of producing molybdenum-99 from fission targets utilizing LEU instead of HEU. Various types of targets and processes are being pursued, with FDA approval of an LEU process projected to occur within two years. The feasibility of LEU fuel conversion for three important DOE research reactors (BMRR, HFBR, and HFIR) has been evaluated by the RERTR program. In spite of the many momentous events which have occurred during the intervening years, and the excellent progress achieved, the most important challenges that the RERTR program faces today are not very different in type from those that were faced during the first RERTR meeting. Now, as then, the most important task is to develop new LEU fuels satisfying requirements which cannot be satisfied by any

  10. The congressional assault on RERTR

    International Nuclear Information System (INIS)

    Lyman, Edwin

    2003-01-01

    Since 9/11, there is a growing appreciation of the importance of the RERTR program and the effort to eliminate HEU commerce. That is why US DOE supports the 'Accelerated RERTR program'. But some legislative developments run contrary to this trend. This could result in loopholes that block completion of RERTR agenda. Current attempt in Congress si to weaken US HEU export controls by modifying the 'Schumer Amendment' - Right-wing criticism of DOE plan to convert US research reactors in the House of Representatives

  11. Status of RERTR activities in Argentina

    International Nuclear Information System (INIS)

    Alvarez, L.; Boero, N.; Pasqualini, E.; Echenique, P.; Ruggirello, G.; Taboada, H.

    2003-01-01

    CNEA, the Atomic Energy Commission of Argentina, has defined development and fabrication activities related with the main fields of the RERTR program. This includes the qualification of high-density fuels for research reactors, the production of radioisotopes from LEU targets and the reduction of the HEU inventory. This paper continues previous presentations in past RERTR meetings and summarizes the main activities performed in these fields during 2002- 2003 and future works. The main part of the CNEA U 3 Si 2 Qualification Program will be completed within this year. The irradiation of the first fuel element was finished and the activities are mainly focused on its PIE works. Regarding the UMo field, development works to provide support for fuel plate fabrication process and to improve understanding of UMo-Al interaction are on progress. Depending on the results of the fuel plate fabrication setup, the manufacturing of full size FE for UMo qualification program could take place during 2004. A program related with the UMo monolithic fuel is also in progress. In the radioisotopes field LEU targets have replaced HEU-Al alloy targets for routine Mo-99 production since November 2002. CNEA approach to minimize the Argentine HEU inventory is also presented (author)

  12. The RERTR program

    International Nuclear Information System (INIS)

    Travelli, A.

    1997-01-01

    The Reduced Enrichment Research and Test Reactor (RERTR) Program was established in 1978 at the Argonne National Laboratory (ANL) by the Department of Energy (DOE), which continues to fund the program and to manage it in coordination with the Department of State (DOS), the Arms Control and Disarmament Agency (ACDA), and the Nuclear Regulatory Commission (NRC). The primary objective of the program is to develop the technology needed to use Low-Enrichment Uranium (LEU) instead of High-Enrichment Uranium (HEU) in research and test reactors, without significant penalties in experiment performance, economics, or safety. Eliminating the continuing need of HEU supplies for research and test reactors has long been an integral part of US nonproliferation policy. This paper reviews the main accomplishments of the program through the years

  13. The U.S. RERTR Program: Overview, status and plans

    International Nuclear Information System (INIS)

    Travelli, A.

    1985-01-01

    The status of the Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a brief review of the accomplishments which the RERTR Program, in cooperation with its many international partners, had achieved by the end of 1983 in the area of LEU research reactor fuels development and application, emphasis is placed on the RERTR Program developments which took place during 1984 and on current plans and schedules. The RERTR progress in 1984 has been significant, with solid accomplishments and few surprises. Most LEU U 3 Si 2 -AI irradiation tests with 48 g U/cm 3 have been successfully completed, and contract negotiations are under way for the procurement of a whole-core demonstration of this fuel in the ORR. The demonstration is to begin in mid-1985 and to last for approximately eighteen months. Qualification of U 3 Si-AI fuel with 7 g U/cm 3 is scheduled for 1989. International cooperation among fuel developers, commercial vendors, and reactor operators has been essential to the progress which has been achieved. With continued international cooperation, it will be feasible to significantly reduce HEU usage in research reactors in the next few years. (author)

  14. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W.; Ross Finlay, M.; Moore, Glenn; Medvedev, Pavel; Meyer, Mitch

    2017-05-01

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U-Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

  15. Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500 °C

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Dennis D., E-mail: dennis.keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Jue, Jan-Fong; Gan, Jian; Miller, Brandon D.; Robinson, Adam B.; Madden, James W. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States); Ross Finlay, M. [Australian Nuclear Science and Technology Organization, PMB 1, Menai, NSW 2234 (Australia); Moore, Glenn; Medvedev, Pavel; Meyer, Mitch [Nuclear Fuels and Materials Division, Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID 83415-6146 (United States)

    2017-05-15

    The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research and test reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500 °C. The results indicated that two types of grain/cell boundaries were observed in the U-7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Finally, the fission gas bubbles that were originally around 3 nm in diameter and resided on a fission gas superlattice (FGS) in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ∼20 nm diameter) during blister testing and, in many areas, are no longer organized as a superlattice.

  16. Abstracts and papers presented at the 2004 international RERTR meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    Oral and poster presentations of the Meeting covered the following topics: National and international programs related to Reduced Enrichment for Research and Test Reactors (RERTR). Development of new fuel types, testing, fabrication, modelling. Studies of reactor cores conversion from highly enriched to low enriched fuel, including licensing. New and converted reactors. Spent fuel management including storage and transportation. Production of Molybdenum 99 under converted core conditions.

  17. Abstracts and papers presented at the 2004 international RERTR meeting

    International Nuclear Information System (INIS)

    2005-01-01

    Oral and poster presentations of the Meeting covered the following topics: National and international programs related to Reduced Enrichment for Research and Test Reactors (RERTR). Development of new fuel types, testing, fabrication, modelling. Studies of reactor cores conversion from highly enriched to low enriched fuel, including licensing. New and converted reactors. Spent fuel management including storage and transportation. Production of Molybdenum 99 under converted core conditions

  18. Detailed measurements of local thickness changes for U-7Mo dispersion fuel plates with Al-3.5Si matrix after irradiation at different powers in the RERTR-9B experiment

    Science.gov (United States)

    Keiser, Dennis D.; Williams, Walter; Robinson, Adam; Wachs, Dan; Moore, Glenn; Crawford, Doug

    2017-10-01

    The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.

  19. Microstructural Characterization of the U-9.1Mo Fuel/AA6061 Cladding Interface in Friction-Bonded Monolithic Fuel Plates Irradiated in the RERTR-6 Experiment

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Medvedev, Pavel; Madden, James; Wachs, Dan; Clark, Curtis; Meyer, Mitch

    2015-09-01

    Low-enrichment (235U < 20 pct) U-Mo monolithic fuel is being developed for use in research and test reactors. The earliest design for this fuel that was investigated via reactor testing consisted of a nominally U-10Mo fuel foil encased in AA6061 (Al-6061) cladding. For a fuel design to be deemed adequate for final use in a reactor, it must maintain dimensional stability and retain fission products throughout irradiation, which means that there must be good integrity at the fuel foil/cladding interface. To investigate the nature of the fuel/cladding interface for this fuel type after irradiation, fuel plates were fabricated using a friction bonding process, tested in INL's advanced test reactor (ATR), and then subsequently characterized using optical metallography, scanning electron microscopy, and transmission electron microscopy. Results of this characterization showed that the fuel/cladding interaction layers present at the U-Mo fuel/AA6061 cladding interface after fabrication became amorphous during irradiation. Up to two main interaction layers, based on composition, could be found at the fuel/cladding interface, depending on location. After irradiation, an Al-rich layer contained very few fission gas bubbles, but did exhibit Xe enrichment near the AA6061 cladding interface. Another layer, which contained more Si, had more observable fission gas bubbles. In the samples produced using a focused ion beam at the interaction zone/AA6061 cladding interface, possible indications of porosity/debonding were found, which suggested that the interface in this location is relatively weak.

  20. RERTR-12 Post-irradiation Examination Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Rice, Francine [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williams, Walter [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robinson, Adam [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason [Idaho National Lab. (INL), Idaho Falls, ID (United States); Meyer, Mitch [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabin, Barry [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    The following report contains the results and conclusions for the post irradiation examinations performed on RERTR-12 Insertion 2 experiment plates. These exams include eddy-current testing to measure oxide growth; neutron radiography for evaluating the condition of the fuel prior to sectioning and determination of fuel relocation and geometry changes; gamma scanning to provide relative measurements for burnup and indication of fuel- and fission-product relocation; profilometry to measure dimensional changes of the fuel plate; analytical chemistry to benchmark the physics burnup calculations; metallography to examine the microstructural changes in the fuel, interlayer and cladding; and microhardness testing to determine the material-property changes of the fuel and cladding.

  1. Abstracts and papers of the 2000 International RERTR Meeting

    International Nuclear Information System (INIS)

    2000-10-01

    This meeting was devoted to progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000. The main activities planned for the year 2001 were discussed. Year 2000 was held by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Qualification of the U-Mo dispersion fuels is proceeding on schedule. Test fuel elements with uranium density of 6 g/cm3 are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo fuel with uranium density of 8-9 g/cm3 is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission 99Mo. Progress was made on irradiation testing of LEU UO 2 dispersion fuel and on LEU conversion feasibility studies in the Russian RERTR program. Conversion of the BER-II reactor in Germany, was completed and conversion of the La Reina reactor in Chile, began. In the fuel development area, the RERTR program is pursuing qualification of high-density LEU U-Mo dispersion fuels, with the dual goal of enabling further conversions and of developing a substitute for LEU silicide fuels that can be more easily disposed of after expiration of the FRR SNF Acceptance Program. The 99Mo effort has reached the point where it appears feasible for all the 99Mo producers of the world to agree jointly to a common course of action leading to the elimination of HEU use in their processes. It was concluded

  2. Status and progress of the RERTR program in the year 2002

    International Nuclear Information System (INIS)

    Travelli, A.; Technology Development

    2003-01-01

    Following the cancellation of the 2001 International RERTR Meeting, which had been planned to occur in Bali, Indonesia, this paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the years 2001 and 2002, and discusses the main activities planned for the year 2003. The past two years have been characterized by very important achievements of the RERTR program, but these technical achievements have been overshadowed by the terrible events of September 11, 2001. Those events have caused the U.S. Government to reevaluate the importance and urgency of the RERTR program goals. A recommendation made at the highest levels of the government calls for an immediate acceleration of the program activities, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors

  3. Status of the RERTR program: overview, progress and plans

    International Nuclear Information System (INIS)

    Travelli, A.

    1985-01-01

    The status of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a summary of the accomplishments which the RERTR Program had achieved by the end of 1984 with its many international partners, emphasis is placed on the progress achieved during 1985 and on current plans and schedules. A new miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels, was fabricated and is well into irradiation. The whole-core ORR demonstration is scheduled to begin in November 1985, with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 . Altogether, 921 full-size test and prototype elements have been ordered for fabrication with reduced enrichment and the new technologies. Qualification of U 3 Si-Al fuel with approx.7 g U/cm 3 is still projected for 1989. This progress could not have been achieved without the close international cooperation which has existed since the beginning, and whose continuation and intensification will be essential to the achievement of the long-term RERTR goals

  4. The status of the RERTR Program: Overview, progress and plans

    International Nuclear Information System (INIS)

    Travelli, A.

    1985-01-01

    The status of the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program is reviewed. After a summary of the accomplishments which the RERTR Program had achieved by the end of 1984 with its many international partners, emphasis is placed on the progress achieved during 1985 and on current plans and schedules. A new miniplate series, concentrating on U 3 Si 2 -AI and U 3 Si-AI fuels, was fabricated and is well into irradiation. The whole-core ORR demonstration is scheduled to begin in November 1985, with U 3 Si 2 -AI fuel at 4.8 g U/cm 3 . Altogether, 921 full-size test and prototype elements have been ordered for fabrication with reduced enrichment and the new technologies. Qualification of U 3 Si-AI fuel with ∼7 g U/cm 3 is still projected for 1989. This progress could not have been achieved without the close international cooperation which has existed since the beginning, and whose continuation and intensification will be essential to the achievement of the long-term RERTR goals. (author)

  5. RERTR activities in Argentina

    International Nuclear Information System (INIS)

    Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; De La Fuente, M.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Perez, A.; Piazza, A.; Ruggirello, G.; Taboada, H.

    2000-01-01

    The Atomic Energy Commission of Argentina has been involved in the Reduced Enrichment for Research and Test Reactors Program since 1978. The most relevant milestones of the program, regarding fuel R and D activities, were the development and manufacturing at industrial scale of U 3 O 8 dispersed fuel assemblies and the conversion of the RA-3 reactor core to LEU fuel. More recently, the activities were focused in the development of high density U 3 Si 2 fuel with a density of 4.8 gU/cm 3 and the improvement of the manufacturing process of U 3 Si 2 powder. Currently one of the main objectives is to develop and qualify the technology for the production of high-density LEU fuel elements using U-Mo alloy. Several alternative ways to obtain U-Mo powder are under development with the aim of evaluating plant scale production and costs. To boost this program the main research reactor of Argentina, the RA-3, will be upgraded to 10 MW early in 2001 and the hot cells at the Ezeiza Atomic Center are fully operational after important investments. Significant progresses were also carried out in the development of LEU targets for the production of Mo 99 . Experimental work has demonstrated the feasibility of the manufacturing and radiochemical processing of miniplate targets prepared with dispersed UAl x , maintaining the geometry and the alkaline processing of the HEU targets used so far. (author)

  6. Current status of the RERTR program

    International Nuclear Information System (INIS)

    Travelli, Armando

    1983-01-01

    The main purpose of this paper is to review the progress which has been accomplished by the RERTR Program during the past year, the present status of the program, the activities which are now in progress, and the program plans for the coming years. The main program objectives have not changed during the past year. The primary goal of the RERTR Program is still to reduce the amount of HEU which is shipped and used around the world in research reactors and, with it, the nuclear proliferation potential which it creates. Also unchanged is the program objective to reduce HEU use without significant penalty to the reactors involved. This means that for every possible conversion, our technical goal is to achieve good experiment capabilities and the economic advantages that come from a long core lifetime and from relatively economical fuels. We also intend to limit as much as possible the modifications which may have to be made to the reactor in order to achieve conversion and, last but not least, we intend to avoid any significant problems that might appear in the safety and licensing areas. Other important goals are the sharing of information about reduced-enrichment technology and providing technical assistance to the U.S. Government on special problems that may affect implementation of its policies. The timetable of the RERTR Program has stretched slightly. At the time of the previous ANL Conference, a long range goal was set which was anticipated to require about five or six years. This happened in 1978 and, therefore, the long range goal was considered to be achievable by 1984. This final deadline, the time when the long term goal of technical convertibility of all research reactors can be achieved, has now been moved to 1986. In the meantime, good progress has been made towards the near-term goals. Much more detailed knowledge is available about what lies ahead from a technical point of view. In addition, the general level of confidence in the achievability of the

  7. Status and progress of the RERTR Program

    International Nuclear Information System (INIS)

    Travelli, A.

    1996-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1996 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1995 in collaboration with its many international partners

  8. Progress of the RERTR program in 1989

    International Nuclear Information System (INIS)

    Travelli, A.

    1991-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1988, the major events, findings, and activities of 1989 are reviewed. (orig.)

  9. Status and progress of the RERTR Program

    International Nuclear Information System (INIS)

    Travelli, A.

    1993-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1993 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1992 in collaboration with its many international partners

  10. The RERTR Program: A status report

    International Nuclear Information System (INIS)

    Travelli, A.

    1991-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. The major events, findings, and activities of 1991 are reviewed after a brief summary of the results which the RERTR Program had achieved by the end of 1990 in collaboration with its many international partners

  11. The RERTR [Reduced Enrichment Research and Test Reactor] program:

    International Nuclear Information System (INIS)

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) program is described. After a brief summary of the results which the RERTR program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results and new developments which ocurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40 % average burnup. Good progress was made in the area of LEU usage for the production of fission 99 Mo, and in the coordination of safety evaluations related to LEU conversions of U.S. university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U 3 Si-Al with 19.75 % enrichment and U 3 Si 2 -Al with 45 % enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR program. (Author)

  12. Progress of the RERTR program in 1999

    International Nuclear Information System (INIS)

    Travelli, A.

    1999-01-01

    This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during 1999 and discusses planned activities for the coming year

  13. Status and progress of the RERTR program in the year 2004

    International Nuclear Information System (INIS)

    Travelli, A.

    2005-01-01

    The overall status of the RERTR program at the time of the last RERTR meeting is reviewed and the progress achieved since that meeting is described. In the fuel area, unexpected failures of LEU U-Mo dispersion plates and tubes under irradiation testing have prompted a revision of the plans to qualify these fuels. While potential solutions to the difficulties with U-Mo dispersion fuels are being explored in collaboration with our international partners, greater emphasis has been placed on accelerating development of monolithic LEU U-Mo fuel. The feasibility of converting several Russian-designed research reactors to LEU fuels has been addressed, and progress has been made in the development of LEU based 99 Mo production processes. The Russian RERTR program has made significant advances. A very important event of 2004 was the US DOE establishment of the Global Threat Reduction Initiative (GTRI). This new program accelerates and combines under the same US DOE management several programs, including RERTR, which aim to secure, remove, or dispose of, nuclear and other radioactive materials throughout the world that are vulnerable to theft by terrorists. (author)

  14. Abstracts and available papers presented at the 2005 International RERTR Meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    Oral and poster presentations of the Meeting covered the following topics: National and international programs related to Reduced Enrichment for Research and Test Reactors (RERTR). Development of new fuel types, testing, fabrication, modelling. Studies of reactor cores conversion from highly enriched to low enriched fuel, including licensing. New and converted reactors. Spent fuel management including storage and transportation. Production of Molybdenum 99 under converted core conditions.

  15. The RERTR program: a status report

    International Nuclear Information System (INIS)

    Travelli, A.

    1998-01-01

    This paper describes the progress achieved by the reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners since its inception in 1978. A brief summary of the results that the program had attained by the end of 1977 is followed by a detailed review of the major events, findings, and activities that took place in 1998. (author)

  16. Current status of the US RERTR Program

    International Nuclear Information System (INIS)

    Travelli, A.

    1982-01-01

    This paper describes the status of the RERTR Program. The main objective is to provide the audience with some orientation and sense of perspective that may assist in viewing the other program presentations as part of an overall effort, rather than as separate and unrelated activities

  17. A premature demise for RERTR [Reduced Enrichment for Research and Test Reactors programme]?

    International Nuclear Information System (INIS)

    Rydell, R.J.

    1990-01-01

    A common commitment from France, Belgium, Germany and the US to eliminate highly enriched uranium from their research reactors is needed to help guard against this material falling into the wrong hands. In the US, an essential part of this commitment would be rekindling the weakened Reduced Enrichment for Research and Test Reactors programme (RERTR). This is an American initiative to develop low-enrichment uranium fuel for research reactors that have previously required weapons-usable material. Underway since 1978 at Argonne National Laboratory, RERTR has achieved some impressive results: the development of higher density, low enriched fuels that are suitable for use at over 90% of the world's research reactors; a net reduction of US exports of highly enriched uranium (HEU) from the annual 700kg levels in the late 1970s to a 1990 level of just over 100kg; the encouragement of international scientific co-operation aimed at developing new fuels and facilitating the conversion of existing reactors to these fuels. However, in recent years, the US commitment to RERTR has been declining -budgets have fallen and advanced fuel development work has terminated. (author)

  18. Status and progress of the RERTR program in the year 2002

    International Nuclear Information System (INIS)

    Travelli, Armando

    2002-01-01

    Following the cancellation of the 2001 International RERTR Meeting, which had been planned to occur in Bali, Indonesia, this paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the years 2001 and 2002, and discusses the main activities planned for the year 2003. The past two years have been characterized by very important achievements of the RERTR program, but these technical achievements have been overshadowed by the terrible events of September 11, 2001. Those events have caused the U.S. Government to reevaluate the importance and urgency of the RERTR program goals. A recommendation made at the highest levels of the government calls for an immediate acceleration of the program activities, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The RERTR program has prepared and submitted to the Department of Energy a plan and a schedule to achieve this goal. The plan makes full use of two very important technical developments that have occurred within the program during the past two years: - Excellent results have been obtained from the irradiation of miniplates containing monolithic LEU U-Mo fuel with uranium density of 15.6 g/cm 3 . If an economically viable manner of fabricating monolithic LEU U-Mo fuel elements is developed, and if the preliminary irradiation tests are confirmed, this fuel holds the promise of enabling LEU operation of all existing and future research reactors in combination with unprecedented performance. - The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of most Russian-designed research and test reactors. The plan is structured to achieve LEU

  19. 2004 status of RERTR activities of CNEA - Argentina

    International Nuclear Information System (INIS)

    Audero, M.; Balart, S.; Boero, N.; Cabot, P.; Manzini, A.; Pasqualini, E.; Ruggirello, G.; Taboada, H.

    2005-01-01

    During 2004 several activities related to RERTR topics has taken place. In what follows, a resume of those activities is presented. During 2004 the Nuclear Fuel Cycle Program has supported the activities of about a hundred of researchers and technicians involved on different aspects of Very high density fuel developpement (VHD) fuel development as R and D, mini-plates, plates and fuel assemblies fabrication, irradiation and PIE analyses of VHD fuels. CNEA's LEU MTR fuel waste disposal plans are concerned with spent fuel from the reactor RA-3. The SF will be kept in interim storage until a geological repository is available for disposal. According to the current planning, this repository would be available in the year 2050 to receive the SF from the NPPs or the HLW from their reprocessing. Since 2002 the domestic production of Mo99 and other fission radioisotopes is performed using LEU targets, fulfilling international technical and quality standards. Up to present more than 600 targets bearing LEU material have been fabricated irradiated and processed. Very recently CNEA has also become an international supplier of targets for Mo-99 production. This status report deals as well with the International framework i.e. Global Threat Reduction Initiative and the historical involvement of Argentina in the peaceful applications of nuclear energy

  20. Using low-enriched uranium in research reactors: The RERTR program

    International Nuclear Information System (INIS)

    Travelli, A.

    1994-01-01

    The goal of the RERTR program is to minimize and eventually eliminate use of highway enriched uranium (HEU) in research and test reactors. The program has been very successful, and has developed low-enriched uranium (LEU) fuel materials and designs which can be used effectively in approximately 90 percent of the research and test reactors which used HEU when the program began. This progress would not have been possible without active international cooperation among fuel developers, commercial vendors, and reactor operators. The new tasks which the RERTR program is undertaking at this time include development of new and better fuels that will allow use of LEU fuels in all research and test reactors; cooperation with Russian laboratories, which will make it possible to minimize and eventually eliminate use of HEU in research reactors throughout the world, irrespective of its origin; and development of an LEU-based process for the production of 99 Mo. Continuation and intensification of international cooperation are essential to the achievement of the ultimate goals of the RERTR program

  1. 20 years of RERTR. Where we are and where to go?

    International Nuclear Information System (INIS)

    Krull, W.; Jager, W.

    1997-01-01

    This review will not survey the 20 years with all the efforts made in different countries to develop higher density fuel, to perform test irradiations and typical prototype fuel irradiation experiments, post irradiation examinations and actual core conversions. The emphasis is directed to the situation a few years ago when efforts have been rapidly slowing down, and the situation today after having had an international wake up call and comparing these efforts with the primary demand to reduce the proliferation risk worldwide and to have this as an indispensable goal of the international politics. The RERTR program is being back on the stony way. There are some (many?) hopes but many doubts to have the RERTR program a real success. Such a success needs the strong continue support and activities of many governments and especially of the IAEA for years. (author)

  2. South Africa and the RERTR programme. Current status. An AEC perspective

    International Nuclear Information System (INIS)

    Ball, G.

    1999-01-01

    The Atomic Energy Corporation (AEC) of South Africa fully endorses the principles of the RERTR programme. To this end in 1994 the AEC, together with Argonne National Laboratory (ANL), undertook a study into the feasibility of converting the SAFARI-1 reactor to LEU silicide fuel. More details are given. Due to the commercial pressures on the AEC at that time, the AEC decided not to convert SAFARI-1 to LEU silicide fuel. The AEC did, however, undertake to reevaluate the situation periodically. Five years have passed since the detailed study of 1994 and the changing conditions now warrant another detailed study taking into account the current situation. A brief historical background is given on the RERTR activities in South Africa since 1993 followed by a summary of the current situation

  3. Twenty years of RERTR in Russia: Past, present and future

    International Nuclear Information System (INIS)

    Arkhangelsky, N.

    2000-01-01

    The Russian RERTR Program started approximately 20 years ago. The USSR always supported principles and goals of the policy of nonproliferation and at the end of 70's the Soviet Government decided to create such national program. Twenty years is the sufficient time to estimate preliminary results of the realization of the program and its prospects on the future. After the first successes of the program, when the enrichment of uranium in fuel elements for foreign supplies was reduced from 80 % to 36 %, the realization of the program was suspended in connection with financial difficulties. But in the beginning of the 90's the Program has received a new pulse connected to the inclusion of Russian scientists and engineers in the international Program. Now basic directions in development of new kinds of fuel are development of works on fuel on a basis dioxide of uranium and development of fuel on a basis of U-Mo alloy. In the future to basic goals of the program the problem of the management of the spent nuclear fuel should be added. The management of HEU at the final stage of a fuel cycle becomes an important objective of the program, since the basic amount of HEU is concentrated in storage of SNF. (author)

  4. Status and progress of the RERTR program in the year 2003

    International Nuclear Information System (INIS)

    Travelli, Armando

    2003-01-01

    One of the most important events affecting the RERTR program during the past year was the decision by the U.S. Department of Energy to request the U.S. Congress to significantly increase RERTR program funding. This decision was prompted, at least in part, by the terrible events of September 11, 2001, and by a high-level U.S./Russian Joint Expert Group recommendation to immediately accelerate RERTR program activities in both countries, with the goal of converting all the world's research reactors to low-enriched fuel at the earliest possible time, and including both Soviet-designed and United States-designed research reactors. The U.S. Congress is expected to approve this request very soon, and the RERTR program has prepared itself well for the intense activities that the 'Accelerated RERTR Program' will require. Promising results have been obtained in the development of a fabrication process for monolithic LEU U-Mo fuel. Most existing and future research reactors could be converted to LEU with this fuel, which has a uranium density between 15.4 and 16.4 g/cm 3 and yielded promising irradiation results in 2002. The most promising method hinges on producing the monolithic meat by cold-rolling a thin ingot produced by casting. The aluminum clad and the meat are bonded by friction stir welding and the cladding surface is finished by a light cold roll. This method can be applied to the production of miniplates and appears to be extendable to the production of full-size plates, possibly with intermediate anneals. Other methods planned for investigation include high temperature bonding and hot isostatic pressing. The progress achieved within the Russian RERTR program, both for the traditional tube-type elements and for the new 'universal' LEU U-Mo pin-type elements, promises to enable soon the conversion of many Russian-designed research and test reactors. Irradiation testing of both fuel types with LEU U-Mo dispersion fuels has begun. Detailed studies are in progress to

  5. The RERTR [Reduced Enrichment Research and Test Reactor] Program: Progress and plans

    International Nuclear Information System (INIS)

    Travelli, A.

    1987-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1986, the activities, results, and new developments which occurred in 1987 are reviewed. Irradiation of the second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels, was completed and postirradiation examinations were performed on many of its miniplates. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 was completed at the end of March with excellent results and with 29 elements estimated to have reached at least 40% average burnup. Good progress was made in the area of LEU usage for the production of fission 99 Mo, and in the coordination of safety evaluations related to LEU conversions of US university reactors. Planned activities include testing and demonstrating advanced fuels intended to allow use of reduced enrichment uranium in very-high-performance reactors. Two candidate fuels are U 3 Si-Al with 19.75% enrichment and U 3 Si 2 -Al with 45% enrichment. Demonstration of these fuels will include irradiation of full-size elements and, possibly, a full-core demonstration. Achievement of the final program goals is still projected for 1990. This progress could not have been possible without the close international cooperation which has existed from the beginning, and which is essential to the ultimate success of the RERTR Program

  6. Abstracts and papers of the 1996 International RERTR Meeting

    International Nuclear Information System (INIS)

    1996-10-01

    The 1996 International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR) was sponsored by the Korea Atomic Energy Research Institute and the International Atomic Energy Agency and was held in Seoul, Korea, on 6-10 October 1996. The abstracts and available papers that were presented at this meeting are provided. The most important subject dealt with was progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program. The major events, findings, and activities of 1996 are reviewed with the results which the RERTR Program has achieved by the end of 1995 in collaboration of the ANL and its many international partners

  7. Abstracts and papers of the 1996 International RERTR Meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    The 1996 International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR) was sponsored by the Korea Atomic Energy Research Institute and the International Atomic Energy Agency and was held in Seoul, Korea, on 6-10 October 1996. The abstracts and available papers that were presented at this meeting are provided. The most important subject dealt with was progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program. The major events, findings, and activities of 1996 are reviewed with the results which the RERTR Program has achieved by the end of 1995 in collaboration of the ANL and its many international partners.

  8. The RERTR [Reduced Enrichment Research and Test Reactor] program: A progress report

    International Nuclear Information System (INIS)

    Travelli, A.

    1986-11-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1985, the activities, results, and new developments which occurred in 1986 are reviewed. The second miniplate series, concentrating on U 3 Si 2 -Al and U 3 Si-Al fuels, was expanded and its irradiation continued. Postirradiation examinations of several of these miniplates and of six previously irradiated U 3 Si 2 -Al full-size elements were completed with excellent results. The whole-core ORR demonstration with U 3 Si 2 -Al fuel at 4.8 g U/cm 3 is well under way and due for completion before the end of 1987. DOE removed an important barrier to conversions by announcing that the new LEU fuels will be accepted for reprocessing. New DOE prices for enrichment and reprocessing services were calculated to have minimal effect on HEU reactors, and to reduce by about 8 to 10% the total fuel cycle costs of LEU reactors. New program activities include preliminary feasibility studies of LEU use in DOE reactors, evaluation of the feasibility to use LEU targets for the production of fission-product 99 Mo, and responsibility for coordinating safety evaluations related to LEU conversions of US university reactors, as required by NRC. Achievement of the final program goals is projected for 1990. This progress could not have been achieved without close international cooperation, whose continuation and intensification are essential to the achievement of the ultimate goals of the RERTR Program

  9. Status and progress of the RERTR program in the year 2000

    International Nuclear Information System (INIS)

    Travelli, Armando

    2000-01-01

    This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was held by important accomplishments and events for the RERTR program: Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy; Postirradiation examinations of three batches of microplates have continued to reveal excellent irradiation behavior of U-Mo dispersion fuels in a variety of compositions and irradiating conditions. Irradiation of two new batches of miniplates of greater sizes is in progress in the ATR to investigate the swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g /cm 3 range; Qualification of the U-Mo dispersion fuels is proceeding on schedule. Test fuel elements with uranium density of 6 g/cm 3 are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo fuel with uranium density of 8-9 g/cm 3 is planned to be qualified by the end of 2005; Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1; Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission 99 Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets; Progress was made on irradiation testing of

  10. Status and progress of the RERTR program in the year 2000

    International Nuclear Information System (INIS)

    Travelli, A.

    2000-01-01

    This paper describes the progress achieved by the Reduced Enrichment for Research and Test Reactors (RERTR) Program in collaboration with its many international partners during the year 2000 and discusses the main activities planned for the year 2001. The past year was characterized by important accomplishments and events for the RERTR program. Four additional shipments containing 503 spent fuel assemblies from foreign research reactors were accepted by the U.S. Altogether, 3,740 spent fuel assemblies from foreign research reactors have been received by the U.S. under the acceptance policy. Postirradiation examinations of three batches of microplates have continued to reveal excellent irradiation behavior of U-MO dispersion fuels in a variety of compositions and irradiating conditions. h-radiation of two new batches of miniplates of greater sizes is in progress in the ATR to investigate me swelling behavior of these fuels under prototypic conditions. These materials hold the promise of achieving the program goal of developing LEU research reactor fuels with uranium densities in the 8-9 g /cm 3 range. Qualification of the U-MO dispersion fuels is proceeding on schedule. Test fuel elements with 6 gU/cm 3 are being fabricated by BWXT and are scheduled to begin undergoing irradiation in the HFR-Petten in the spring of 2001, with a goal of qualifying this fuel by the end of 2003. U-Mo with 8-9 gU/cm 3 is planned to be qualified by the end of 2005. Joint LEU conversion feasibility studies were completed for HFR-Petten and for SAFARI-1. Significant improvements were made in the design of LEU metal-foil annular targets that would allow efficient production of fission 99 Mo. Irradiations in the RAS-GAS reactor showed that these targets can formed from aluminum tubes, and that the yield and purity of their product from the acidic process were at least as good as those from the HEU Cintichem targets. Progress was made on irradiation testing of LEU UO 2 dispersion fuel and on

  11. Current status of the U.S. RERTR program

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-01-01

    The main objective of this paper is to provide the audience with some orientation and sense of perspective that may assist in viewing the other program presentations as part of an overall effort, rather than as separate and unrelated activities. In general, the purpose, goals, strategy and organization of the program have not changed from what they were when described in previous papers and conferences. From the technical point of view, on the other hand, this has probably been the year in which the largest number of significant new developments have taken place in the RERTR Program. Many of these new developments were favorable and corresponded to positive steps forward towards the program goals. Others corresponded to significant changes in program plans and projections. The major and minor headings of the areas whose main developments discussed are: Analysis and design (Methods and Codes, Generic Analysis Specific Support); Fuel development and element demonstration (UAI - Al, U 3 O 8 -Al, U 3 Si x Al y -Al, UZrH x ); Reprocessing; Whole core demonstrations. Some indication is given of the impact which the changes are likely to have on future program activities. This paper includes an outline of what appears today as the most probable program schedule for the next few years

  12. Investigation of the Cause of Low Blister Threshold Temperatures in the RERTR-12 and AFIP-4 Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell K Meyer

    2012-06-01

    Blister–threshold testing of fuel plates is a standard method through which the safety margin for operation of plate-type in research and test reactors is assessed. The blister-threshold temperature is indicative of the ability of fuel to operate at high temperatures for short periods of time (transient conditions) without failure. This method of testing was applied to the newly developed U-Mo monolithic fuel system. Blister annealing studies on the U-Mo monolithic fuel plates began in 2007, with the Reduced Enrichment for Research and Test Reactors (RERTR)-6 experiment, and they have continued as the U-Mo fuel system has evolved through the research and development process. Blister anneal threshold temperatures from early irradiation experiments (RERTR-6 through RERTR-10) ranged from 400 to 500°C. These temperatures were projected to be acceptable for NRC-licensed research reactors and the high-power Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) based on current safety-analysis reports (SARs). Initial blister testing results from the RERTR-12 experiment capsules X1 and X2 showed a decrease in the blister-threshold temperatures. Blister threshold temperatures from this experiment ranged from 300 to 400°C. Selected plates from the AFIP-4 experiment, which was fabricated using a process similar to that used to fabricate the RERTR-12 experiment, also underwent blister testing to determine whether results would be similar. The measured blister-threshold temperatures from the AFIP-4 plates fell within the same blister-threshold temperature range measured in the RERTR-12 plates. Investigation of the cause of this decrease in bister threshold temperature is being conducted under the guidance of Idaho National Laboratory PLN-4155, “Analysis of Low Blister Threshold Temperatures in the RERTR-12 and AFIP-4 Experiments,” and is driven by hypotheses. The main focus of the investigation is in the following areas: 1. Fabrication variables 2. Pre

  13. Proceedings of the international meeting on development, fabrication and application of reduced enrichment fuels for research and test reactors

    International Nuclear Information System (INIS)

    1983-08-01

    Separate abstracts were prepared for each of the papers presented in the following areas: (1) Reduced Enrichment Fuels for Research and Test Reactors (RERTR) Program Status; (2) Fuel Development; (3) Fuel Demonstrations; (4) General Topics; and (5) Specific Reactor Applications

  14. Comments on the future activities of the RERTR program - part II

    Energy Technology Data Exchange (ETDEWEB)

    Krull, W. [GKSS Research Centre Geesthacht GmbH (Germany)

    1997-08-01

    At the last RERTR meeting an historical overview was given and the status and consequences of enrichment reduction were discussed. At that time and somewhat more today many doubts are raised that enrichment reduction, as a tool for reducing the proliferation risk, is being done in the most efficient and convincing manner. The informations presented in this report were taken from IAEA, US-DOE, US-GAO publications and from proceedings of the RERTR meetings. The data presented should be compared only on a relative basis. It was not the intention for many reasons to present quantitative exact values as some figures used for developing the conclusions are being confidential. Others are available in the above mentioned publications and proceedings. But nevertheless conclusions drawn and recommendations developed are believed to be worth to be taken into account by research reactor operators, their funding organizations and administrators when they are faced with decisions about the future of their facilities. All of the international effort on reducing the enrichment on research and test reactors has their only justification from the INFCE (International Nuclear Fuel Cycle Evaluation) conclusions to reduce the proliferation risk worldwide to a large extent. Worldwide is important. IAEA and others have to convince operators and organizations upon the necessity in doing so. One of the best way convincing people is giving convincing examples. At present none of the nations having a military nuclear weapons program is giving such a convincing example. Even in cases where the qualified fuel and all other necessary tools for converting specific research reactors are existing only small or no progress can be seen. Only non-weapon states are being pressed today.

  15. Comments on the future activities of the RERTR program - part II

    International Nuclear Information System (INIS)

    Krull, W.

    1997-01-01

    At the last RERTR meeting an historical overview was given and the status and consequences of enrichment reduction were discussed. At that time and somewhat more today many doubts are raised that enrichment reduction, as a tool for reducing the proliferation risk, is being done in the most efficient and convincing manner. The informations presented in this report were taken from IAEA, US-DOE, US-GAO publications and from proceedings of the RERTR meetings. The data presented should be compared only on a relative basis. It was not the intention for many reasons to present quantitative exact values as some figures used for developing the conclusions are being confidential. Others are available in the above mentioned publications and proceedings. But nevertheless conclusions drawn and recommendations developed are believed to be worth to be taken into account by research reactor operators, their funding organizations and administrators when they are faced with decisions about the future of their facilities. All of the international effort on reducing the enrichment on research and test reactors has their only justification from the INFCE (International Nuclear Fuel Cycle Evaluation) conclusions to reduce the proliferation risk worldwide to a large extent. Worldwide is important. IAEA and others have to convince operators and organizations upon the necessity in doing so. One of the best way convincing people is giving convincing examples. At present none of the nations having a military nuclear weapons program is giving such a convincing example. Even in cases where the qualified fuel and all other necessary tools for converting specific research reactors are existing only small or no progress can be seen. Only non-weapon states are being pressed today. being pressed today

  16. IAEA activities in support of RERTR programme

    International Nuclear Information System (INIS)

    Akhtar, K.M.

    2004-01-01

    The International Atomic Energy Agency has supported the programme for Reduced Enrichment for Research and Test Reactors from the very initial stage. As part of its research reactor programme, the Agency has convened several technical meetings and seminars, issued many publications on the subject, and provided technical and financial assistance to many reactor operators in the developing countries. The worldwide current status of fuel enrichment for research reactors and a resume of Agency activities are presented in this paper. (author)

  17. RERTR end-game: A win-win framework. Phasing out remaining global HEU commerce by conditionally and temporarily renewing U.S. exports of HEU

    International Nuclear Information System (INIS)

    Kuperman, Alan J.; Leventhal, Paul L.

    1997-01-01

    The RERTR program stands on the brink of fulfilling its historic mission. However, a series of missteps and misunderstandings have recently raised the risk that defeat will be snatched from the jaws of victory. Perhaps the most serious threat to the RERTR regime is posed by France's pending import of 625 kilograms of bomb-grade, highly enriched uranium (HEU) from Russia, intended primarily to fuel its high-flux research reactor at the Institute Laue-Langevin in Grenoble, as well as its Orphee research reactor. As the first export of HEU from Russia to a facility outside the former Soviet bloc, this precedential transaction would establish Russia as a new global supplier of bomb-grade uranium, potentially setting the stage for a rise in international HEU commerce, rather than its phase-out as envisioned under the RERTR program. Apparently, France turned to Russia for supply of the fuel because the United States was perceived as unable or unwilling to continue supplying such fuel in the wake of the U.S. Energy Policy Act of 1992, which, pursuant to its so-called Schumer Amendment, places sharp restrictions on HEU exports. Unexplained delays in Russia's shipment of this material to France provide a fortuitous window of opportunity in which efforts can and should be made by France and the United States to resolve present differences in a manner beneficial to each, as well as in the interest of global security. This paper proposes an arrangement under which the United States would renew exports of HEU to France, in exchange for pledges from France enabling the export to comply with the principles and objectives of the RERTR program as embodied in U.S. law. In so doing, the arrangement would obviate the need for Russian HEU export, thereby avoiding its dangerous precedent. By enabling high quality scientific research to continue, while simultaneously helping to fulfill the RERTR program's original goal, such an arrangement would truly be a 'win-win' solution. (author)

  18. Progress of the RERTR [Reduced Enrichment Research and Test Reactor] Program in 1989

    International Nuclear Information System (INIS)

    Travelli, A.

    1989-01-01

    The progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program is described. After a brief summary of the results which the RERTR Program, in collaboration with its many international partners, had achieved by the end of 1988, the major events, findings, and activities of 1989 are reviewed. The scope of the RERTR Program activities was curtailed, in 1989, by an unexpected legislative restriction which limited the ability of the Arms Control and Disarmament Agency to adequately fund the program. Nevertheless, the thrust of the major planned program activities was maintained, and meaningful results were obtained in several areas of great significance for future work. 15 refs., 12 figs

  19. Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad Abir; Fahima Islam; Hyoung Koo Lee; Daniel Wachs

    2014-11-01

    The Convert branch of the National Nuclear Security Administration (NNSA) Global Threat Reduction Initiative (GTRI) focuses on the development of high uranium density fuels for research and test reactors for nonproliferation. This fuel is aimed to convert low density high enriched uranium (HEU) based fuel to high density low enriched uranium (LEU) based fuel for high performance research reactors (HPRR). There are five U.S. reactors that fall under the HPRR category, including: the Massachusetts Institute of Technology Reactor (MITR), the National Bureau of Standards Reactor (NBSR), the Missouri University Research Reactor (UMRR), the Advanced Test Reactor (ATR), and the High Flux Isotope Reactor (HFIR). U-Mo alloy fuel phase in the form of either monolithic or dispersion foil type fuels, such as ATR Full-size In center flux trap Position (AFIP) and Reduced Enrichment for Research and Test Reactor (RERTR), are being designed for this purpose. The fabrication process1 of RERTR is susceptible to introducing a variety of fuel defects. A dependable quality control method is required during fabrication of RERTR miniplates to maintain the allowable design tolerances, therefore evaluating and analytically verifying the fabricated miniplates for maintaining quality standards as well as safety. The purpose of this work is to analyze the thickness of the fabricated RERTR-12 miniplates using non-destructive technique to meet the fuel plate specification for RERTR fuel to be used in the ATR.

  20. Russian RERTR program as a part of Joint US DOE-RF MINATOM collaboration on elimination of the threat connected to the use of HEU in research reactors

    International Nuclear Information System (INIS)

    Arkhangelsky, N.

    2002-01-01

    The Russian RERTR Program started at the end of 70's, the final goal of the program is to eliminate supplies of HEU in fuel elements and assemblies for foreign research reactors that were designed according to Russian projects. Basic directions of the work include: completion of the development of the fuel elements and assemblies on a basis of uranium dioxide; development of the fuel on a basis of U-Mo alloy; and development of pin type fuel elements. Fuel assemblies of WWR-M2 type with LEU were developed and qualified for using in foreign research reactors that use such type of fuel assemblies. These assemblies are ready for the supplying several operating foreign research reactors. There are more than 20 sites in Eastern European countries, former Soviet republics and another countries that have big amount of Russian origin HEU in fresh and spent fuel. The problem of the shipment of SNF from sites of research reactors is also very important for domestic Russian research reactors. More than ten years from its beginning the Russian RERTR program developed practically independently from the international RERTR program and only at the begin of 90's the Russian specialists started to contact with foreign scientists and the exchange of the scientific information has become more intensive. In September 1994, representatives of Minatom and DOE signed a protocol of intent to reduce an enrichment of uranium in research reactors. The main aspects of collaboration involve: Several domestic Russian research reactors such as WWR-M, IR-8 and others were investigated from the point of view of possibility of reducing of enrichment; financial support of the program from US DOE which is insufficient. The important part of international collaboration is the import of Russian origin spent and fresh fuel of research reactors to Russia. In August 2002 an impressive result of the Russian-American collaboration with support of IAEA and with the help and assistance of Yugoslavian side was

  1. The RERTR demonstration experiments program at the Ford Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wehe, D K; King, J S [Department of Nuclear Engineering, University of Michigan (United States)

    1983-08-01

    The purpose of this paper is to highlight a major part of the experimental work which is being carried out at the Ford Nuclear Reactor (FNR) in conjunction with the RERTR program. A demonstration experiments program has been developed to: 1) characterize the FNR in sufficient detail to discern and quantify neutronic differences between the high and low enriched cores; 2) provide the theoretical group with measurements to benchmark their calculations. As with any experimental program associated with a reactor, stringent constraints limit the experiments which can be performed. Some experiments are performed routinely on the FNR (such as control rod calibrations), and much data is already available. Unfortunately, the accuracy we demand precludes using much of this earlier data. And in many cases, the requirement of precise (and copious) data has led to either developing new techniques (as in the case of rhodium mapping and neutron diffraction) or to further refinements on existing methods (as in the case of spectral unfolding). Nevertheless, we have tried to stay within the realm of recognized, well-established experimental methods in order to assuage any doubts about measured differences between HEU and LEU core parameters. This paper describes the principal results of the experiments performed so far.

  2. RERTR progress in Mo-99 production from LEU

    Energy Technology Data Exchange (ETDEWEB)

    Vandegrift, G.F.; Conner, C.; Aase, S.; Bakel, A.; Bowers, D.; Freiberg, E.; Gelis, A.; Quigley, K.J.; Snelgrove, J.L. [Argonne National Laboratory 9700 S. Cass Avenue, Argonne, IL (United States)

    2002-07-01

    The ANL RERTR program is performing R and D supporting conversion of {sup 99}Mo production from HEU to LEU targets. Irradiation and processing of LEU targets were demonstrated at the Argentine Ezeiza Atomic Center. Target irradiation and disassembly were flawless, but the processing is not fully developed. In addition to preparing for, assisting in, and analyzing results of the demonstration, we performed other R and D related to LEU conversion: (1) designing a prototype production dissolver for digesting irradiated LEU foils in alkaline solutions and developing means to simplify digestion, (2) modifying ion-exchange columns used in the CNEA recovery and purification of {sup 99}Mo to deal with the lower volumes generated from LEU-foil digestion, (3) measuring the performance of new inorganic sorbents that outperform alumina for recovering Mo(VI) from nitric acid solutions containing high concentrations of uranium nitrate, and (4) developing means to facilitate the concentration and calcination of waste nitric-acid/LEU-nitrate solutions from {sup 99} Mo production. (author)

  3. Modification and verification of the program COBRA-RERTR for the application in the thermohydraulic analysis of research reactors

    International Nuclear Information System (INIS)

    Hainoun, A.; Ghazi, N.

    2005-01-01

    In the frame work of testing, evaluation and application of computer codes in the design and safety analysis of research reactors, the thermal hydraulic code COBRA-RERTR (Reduced Enriched Research and Test Reactor) has been tested and partially validated. COBRA-RERTR has been selected due to the available options, which are suitable for the analysis of research reactors which are operated at low temperatures, and which may use plate-type fuel elements and heavy water as the coolant. In addition to that, the code enables the consideration of cross-flow that is important in case of parallel and open coolant channel. The test of the code shows an overestimation of the wall temperature with an addition to some fluctuation from node to node. This results from the solution scheme that uses an explicit, non-iterative solution for heat conduction and heat transfer to the coolant. The code evaluation regarding the basic thermal hydraulic phenomena indicates the necessity to modify and extent the physical models deals with the estimation of slip ratio and simulation of void content in the sub-cooled boiling. The code has been validated by recalculation of special experiments on axial void distribution and thermal hydraulic instability in the subcooled boiling regime. The validation indicates significant improvement of the code in prediction the axial void distribution in subcooled boiling. The discrepancy between calculation and experiments was about 20% after the modification comparing to 100% in the original models. On the other hand the validation shows the capability of the modified code to stimulate thermal hydraulic flow instability characterized by the critical inlet flow velocity at which the flow just become unstable. This point is identical to the minimum in the integral pressure drop curve. The code results show, from the view point of reactor safety, conservative estimation since the predicted values of critical inlet velocity are higher than the experimental

  4. Modification and verification of the program COBRA-RERTR for the application in the thermohydraulic analysis of research reactors

    International Nuclear Information System (INIS)

    Hainoun, A.; Ghazi, N.

    2004-02-01

    In the frame work of testing, evaluation and application of computer codes in the design and safety analysis of research reactors, the thermal hydraulic code COBRA-RERTR (Reduced Enriched Research and Test Reactor) has been tested and partially validated. COBRA-RERTR has been selected due to the available options, which are suitable for the analysis of research reactors which are operated at low temperatures, and which may use plate-type fuel elements and heavy water as the coolant. In addition to that, the code enables the consideration of cross-flow that is important in case of parallel and open coolant channel. The test of the code shows an overestimation of the wall temperature with an addition to some fluctuation from node to node. This results from the solution scheme that uses an explicit, non-iterative solution for heat conduction and heat transfer to the coolant. The code evaluation regarding the basic thermal hydraulic phenomena indicates the necessity to modify and extent the physical models deals with the estimation of slip ratio and simulation of void content in the sub-cooled boiling. The code has been validated by recalculation of special experiments on axial void distribution and thermal hydraulic instability in the subcooled boiling regime. The validation indicates significant improvement of the code in prediction the axial void distribution in subcooled boiling. The discrepancy between calculation and experiments was about 20% after the modification comparing to 100% in the original models. On the other hand the validation shows the capability of the modified code to stimulate thermal hydraulic flow instability characterized by the critical inlet flow velocity at which the flow just become unstable. This point is identical to the minimum in the integral pressure drop curve. The code results show, from the view point of reactor safety, conservative estimation since the predicted values of critical inlet velocity are higher than the experimental

  5. Cost of the external MTR-fuel cycle. (Uranium , reprocessing and related services)

    International Nuclear Information System (INIS)

    Mueller, H.; Gruber, G.

    1991-01-01

    This paper points out how the RERTR program has affected NUKEM's fuel supplies for MTRs and how the prices in the External MTR Fuel Cycle have developed during this period. In addition other potential fuel sources and services on the External MTR Fuel Cycle are given. (orig.)

  6. HEU core conversion of Russian production reactors: a major threat to the international RERTR regime

    International Nuclear Information System (INIS)

    Kuperman, Alan J.; Leventhal, Paul L.

    1998-01-01

    This paper calls the attention for the major threat to the International Reduced Enrichment for Research and Test Reactors (RERTR) program, represented by the HEU core conversion of russian production reactors. This program aims to reduce and eventually eliminate international civilian commerce in nuclear weapons-usable, highly enriched uranium , and thereby significantly lower risks of the material being stolen or diverted by terrorist or states for producing nuclear weapons

  7. Fuel Cycle Concept with Advanced METMET and Composite Fuel in LWRs

    International Nuclear Information System (INIS)

    Savchenko, A.; Skupov, M.; Vatulin, A.; Glushenkov, A.; Kulakov, G.; Lipkina, K.

    2014-01-01

    The basic factor that limits the serviceability of fuel elements developing in the framework of RERTR Program (transition from HEU to LEU fuel of research reactors) is interaction between U10Mo fuel and aluminium matrix . Interaction results in extra swelling of fuels, disappearance of a heat conducting matrix, a temperature rise in the fuel centre, penetration porosity, etc. Several methods exist to prevent fuel-matrix interaction. In terms of simplifying fuel element fabrication technology and reducing interaction, doping of fuel is the most optimal version

  8. Current status and future trends of the Russian RERTR program

    International Nuclear Information System (INIS)

    Bezzubtsev, V.; Arkhangelsky, N.; Aden, V.; Chernyshov, V.; Vatulin, A.

    2003-01-01

    The Russian program started at the end of 70's. The final goal of the program was to minimize and in perspective to eliminate using of HEU in fuel elements and assemblies supplied to foreign research reactors with the minimum of penalties for the experimenters. Later officials, scientists and engineers began political and scientific discussion about reducing enrichment domestic reactors as well as return of fresh and spent HEU fuel. Important issues of the Russian program is concerned with cooperation with USA, especially related to new types of LEU fuel elements development and fabrication to be used in Russian designed foreign research reactors. According to present results this goal will be achieved by 2006

  9. Results of fuel elements fabrication on the basis of increased concentration dioxide fuel for research reactors

    International Nuclear Information System (INIS)

    Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

    1996-01-01

    According to the Russian Reduced Enrichment for Research and Test Reactors (RERTR) program, that were constructed under the Russian projects, at the Novosibirsk Chemical Concentrates Plant the pilot series of different configuration (WR-M2, MR, IRT-4M) fuel elements, based on increased concentration uranium dioxide fuel, have been fabricated for reactor tests. Comprehensive fabricated fuel elements quality estimation has been carried out. (author)

  10. LEU fuel fabrication in Argentina

    International Nuclear Information System (INIS)

    Giorsetti, D.R.; Gomez, J.O.; Marajofsky, A.; Kohut, C.

    1985-01-01

    As an Institution, aiming to meet with its own needs, CNEA has been intensively developing reduced enriched fuel to use in its own research and test reactors. Development of the fabrication technology as well as the design, installation and operation of the manufacturing plant, have been carried out with its own funds. Irradiation and post-irradiation of test miniplates have been taking place within the framework of the RERTR program. During the last years, CNEA has developed three LEU fuel types. In the previous RERTR meetings, we presented the technological results obtained with these fuel types. This paper focuses on CNEA LEU fuel element manufacturing status and the trained personnel we can offer in design and manufacture fuel capability. CNEA has its own fuel manufacturing technology; the necessary facilities to start the fuel fabrication; qualified technicians and professionals for: fuel design and behaviour analysis; fuel manufacturing and QA; international recognition of its fuel development and manufacturing capability through its ORR miniplate irradiation; its own natural uranium and the future possibility to enrich up to 20% U 235 ; the probability to offer a competitive fuel manufacturing cost in the international market; the disposition to cooperate with all countries that wish to take part and aim to reach an self-sufficiency in their own fuel supply needs

  11. Fuel powder production from ductile uranium alloys

    International Nuclear Information System (INIS)

    Clark, C.R.; Meyer, M.K.

    1998-01-01

    Metallic uranium alloys are candidate materials for use as the fuel phase in very-high-density LEU dispersion fuels. These ductile alloys cannot be converted to powder form by the processes routinely used for oxides or intermetallics. Three methods of powder production from uranium alloys have been investigated within the US-RERTR program. These processes are grinding, cryogenic milling, and hydride-dehydride. In addition, a gas atomization process was investigated using gold as a surrogate for uranium. (author)

  12. Shipment of VINCA Institute's HEU fresh fuel to Russia

    International Nuclear Information System (INIS)

    Pesic, Milan; Sotic, Obrad

    2002-01-01

    This paper shows, for the first time, the basic data related to the recent shipment of the fresh HEU fuel elements from Yugoslavia back to Russia for uranium down blending. In this way, Yugoslavia gives its contribution to the RERTR program and to the world's joint efforts to prevent possible terrorist action against nuclear material potentially usable for production of nuclear weapons. (author)

  13. Qualification of uranium-molybdenum alloy fuel - conclusions of an international workshop

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Languille, A.

    2000-01-01

    Thirty-one participants representing 21 reactors, fuel developers, fuel fabricators, and fuel reprocessors in 11 countries discussed the requirements for qualification of U-Mo alloy fuel at a workshop held at Argonne National Laboratory on January 17-18, 2000. Consensus was reached that the qualification plans of the U.S. RERTR program and the French U-Mo fuel development program are valid. The items to be addressed during qualification are summarized in the paper. (author)

  14. Thermomechanical DART code improvements for LEU VHD dispersion and monolithic fuel element analysis

    International Nuclear Information System (INIS)

    Taboada, H.; Saliba, R.; Moscarda, M.V.; Rest, J.

    2005-01-01

    A collaboration agreement between ANL/US DOE and CNEA Argentina in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual FASTDART version and also a DART THERMAL version were presented during RERTR 2000, 2002 and RERTR 2003 Meetings. During this past year the following activities were completed: Optimization of DART TM code Al diffusion parameters by testing predictions against reliable data from RERTR experiments. Improvements on the 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic fuel. Concerning the first point, by means of an optimization of parameters of the Al diffusion through the interaction product theoretical expression, a reasonable agreement between DART temperature calculations with reliable RERTR PIE data was reached. The 3-D thermomechanical code complex is based upon a finite element thermal-elastic code named TERMELAS, and irradiation behavior provided by the DART code. An adequate and progressive process of coupling calculations of both codes at each time step is currently developed. Compatible thermal calculation between both codes was reached. This is the first stage to benchmark and validate against RERTR PIE data the coupling process. (author)

  15. Irradiation behavior of miniature experimental uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Hofman, G.L.; Neimark, L.A.; Mattas, R.F.

    1983-01-01

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk from, on the order of 7 x 10 20 cm -3 , far short of the approximately 20 x 10 20 cm -3 goal established for the RERTR program. The purpose of the irradiation experiments on silicide fuels on the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix

  16. Status of research reactor fuel development in KAERI

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Ryu, Woo-Seok; Park, Jong-Man; Lee, Don-Bae; Kim, Ki-Hwan; Kuk, Il-Hyun

    1996-01-01

    The development of uranium silicide dispersion fuel fabrication technology has been carried out in KAERI. LEU fuel bundle was prepared for irradiation test. In order to compare the performance of atomized and comminuted U 3 Si dispersed fuels, the bundle of two kinds of fuel elements were prepared. Irradiation test will be performed in the OR-hole of HANARO in the near future. U 3 Si 2 atomization technology has been improved by using ceramic crucible and nozzle. Irradiation test for atomized U 3 Si 2 plate type fuel will be carried out in cooperation with ANL by using HANARO in connection with RERTR advanced fuel development. (author)

  17. Parametric study of fission-induced U-Mo fuel creep and structural analysis of fuel plates in view of implications for microstructure evolution

    International Nuclear Information System (INIS)

    Kim, Y.S.; Hofman, G.L.; Choo, Y.S.; Robinson, A.B.

    2010-01-01

    U-Mo fuel deformation during irradiation in U-Mo/Al dispersion plates is investigated by using the irradiation data from the RERTR-3 through -9 tests. The observation of fuel particle sintering during irradiation is also presented and its influence for fuel performance is discussed. Structural analysis was also performed to examine the relationship between the stress distribution in the plate and the location of matrix-pore formation in the plate. (author)

  18. Irradiation tests of U3Si2-Al fuels up to very high fission densities

    International Nuclear Information System (INIS)

    Nuding, M.; Boening, K.

    2001-01-01

    The new research reactor of the Munich Technical University (TUM), the FRM-II, will have U 3 Si 2 -Al as the fuel. This fuel is considered qualified and optimally usable in the light of findings obtained in the RERTR program (Reduced Enrichment for Research and Test Reactors). The RERTR program was conducted to develop new fuel for the use of low enriched uranium (LEU) in research reactors. As the unique properties of the FRM-II in research and application are based also on achieving a very compact reactor core with highly enriched uranium (HEU), additional irradiation tests were performed on the basis of the RERTR program. They were run in close cooperation with the French Commissariat a l'Energie Atomique (CEA) in its SILOE and OSIRIS facilities, among others. After extensive evaluation, also of other studies, these tests confirm the RERTR findings about fuel swelling behavior and, consequently, the suitability of U 3 Si 2 -Al (HEU) for use in the compact core of the FRM-II. (orig.) [de

  19. DART-TM: A thermomechanical version of DART for LEU VHD dispersed and monolithic fuel analysis

    International Nuclear Information System (INIS)

    Saliba, Roberto; Taboada, Horacio; Moscarda, Ma.Virginia; Rest, Jeff

    2003-01-01

    A collaboration agreement between ANL/USDOE and CNEA Argentina, in the area of Low Enriched Uranium Advanced Fuels has been in place since October 16, 1997 under the 'Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy'. An annex concerning DART code optimization has been operative since February 8, 1999. Previously, as a part of this annex a visual thermal FASTDART version was developed that includes mechanistic models for the calculation of the fission-gas-bubble and fuel particle size distribution, reaction layer thickness, and meat thermal conductivity. FASTDART was presented at the last RERTR Meeting that included validation against RERTR 3 irradiation data. The thermal FASTDART version was assessed as an adequate tool for modeling the behavior of LEU U-Mo dispersed fuels under irradiation against PIE RERTR irradiation data. During this past year the development of a 3-D thermo-mechanical version of the code for modeling the irradiation behavior of LEU U-Mo monolithic and dispersion fuel was initiated. Some preliminary results of this work will be shown during RERTR-2003 meeting. (author)

  20. CERCA's 25 years experience in U3Si2 fuel manufacturing

    International Nuclear Information System (INIS)

    Durand, JP.; Duban, B.; Lavastre, Y.; Perthuis, S. de

    2003-01-01

    This paper documents the experience gained at CERCA in manufacturing, testing, and inspecting U 3 Si 2 fuel elements for various Material Test Reactors (MTR) since the beginning of the RERTR Program in 1978, up to now. It emphasises how the company controls the product to insure compliance with the fuel-related safety parameters. Finally, those statements are considered in the UMo fuel production perspective. (author)

  1. Postirradiation examination of high-density uranium alloy dispersion fuels

    International Nuclear Information System (INIS)

    Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Strain, R.V.

    1998-01-01

    Two irradiation test vehicles, designated RERTR-2, were inserted into the Advanced Test reactor in Idaho in August 1997. These tests were designed to obtain irradiation performance information on a variety of potential new, high-density uranium alloy dispersion fuels, including U-10Mo, U-8Mo, U-6Mo, U-4Mo, U-9Nb-3Zr, U-6Nb-4Zr, U-5Nb-3Zr, U-6Mo-1Pt, U-6Mo-0.6Ru and U-10Mo-0.05Sn: the intermetallic compounds U 2 Mo and U-10Mo-0.-5Sn; the intermetallic compounds U 2 Mo and U 3 Si 2 were also included in the fuel test matrix. These fuels are included in the experiments as microplates (76 mm x 22 mm x 1.3mm outer dimensions) with a nominal fuel volume loading of 25% and irradiated at relatively low temperature (∼100 deg C). RERTR-1 and RERTR-2 were discharged from the reactor in November 1997 and July 1998, respectively at calculated peak fuel burnups of 45 and 71 at %-U 235 Both experiments are currently under examination at the Alpha Gamma Hot Cell Facility at Argonne National Laboratory in Chicago. This paper presents the postirradiation examination results available to date from these experiments. (author)

  2. Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K.; Ross Finlay, M.

    2012-06-01

    The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

  3. Effects of irradiation on the microstructure of U-7Mo dispersion fuel with Al-2Si matrix

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, Dennis D., E-mail: Dennis.Keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Jue, Jan-Fong; Robinson, Adam B.; Medvedev, Pavel; Gan, Jian; Miller, Brandon D.; Wachs, Daniel M.; Moore, Glenn A.; Clark, Curtis R.; Meyer, Mitchell K. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Ross Finlay, M. [Australian Nuclear Science and Technology Organization, PMB 1, Menai, NSW 2234 (Australia)

    2012-06-15

    The Reduced Enrichment for Research and Test Reactor (RERTR) program is developing low-enriched uranium U-Mo dispersion fuels for application in research and test reactors around the world. As part of this development, fuel plates have been irradiated in the Advanced Test Reactor and then characterized using optical metallography (OM) and scanning electron microscopy (SEM) to determine the as-irradiated microstructure. To demonstrate the irradiation performance of U-7Mo dispersion fuel plates with 2 wt.% Si added to the matrix, fuel plates were tested to moderate burnups at intermediate fission rates as part of the RERTR-6 experiment. Further testing was performed to higher fission rates as part of the RERTR-7A experiment, and very aggressive testing (high temperature, high fission density, and high fission rate) was performed in the RERTR-9A, RERTR-9B, and AFIP-1 experiments. As-irradiated microstructures were compared to those observed after fabrication to determine the effects of irradiation on the microstructure. Based on comparison of the microstructural characterization results for each irradiated sample, some general conclusions can be drawn about how the microstructure evolves during irradiation: there is growth during irradiation of the fuel/matrix interaction (FMI) layer created during fabrication; Si diffuses from the FMI layer to deeper depths in the U-7Mo particles as the irradiation conditions are made more aggressive; lowering of the Si content in the FMI layer results in an increase in the size of the fission gas bubbles; as the FMI layer grows during irradiation, more Si diffuses from the matrix to the FMI layer/matrix interface; and interlinking of fission gas bubbles in the fuel plate microstructure that may indicate breakaway swelling is not observed.

  4. Progress in irradiation performance of experimental uranium - Molybdenum dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.

    2002-01-01

    High-density dispersion fuel experiment, RERTR-4, was removed from the Advanced Test Reactor (ATR) after reaching a peak U-235 burnup of ∼80% and is presently undergoing postirradiation examination at the ANL alpha-gamma hot cells. This test consists of 32 mini fuel plates of which 27 were fabricated with nominally 6 and 8 g cm -3 atomized and machined uranium alloy powders containing 7 wt% and 10 wt% molybdenum. In addition, two miniplates containing solid U-10 wt% Mo foils and three containing 6 g cm -3 U 3 Si 2 are part of the test. The results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from previous tests performed to lower burnup will be presented. (author)

  5. CERCA'S experience in UMO fuel manufacturing

    International Nuclear Information System (INIS)

    Jarousse, Ch.; Lavastre, Y.; Grasse, M.

    2003-01-01

    Considered as a suitable solution for non-proliferation and reprocessing purposes, UMo fuel has been chosen and studied by the RERTR program since 1996. Involved in the RERTR fuel developments since 1978, with more than 20 years of U 3 SI 2 fuel production, and closely linked to the French Commissariat a l'Energie Atomique, CERCA was able to define properly, from the beginning, the right R and D actions plan for UMo fuel development. CERCA has already demonstrated during the last 4 years its ability to manufacture plates and fuel elements with high density UMo fuel. UMo full size plates produced for 4 irradiation experiments in 3 European reactors afforded us a unique experience. In addition, as a main part of our R and D effort, we have always studied in depth a key part of the CERCA process outline which is the plate rolling stage. After some preliminary investigation in order to define the phenomenological model describing the behavior of the fuel core when rolling, we have developed a rolling digital simulator. (author)

  6. MTR fuel element supply by CERCA through CECCN after the production transfer from NUKEM

    International Nuclear Information System (INIS)

    Hassel, H.W.

    1991-01-01

    The transfer of fuel element supply contracts, the corresponding Al-materials, structure parts, documents, uranium metal, customers related know-how, tools and equipment from NUKEM to CERCA has been completed, thus now giving a high flexibility for CERCA's workshop to fabricate and inspect large quantities of several types of fuel elements simultaneously. Based on this fact, on strategic planning for the next couple of years and on the fact that after 10 years of RERTR program the necessary high density fuel has been successfully developed and implemented, 'business as usual' in the field of fabrication has well become possible. The RERTR community should now use the great chance to concentrate all its efforts on problems which still strongly influence the fabrication and the use of MTR fuel elements: supply of enriched uranium,reprocessing capabilities and politics, transports of nuclear materials. (author)

  7. Observation on the irradiation behavior of U-Mo alloy dispersion fuel

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Meyer, Mitchell K.; Park, Jong-Man

    2000-01-01

    Initial results from the postirradiation examination of high-density dispersion fuel test RERTR-3 are discussed. The U-Mo alloy fuels in this test were irradiated to 40% U-235 burnup at temperature ranging from 140 0 C to 240 0 C. Temperature has a significant effect on overall swelling of the test plates. The magnitude of the swelling appears acceptable and no unstable irradiation behavior is evident. (author)

  8. Reduction of fuel enrichment for research reactors built-up in accordance with Russian (Soviet) projects

    International Nuclear Information System (INIS)

    Aleksandrov, A.B.; Enin, A.A.; Tkachyov, A.A.

    2001-01-01

    In accordance with the Russian program of reduced enrichment for research and test reactors (RERTR) built-up in accordance with Russian (Soviet) projects, AO 'NCCP' performs works on FA fabrication with reduced enrichment fuel. The main trends and results of performed works on research reactors FEs and FAs based on UO 2 and U-9%Mo fuel with U 235 19.7% enrichment are described. (author)

  9. Fabrication of fuel elements on the basis of increased concentration fuel composition

    International Nuclear Information System (INIS)

    Alexandrov, A.B.; Afanasiev, V.L.; Enin, A.A.; Suprun, V.B.

    2004-01-01

    As a part of Russian Program RERTR Reduced Enrichment for Research and Test Reactors), at NCCP, Inc. jointly with the State Scientific Centre VNIINM the mastering in industrial environment of design and fabrication process of fuel elements (FE) with increased concentration fuel compositions is performed. Fuel elements with fuel composition on the basis of dioxide uranium with nearly 4 g/cm 3 fuel concentration have been produced thus confirming the principal possibility of fuel enrichment reduction down to 20% for research reactors which were built up according to the projects of the former USSR, by increasing the oxide fuel concentration in fuel assemblies (FAs). The form and geometrical dimensions of FEs and FAs shall remain unchanged, only uranium mass in FA shall be increased. (author)

  10. Irradiation behavior of experimental miniature uranium silicide fuel plates

    International Nuclear Information System (INIS)

    Hofman, Gerard L.; Neimark, L.A.; Mattas, R.F.

    1983-01-01

    Uranium silicides, because of their relatively high uranium density, were selected as candidate dispersion fuels for the higher fuel densities required in the Reduced Enrichment Research and Test Reactor (RERTR) Program. Irradiation experience with this type of fuel, however, was limited to relatively modest fission densities in the bulk form, on the order of 7 x 10 20 cm -3 , far short of he approximately 20 x 10 20 cm -3 goal established for the RERTR Program. The purpose of the irradiation experiments on silicide fuels in the ORR, therefore, was to investigate the intrinsic irradiation behavior of uranium silicide as a dispersion fuel. Of particular interest was the interaction between the silicide particles and the aluminum matrix, the swelling behavior of the silicide particles, and the maximum volume fraction of silicide particles that could be contained in the aluminum matrix. The first group of experimental 'mini' fuel plates have recently reached the program's goal burnup and are in various stages of examination. Although the results to date indicate some limitations, it appears that within the range of parameters examined thus far the uranium silicide dispersion holds promise for satisfying most of the needs of the RERTR Program. The twelve experimental silicide dispersion fuel plates that were irradiated to approximately their goal exposure show the 30-vol % U 3 Si-Al plates to be in a stage of relatively rapid fission-gas-driven swelling at a fission density of 2 x 10 20 cm -3 . This fuel swelling will likely result in unacceptably large plate-thickness increases. The U 3 Si plates appear to be superior in this respect; however, they, too, are starting to move into the rapid fuel-swelling stage. Analysis of the currently available post irradiation data indicates that a 40-vol % dispersed fuel may offer an acceptable margin to the onset of unstable thickness changes at exposures of 2 x 10 21 fission/cm 3 . The interdiffusion between fuel and matrix

  11. Nitride Coating Effect on Oxidation Behavior of Centrifugally Atomized U-Mo Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Jin; Cho, Woo Hyoung; Park, Jong Man; Lee, Yoon Sang; Yang, Jae Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    Uranium metal and uranium compounds are being used as nuclear fuel materials and generally known as pyrophoric materials. Nowadays the importance of nuclear fuel about safety is being emphasized due to the vigorous exchanges and co-operations among the international community. According to the reduced enrichment for research and test reactors (RERTR) program, the international research reactor community has decided to use low-enriched uranium instead of high-enriched uranium. As a part of the RERTR program, KAERI has developed centrifugally atomized U-Mo alloys as a promising candidate of research reactor fuel. Kang et al. studied the oxidation behavior of centrifugally atomized U-10wt% Mo alloy and it showed better oxidation resistance than uranium. In this study, the oxidation behavior of nitride coated U-7wt% Mo alloy is investigated to enhance the safety against pyrophoricity

  12. Advanced Research Reactor Fuel Development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, C. K.; Park, H. D.; Kim, K. H. (and others)

    2006-04-15

    RERTR program for non-proliferation has propelled to develop high-density U-Mo dispersion fuels, reprocessable and available as nuclear fuel for high performance research reactors in the world. As the centrifugal atomization technology, invented in KAERI, is optimum to fabricate high-density U-Mo fuel powders, it has a great possibility to be applied in commercialization if the atomized fuel shows an acceptable in-reactor performance in irradiation test for qualification. In addition, if rod-type U-Mo dispersion fuel is developed for qualification, it is a great possibility to export the HANARO technology and the U-Mo dispersion fuel to the research reactors supplied in foreign countries in future. In this project, reprocessable rod-type U-Mo test fuel was fabricated, and irradiated in HANARO. New U-Mo fuel to suppress the interaction between U-Mo and Al matrix was designed and evaluated for in-reactor irradiation test. The fabrication process of new U-Mo fuel developed, and the irradiation test fuel was fabricated. In-reactor irradiation data for practical use of U-Mo fuel was collected and evaluated. Application plan of atomized U-Mo powder to the commercialization of U-Mo fuel was investigated.

  13. IAEA activities related to research reactor fuel conversion and spent fuel return programs

    International Nuclear Information System (INIS)

    Goldman, Ira N.; Adelfang, Pablo; Ritchie, Iain G.

    2005-01-01

    The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programs have supported research reactor fuel conversion from HEU to low enriched uranium (LEU), and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. (author)

  14. IAEA activities related to research reactor fuel conversion and spent fuel return programmes

    International Nuclear Information System (INIS)

    Ritchie, I.G.; Adelfang, P.; Goldman, I.N.

    2004-01-01

    Full text: The IAEA has been involved for more than twenty years in supporting international nuclear non-proliferation efforts associated with reducing the amount of highly enriched uranium (HEU) in international commerce. IAEA projects and activities have directly supported the Reduced Enrichment for Research and Test Reactors (RERTR) programme, as well as directly associated efforts to return research reactor fuel to the country of origin where it was originally enriched. IAEA efforts have included the development and maintenance of several data bases with information related to research reactors and research reactor spent fuel inventories that have been essential in planning and managing both RERTR and spent fuel return programmes. Other IAEA regular budget programmes have supported research reactor fuel conversion from HEU to low enriched uranium, and in addressing issues common to many member states with spent fuel management problems and concerns. The paper briefly describes IAEA involvement since the early 1980's in these areas, including regular budget and Technical Co-operation programme activities, and focuses on efforts in the past five years to continue to support and accelerate U.S. and Russian research reactor spent fuel return programmes. It is hoped that an announcement of the extension of the U.S. Acceptance Programme, which is expected in the very near future, will facilitate the life extensions of many productive TRIGA reactors around the world. (author)

  15. COMPARATIVE ANALYSIS OF STRUCTURAL CHANGES IN U-MO DISPERSED FUEL OF FULL-SIZE FUEL ELEMENTS AND MINI-RODS IRRADIATED IN THE MIR REACTOR

    OpenAIRE

    ALEKSEY. L. IZHUTOV; VALERIY. V. IAKOVLEV; ANDREY. E. NOVOSELOV; VLADIMIR. A. STARKOV; ALEKSEY. A. SHELDYAKOV; VALERIY. YU. SHISHIN; VLADIMIR. M. KOSENKOV; ALEKSANDR. V. VATULIN; IRINA. V. DOBRIKOVA; VLADIMIR. B. SUPRUN; GENNADIY. V. KULAKOV

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%235U; th...

  16. Development of quality assurance methods for low enriched fuel assemblies

    International Nuclear Information System (INIS)

    Woolstenhulme, N.E.; Moore, G.A.; Perez, D.M.; Wachs, D.M.

    2010-01-01

    As the Reduced Enrichment for Research and Test Reactors (RERTR) fuel development program has furthered the technology of low enriched uranium fuels, much effort has been expended to specify requirements, perform appropriate inspections, and to qualify experimental fuel plates and assemblies for irradiation. A great deal of consideration has been given to generate examinations and criteria that are both applicable to the unique fuel types being developed and consistent with industry practices for inspecting plate-type reactor fuel. Recent developments in quality assurance (QA) methodologies have given a heightened confidence in satisfactory fuel plate performance. At the same time, recommendations are given to further develop a system suitable for the testing and acceptance of production fuel elements containing low enriched uranium fuels. (author)

  17. Final disposition of MTR fuel

    International Nuclear Information System (INIS)

    Jonnson, Erik B.

    1996-01-01

    The final disposition of power reactor fuel has been investigated for a long time and some promising solutions to the problem have been shown. The research reactor fuels are normally not compatible with the zirkonium clad power reactor fuel and can thus not rely on the same disposal methods. The MTR fuels are typically Al-clad UAl x or U 3 Si 2 , HEU resp. LEU with essentially higher remaining enrichment than the corresponding power reactor fuel after full utilization of the uranium. The problems arising when evaluating the conditions at the final repository are the high corrosion rate of aluminum and uranium metal and the risk for secondary criticality due to the high content on fissionable material in the fully burnt MTR fuel. The newly adopted US policy to take back Foreign Research Reactor Spent Fuel of US origin for a period of ten years have given the research reactor society a reasonable time to evaluate different possibilities to solve the back end of the fuel cycle. The problem is, however, complicated and requires a solid engagement from the research reactor community. The task would be a suitable continuation of the RERTR program as it involves both the development of new fuel types and collecting data for the safe long-term disposal of the spent MTR fuel. (author)

  18. C.E.R.C.A. contribution to the RERTR program. Status of development, November 1982

    International Nuclear Information System (INIS)

    Savornin, B.; Fanjas, Y.R.

    1983-01-01

    For more than 20 years, CERCA has been manufacturing MTR fuel plates and fuel elements. 200.000 fuel plates have thus been produced and irradiated in reactors of 15 countries through the world, including the U.S.A. and Japan. On the way to the use of LEU in research reactors, we have adapted our fuel plate fabrication technology and developed new fuels in order to achieve high uranium densities in the meat. These technological developments have been performed on our own funds. Aluminide (U Al x ), Oxide (U 3 O 8 ) and Silicide (U 3 Si, U 3 Si 2 , U 3 Si Al) fuels have been successfully developed. Upon completion of the aluminide and oxide developments, we have intensified our effort on silicide fuels. After a brief review of the technological results obtained with U Al x and U 3 O 8 materials, silicide fuels results are examined in more detail. The irradiation status of these various fuels is discussed. It has been shown that CERCA is in position to provide on industrial scale UAl x , U 3 O 8 and U 3 Si 2 fuels with uranium densities up to 5 g/cm 3 . New equipment has been installed in MTR workshop in ROMANS for this purpose. For densities between and 74 g U/cm 3 , pilot technology is available. Development is carried out towards industrialization of U 3 Si fuel production. The commercial availability of these fuels depends on their qualification by post irradiation examination. It can be seen that aluminide fuel will be commercially available in 1983, oxide fuel in 1984. As far as silicide fuels are concerned, their availability will depend on the number of tests to be performed and the burn-up levels required by the Safety Commissions to consider the fuel qualified. These requirements may be different from one reactor to the other. This is why the corresponding dates stretch from 1985 to 1988. The goal of technological development has been reached. Densities above 7 g U/cm 3 have been achieved. Assuming silicide fuels reprocessing does not present major

  19. Progress in development of low-enriched U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Hofman, G.L.; Snelgrove, J.L.; Hayes, S.L.; Meyer, M.K.

    2002-01-01

    Results from post irradiation examinations and analyses of U-Mo/Al dispersion mini plates are presented. Irradiation test RERTR-5 contained mini- fuel plates with fuel loadings of 6 and 8 g U cm -3 . The fuel material consisted of 6, 7 and 10 wt. % Mo-uranium-alloy powders in atomized and machined form. The swelling behavior of the various fuel types is analyzed, indicating athermal swelling of the U-Mo alloy and temperature-dependent swelling owing to U-Mo/Al interdiffusion. (author)

  20. Effect of reduced enrichment on the fuel cycle for research reactors

    International Nuclear Information System (INIS)

    Travelli, A.

    1982-01-01

    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

  1. Transport of high enriched uranium fresh fuel from Yugoslavia to the Russian federation

    OpenAIRE

    Pešić Milan P.; Šotić Obrad; Hopwood William H.Jr

    2002-01-01

    This paper presents the relevant data related to the recent shipment (August 2002) of fresh highly enriched uranium fuel elements from Yugoslavia back to the Russian Federation for uranium down blending. In this way, Yugoslavia gave its contribution to the Reduced Enrichment for Research and Test Reactors (RERTR) Program and to the world's joint efforts to prevent possible terrorist actions against nuclear material potentially usable for the production of nuclear weapons.

  2. Transport of high enriched uranium fresh fuel from Yugoslavia to the Russian federation

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2002-01-01

    Full Text Available This paper presents the relevant data related to the recent shipment (August 2002 of fresh highly enriched uranium fuel elements from Yugoslavia back to the Russian Federation for uranium down blending. In this way, Yugoslavia gave its contribution to the Reduced Enrichment for Research and Test Reactors (RERTR Program and to the world's joint efforts to prevent possible terrorist actions against nuclear material potentially usable for the production of nuclear weapons.

  3. Status of the RERTR [Reduced Enrichment Research and Test Reactor] program in Argentina

    International Nuclear Information System (INIS)

    Giorsetti, D.R.

    1987-01-01

    The Argentine Atomic Energy Commission started in 1978 the Reduced Enrichment Research and Test Reactors in the field of reactor engineering; engineering, development and manufacturing of fuel elements and research reactors operators. This program was initiated with the conviction that it would contribute to the international efforts to reduce risks of nuclear weapons proliferation owing to an uncontrolled use of highly enriched uranium. It was intended to convert RA-3 reactor to make possible its operation with low enriched fuel (LEU), instead of high enriched fuel (HEU) and to develop manufacturing techniques for said LEU. Afterwards, this program was adapted to assist other countries in reactors conversion, development of the corresponding fuel elements and supply of fuel elements to other countries. (Author)

  4. Fuel conversion of JRR-4 from HEU to LEU

    International Nuclear Information System (INIS)

    Ichikawa, Hiroki; Nakajima, Teruo

    1997-01-01

    Japanese JRR-4 (Japan Research Reactor No.4) is a pool type, light water moderated and cooled, ETR type fuel reactor used for Shielding experiments, isotope production, neutron activation analyses, Si doping, reactor students training. It acieved first criticality on January 28, 1965 with maximum thermal power 3.5MW. The standard core consistes of 20 Fuel elements, 7 control rods 5 Irradiation holes, neutron source, graphite reflectors. Available thermal flux is 7x1013 n/cm2/s. Within the RERTR program plans are made for core conversion from HEU to LEU

  5. The U.S. reduced enrichment research and test reactor (RERTR) program

    International Nuclear Information System (INIS)

    Travelli, A.

    1993-01-01

    Research and test reactors are widely deployed to study the irradiation behavior of materials of interest in nuclear engineering, to produce radioisotopes for medicine, industry, and agriculture, and as a basic research and teaching tool. In order to maximize neutron flux per unit power and/or to minimize capital costs and fuel cycle costs, most of these reactors were de- signed to utilize uranium with very high enrichment (in the 70% to 95% range). Research reactor fuels with such high uranium enrichment cause a potential risk of nuclear weapons proliferation. Over 140 research and test reactors of significant power (between 10 kW and 250 MW) are in operation with very highly enriched uranium in more than 35 countries, with total power in excess of 1,700 MW. The overall annual fuel requirement of these reactors corresponds to approximately 1,200 kg of 235 U. This highly strategic material is normally exported from the United States, converted to metal form, shipped to a fuel fabricator, and then shipped to the reactor site in finished fuel element form. At the reactor site the fuel is first stored, then irradiated, stored again, and eventually shipped back to the United States for reprocessing. The whole cycle takes approximately four years to complete, bringing the total required fuel inventory to approximately 5,000 kg of 235 U. The resulting international trade in highly-enriched uranium may involve several countries in the process of refueling a single reactor and creates a considerable concern that the highly-enriched uranium may be diverted for non-peaceful purposes while in fabrication, transport, or storage, particularly when it is in the unirradiated form. The proliferation resistance of nuclear fuels used in research and test reactors can be considerably improved by reducing their uranium enrichment to a value less than 20%, but significantly greater than natural to avoid excessive plutonium production

  6. Description of ECRI (CNEA'S MTR fuel fabrication plant)

    International Nuclear Information System (INIS)

    Echenique, P.; Fabro, J.; Podesta, D.; Restelli, M.; Rossi, G.; Alvarez, L.; Adelfang, P.

    2002-01-01

    The ECRI Plant is dedicated to the development and fabrication of high-density fuel elements and targets for 99 Mo. In this sector had been done the start up Fuel Elements for the Reactors of Peru, Iran, Algeria and Egypt. All of them were made with U 3 O 8 . The targets for 99 Mo using HEU were fabricated too in the last years. The new material of high-density for Fuel Elements as U 3 Si 2 were done in this sector, three prototypes were fabricated, two are still under irradiation. (P06 and P07). As new developments we are working with U-Mo (7%) Fuel Plates with both material Korean and HMD. This work is under the RERTR Program and two fuel elements, manufactured by us, with both powders, will be irradiated in Petten. For 99 Mo targets, we are fabricating miniplates of LEU with an AlUx powder by pulvi-metallurgy technique. And it is under development the foils targets under the RERTR Program. A general view of the fabrication facilities and control sector will be shown. The different operations that are done in each sector will be explained. All our activities will be certified under the ISO 9000 and we are working hard to get it in the middle of 2003. (author)

  7. The United States spent fuel acceptance policy - A year in review

    International Nuclear Information System (INIS)

    Huizenga, David G.

    1997-01-01

    Through the combined efforts of the United States and many other nations participating in the Reduced Enrichment for Research and Test Reactors (RERTR) program, much progress has been made toward reducing the amount of highly enriched uranium (HEU) in international commerce. A little more than a year ago, the U.S. Department of Energy adopted a new 1 0-year policy to accept research reactor spent nuclear fuel into the United States from other nations. The policy supports U.S. nuclear weapons nonproliferation objectives and demonstrates the continued commitment of the U.S. to the RERTR program. This paper is a review of the past year's activities and addresses the progress made since the policy was implemented

  8. Microstructure of RERTR DU-alloys irradiated with krypton ions up to 100 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D., E-mail: Dennis.Keiser@inl.gov [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D.; Wachs, D.M. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Allen, T.R. [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.; Rest, J. [Materials Science Division, Argonne National Laboratory, 9700 S. Cass Ave, Argonne, IL 60439 (United States)

    2011-04-15

    The radiation stability of the interaction product formed at the fuel-matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al){sub 3}, (U, Mo)(Si, Al){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43}, and UAl{sub 4}. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 deg. C to doses up to {approx}100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al){sub 3} and UAl{sub 4} phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al){sub 3} and UMo{sub 2}Al{sub 20} phases become amorphous at 1 and {approx}2 dpa, respectively, and show no evidence of voids at 100 dpa. The U{sub 6}Mo{sub 4}Al{sub 43} phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

  9. Microstructure of RERTR DU-alloys irradiated with krypton ions up to 100 dpa

    Science.gov (United States)

    Gan, J.; Keiser, D. D., Jr.; Miller, B. D.; Wachs, D. M.; Allen, T. R.; Kirk, M.; Rest, J.

    2011-04-01

    The radiation stability of the interaction product formed at the fuel-matrix interface of research reactor dispersion fuels, under fission-product bombardment, has a strong impact on fuel performance. Three depleted uranium alloys were cast that consisted of the following five phases to be investigated: U(Si, Al) 3, (U, Mo)(Si, Al) 3, UMo 2Al 20, U 6Mo 4Al 43, and UAl 4. Irradiation of transmission electron microscopy (TEM) disc samples with 500-keV Kr ions at 200 °C to doses up to ˜100 displacements per atom (dpa) were conducted using a 300-keV electron microscope equipped with an ion accelerator. TEM results show that the U(Si, Al) 3 and UAl 4 phases remain crystalline at 100 dpa without forming voids. The (U, Mo)(Si, Al) 3 and UMo 2Al 20 phases become amorphous at 1 and ˜2 dpa, respectively, and show no evidence of voids at 100 dpa. The U 6Mo 4Al 43 phase goes to amorphous at less than 1 dpa and reveals high density voids at 100 dpa.

  10. DEVELOPMEMT OF PILOT SCALE DEALUMINATION UNIT OF 2.5 ...

    African Journals Online (AJOL)

    ... sulphuric acid at the reaction temperature > 140°C. The capacity of the componentsare; acid holding tank 5 liters;metakaolin slurry mixing tank 20 liters, with 70 W capacity motor and the dealuminator 50 liters, which had pressure relief valve, pressure gauge (0-10 bars) and temperature gauge (0 - 300°C) on the cover.

  11. DEVELOPMEMT OF LIQUIDITY REGULATION INSTRUMENTS OF REGIONAL BANK SECTOR ORGANIZATIONS

    Directory of Open Access Journals (Sweden)

    A. V. Roussavskaya

    2012-01-01

    Full Text Available Discussed in the article is a model of an improved regulatory framework for credit institutions dealing with investment projects implementation of which will make it possible for business enterprises of various economy sectors whose activities are dependent on investment resources, such as small-scale and medium-scale as well as innovation business enterprises, to reach in the coming years an entirely different level of development.

  12. DEVELOPMEMT OF PILOT SCALE DEALUMINATION UNIT OF 2.5 ...

    African Journals Online (AJOL)

    HOD

    alumina which reacted with the sulphuric acid was obtained in liquid form as aluminum ... dealumination unit comprises of the dealumination reactor, known as ... Clay catalysts have attracted much interest in catalytic ... since the dealuminator is not being designed as a ... of gases), a safety factor of 25% was chosen.

  13. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, New York (United States)

    2007-07-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat.

  14. Performance Evaluation of Metallic Dispersion Fuel for Advanced Research Reactors

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Kim, Chang Kyu; Chae, Hee Taek; Song, Kee Chan; Kim, Yeon Soo

    2007-01-01

    Uranium alloys with a high uranium density has been developed for high power research reactor fuel using low-enriched uranium (LEU). U-Mo alloys have been developed as candidate fuel material because of excellent irradiation behavior. Irradiation behavior of U-Mo/Al dispersion fuel has been investigated to develop high performance research reactor fuel as RERTR international research program. While plate-type and rod-type dispersion fuel elements are used for research reactors, HANARO uses rod-type dispersion fuel elements. PLATE code is developed by Argonne National Laboratory for the performance evaluation of plate-type dispersion fuel, but there is no counterpart for rod-type dispersion fuel. Especially, thermal conductivity of fuel meat decreases during the irradiation mainly because of interaction layer formation at the interface between the U-Mo fuel particle and Al matrix. The thermal conductivity of the interaction layer is not as high as the Al matrix. The growth of interaction layer is interactively affected by the temperature of fuel because it is associated with a diffusion reaction which is a thermally activated process. It is difficult to estimate the temperature profile during irradiation test due to the interdependency of fuel temperature and thermal conductivity changed by interaction layer growth. In this study, fuel performance of rod-type U-Mo/Al dispersion fuels during irradiation tests were estimated by considering the effect of interaction layer growth on the thermal conductivity of fuel meat

  15. The series production in a standardized fabrication line for silicide fuels and commercial aspects

    International Nuclear Information System (INIS)

    Wehner, E.L.; Hassel, H.W.

    1987-01-01

    NUKEM has been responsible for the development and fabrication of LEU fuel elements for MTR reactors under the frame of the German AF program since 1979. The AF program is part of the international RERTR efforts, which were initiated by the INFCE Group in 1978. This paper describes the actual status of development and the transition from the prototype to the series production in a standardized manufacturing line for silicide fuels at NUKEM. Technical provisions and a customer oriented standardized product range aim at an economized manufacturing. (Author)

  16. Modeling a failure criterion for U-Mo/Al dispersion fuel

    Science.gov (United States)

    Oh, Jae-Yong; Kim, Yeon Soo; Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik

    2016-05-01

    The breakaway swelling in U-Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U-Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

  17. Modeling a failure criterion for U–Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jae-Yong, E-mail: tylor@kaeri.re.kr [Korea Atomic Energy Research Institute, 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Kim, Yeon Soo [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Tahk, Young-Wook; Kim, Hyun-Jung; Kong, Eui-Hyun; Yim, Jeong-Sik [Korea Atomic Energy Research Institute, 111, Daedeok-Daero 989 Beon-Gil, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

    2016-05-15

    The breakaway swelling in U–Mo/Al dispersion fuel is known to be caused by large pore formation enhanced by interaction layer (IL) growth between fuel particles and Al matrix. In this study, a critical IL thickness was defined as a criterion for the formation of a large pore in U–Mo/Al dispersion fuel. Specifically, the critical IL thickness is given when two neighboring fuel particles come into contact with each other in the developed IL. The model was verified using the irradiation data from the RERTR tests and KOMO-4 test. The model application to full-sized sample irradiations such as IRISs, FUTURE, E-FUTURE, and AFIP-1 tests resulted in conservative predictions. The parametric study revealed that the fuel particle size and the homogeneity of the fuel particle distribution are influential for fuel performance.

  18. Nuclear fuel management in JMTR

    International Nuclear Information System (INIS)

    Naka, Michihiro; Miyazawa, Masataka; Sato, Hiroshi; Nakayama, Fusao; Ito, Haruhiko

    1999-01-01

    The Japan Materials Testing Reactor (JMTR) is the largest scale materials (author)ted the fission gas release compared with the steady state opkW/l in Japan. JMTR as a multi-purpose reactor has been contributing to research and development on nuclear field with a wide variety of irradiation for performing engineering tests and safety research on fuel and component for light water reactor as well as fast breeder reactor, high temperature gas-cooled reactor etc., for research and development on blanket material for fusion reactor, for fundamental research, and for radio-isotope (RI) production. The driver nuclear fuel used in JMTR is aluminum based MTR type fuel. According to the Reduced Enrichment for Research and Test Reactors (RERTR) Program, the JMTR fuel elements had been converted from 93% high enriched uranium (HEU) fuel to 45% medium enriched uranium (MEU) fuel in 1986, and then to 20% low enriched uranium (LEU) fuel in 1994. The cumulative operation cycles until March 1999 reached to 127 cycles since the first criticality in 1968. JMTR has used 1,628 HEU, 688 MEU and 308 LEU fuel elements for these operation cycles. After these spent fuel elements were cooled in the JMTR water canal more than one year after discharged from the JMTR core, they had been transported to reprocessing plants in Europe, and then to plants in USA in order to extract the uranium remaining in the spent fuel. The JMTR spent fuel transportation for reprocessing had been continued until the end of 1988. However, USA had ceased spent fuel reprocessing in 1989, while USDOE committed to prepare an environmental review of the impacts of accepting spent fuels from foreign research reactors. After that, USDOE decided to implement a new acceptance policy in 1996, the spent fuel transportation from JMTR to Savannah River Site was commenced in 1997. It was the first transportation not only in Japan but in Asia also. Until resuming the transportation, the spent fuel elements stored in JMTR

  19. The irradiation behavior of atomized U-Mo alloy fuels at high temperature

    Science.gov (United States)

    Park, Jong-Man; Kim, Ki-Hwan; Kim, Chang-Kyu; Meyer, M. K.; Hofman, G. L.; Strain, R. V.

    2001-04-01

    Post-irradiation examinations of atomized U-10Mo, U-6Mo, and U-6Mo-1.7Os dispersion fuels from the RERTR-3 experiment irradiated in the Advanced Test Reactor (ATR) were carried out in order to investigate the fuel behavior of high uranium loading (8 gU/cc) at a high temperature (higher than 200°C). It was observed after about 40 at% BU that the U-Mo alloy fuels at a high temperature showed similar irradiation bubble morphologies compared to those at a lower temperature found in the RERTR-1 irradiation result, but there was a thick reaction layer with the aluminum matrix which was found to be greatly affected by the irradiation temperature and to a lesser degree by the fuel composition. In addition, the chemical analysis for the irradiated U-Mo fuels using the Electron Probe Micro Analysis (EPMA) method were conducted to investigate the compositional changes during the formation of the reaction product.

  20. Fuel performance of rod-type research reactor fuel using a centrifugally atomized U-Mo powder

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Man; Lee, Yoon Sang; Kim, Chang Kyu

    2009-01-01

    A low enriched uranium nuclear fuel for research reactors has been developed in order to replace a highly enriched uranium fuel according to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program. In KAERI, a rod-type U 3 Si dispersion fuel has been developed for a localization of the HANARO fuel and a U 3 Si/Al dispersion fuel of 3.15 gU/cc has been used at HANARO as a driver fuel since 2005. Although uranium silicide dispersion fuels such as U 3 Si 2 /Al and U 3 Si/Al are being used widely, high uranium density dispersion fuels (8-9 g/cm 3 ) are required for some high performance research reactors. U-Mo alloys have been considered as one of the most promising uranium alloys for a dispersion fuel due to their good irradiation performance. An international qualification program on U-Mo fuel to replace a uranium silicide dispersion fuel with a U-Mo dispersion fuel has been carried out

  1. Present status of JMTR spent fuel shipment

    International Nuclear Information System (INIS)

    Miyazawa, Masataka; Watanabe, Masao; Yokokawa, Makoto; Sato, Hiroshi; Ito, Haruhiko

    2002-01-01

    The Japan Atomic Energy Research Institute (JAERI) has been consistently making the enrichment reduction of reactor fuels in cooperation with RERTR Program and FRR SNF Acceptance Program both conducted along with the U.S. Nuclear Non-Proliferation Policy and JMTR, 50 MW test reactor in Oarai Research Establishment, has achieved core conversion, from its initial 93% enriched UAl alloy to 45% enriched uranium-aluminide fuel, and then to the current 19.8% enriched uranium-silicide fuel. In order to return all of JMTR spent fuels, to be discharged from the reactor by May 12, 2006, to the U.S.A. by May 12, 2009, JAERI is planning the transportation schedule based on one shipment per year. The sixth shipment of spent fuels to U.S. was carried out as scheduled this year, where the total number of fuels shipped amounts to 651 elements. All of the UAl alloy elements have so far been shipped and now shipments of 45% enriched uranium-aluminide type fuels are in progress. Thus far the JMTR SFs have been transported on schedule. From 2003 onward are scheduled more then 850 elements to be shipped. In this paper, we describe our activities on the transportation in general and the schedule for the SFs shipments. (author)

  2. From high to low. The IAEA is helping to reduce the use of high-risk nuclear fuel at the world's research reactors

    International Nuclear Information System (INIS)

    Adelfang, P.; Goldman, I.

    2006-01-01

    Research reactors play a key role in the development of peaceful uses of atomic energy. They are used for the production of isotopes for medicine and industry, for research in physics, biology and materials science, and for scientific education and training. They also continue to play an important role in support of nuclear power programmes. The IAEA's data shows there are 249 operational research reactors worldwide. Of these, more than 100 reactors are still fuelled with highly enriched uranium (HEU). It is considered high-risk nuclear material since it can be easily used for a nuclear explosive device. As part of a developing international norm to minimize and eventually eliminate HEU in civilian nuclear applications, research reactor operators increasingly are working with national and international agencies. They are being encouraged and supported to improve their physical security arrangements, convert their reactors to low-enriched uranium (LEU) fuel, and ship irradiated fuel back to the country of origin.For more than twenty years the IAEA has been supporting international efforts associated with reducing the amount of HEU in international commerce. Projects and activities have directly supported a programme the United States initiated in 1978, called Reduced Enrichment for Research and Test Reactors (RERTR). The IAEA's work additionally supports efforts to return research reactor fuel to the country where it was originally enriched so-called take back activities. IAEA initiatives have included the development and maintenance of several databases with information related to research reactors and research reactor spent fuel inventories. These databases have been essential in planning and managing both RERTR and take-back programmes. Other Agency activities through technical cooperation and other channels have supported the conversion of research reactors to using lower enriched fuels. In other ways, the IAEA supports the exchange of information among experts

  3. Handling of spent fuel from research reactors in Japan

    International Nuclear Information System (INIS)

    Kanda, K.

    1997-01-01

    In Japan eleven research reactors are in operation. After the 19th International Meeting on Reduced Enrichment for Research Reactors and Test Reactors (RERTR) on October 6-10, 1996, Seoul, Korea, the Five Agency Committee on Highly Enriched Uranium, which consists of Science and Technology Agency, the Ministry of Education, Science and Culture, the Ministry of Foreign Affairs, Japan Atomic Energy Research Institute (JAERI) and Kyoto University Research Reactor Institute (KURRI) met on November 7,1996, to discuss the handling of spent fuel from research reactors in Japan. Advantages and disadvantages to return spent fuel to the USA in comparison to Europe were discussed. So far, a number of spent fuel elements in JAERI and KURRI are to be returned to the US. The first shipment to the US is planned for 60 HEU elements from JMTR in 1997. The shipment from KURRI is planned to start in 1999. (author)

  4. Update on Fresh Fuel Characterization of U-Mo Alloys

    International Nuclear Information System (INIS)

    Burkes, D.E.; Wachs, D.M.; Keiser, D.D.; Okuniewski, M.A.; Jue, J.F.; Rice, F.J.; Prabhakaran, R.

    2009-01-01

    The need to provide more accurate property information on U-Mo fuel alloys to operators, modellers, researchers, fabricators, and government increases as success of the GTRI Reactor Convert program continues. This presentation provides an update on fresh fuel characterization activities that have occurred at the INL since the RERTR 2008 conference in Washington, D.C. The update is particularly focused on properties recently obtained and on the development progress of new measurement techniques. Furthermore, areas where useful and necessary information is still lacking is discussed. The update deals with mechanical, physical, and microstructural properties for both integrated and separate effects. Appropriate discussion of fabrication characteristics, impurities, thermodynamic response, and effects on the topic areas are provided, along with a background on the characterization techniques used and developed to obtain the information. Efforts to measure similar characteristics on irradiated fuel plates are discussed.

  5. A Prediction Study of Aluminum Alloy Oxidation of the Fuel Cladding in Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tahk, Y. W.; Oh, J. Y.; Lee, B. H.; Seo, C. G.; Chae, H. T.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    U{sub 3}Si{sub 2}-Al dispersion fuel with Al cladding will be used for Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding experiences the oxidation layer growth on the surface during the reactor operation. The formation of oxides on the cladding affects fuel performance by increasing fuel temperature. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, a fresh fuel is discharged after 900 effective full power days (EFPD) with 18 cycles of 50 days loading. For the proper prediction of the aluminum oxide thickness of fuel cladding during the long residence time, a reliable model is needed. In this work, several oxide thickness prediction models are compared with the measured data from in-pile test by RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model are performed for JRTR fuel

  6. Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Snelgrove, J.L.; Cornella, R.J.

    1983-01-01

    A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel ( 235 U enrichment > 90%) with low-enriched uranium (LEU) fuel ( 235 U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

  7. Methodology for thermal-hydraulics analysis of pool type MTR fuel research reactors

    International Nuclear Information System (INIS)

    Umbehaun, Pedro Ernesto

    2000-01-01

    This work presents a methodology developed for thermal-hydraulic analysis of pool type MTR fuel research reactors. For this methodology a computational program, FLOW, and a model, MTRCR-IEAR1 were developed. FLOW calculates the cooling flow distribution in the fuel elements, control elements, irradiators, and through the channels formed among the fuel elements and among the irradiators and reflectors. This computer program was validated against experimental data for the IEA-R1 research reactor core at IPEN-CNEN/SP. MTRCR-IEAR1 is a model based on the commercial program Engineering Equation Solver (EES). Besides the thermal-hydraulic analyses of the core in steady state accomplished by traditional computational programs like COBRA-3C/RERTR and PARET, this model allows to analyze parallel channels with different cooling flow and/or geometry. Uncertainty factors of the variables from neutronic and thermalhydraulic calculation and also from the fabrication of the fuel element are introduced in the model. For steady state analyses MTRCR-IEAR1 showed good agreement with the results of COBRA-3C/RERTR and PARET. The developed methodology was used for the calculation of the cooling flow distribution and the thermal-hydraulic analysis of a typical configuration of the IEA-R1 research reactor core. (author)

  8. Thermal behavior analysis of U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

  9. Thermal behavior analysis of U-Mo/Al dispersion fuel

    International Nuclear Information System (INIS)

    Ryu, Ho Jin; Park, Jong Mang; Lee, Yoon Sang; Kim, Chang Kyu

    2004-01-01

    According to the non-proliferation policy under the reduced enrichment for research and test reactors (RERTR) program, low enriched uranium(LEU) fuel such as uranium silicide dispersion fuels are being used in research reactors. Because of a lower enrichment higher uranium density fuels are required for some high performance research reactors. Some uranium alloys with a high uranium density such as U-Mo alloys have been considered as one of the most promising candidates for a dispersion fuel due to the good irradiation performance. An international qualification program to replace the uranium silicide dispersion fuel with U-Mo dispersion fuel is being carried out under the RERTR program. Although U-Mo powders are conventionally supplied by the mechanical comminuting of as-cast U-Mo alloys, KAERI developed a centrifugal atomization method in order to simplify the preparation process and improve the properties. The centrifugally atomized powders have a rapidly solidified gamma uranium structure and a spherical shape. During the in-reactor operation of a dispersion fuel, interdiffusion or chemical reactions between the fuel particles and the matrix occurr. Intermetallic compounds in the form of UAlx are formed as a result of the diffusional reaction. Because the intermetallic compounds are less dense than the combined reactants, the volume of the fuel element increases after the reaction. In addition to the effect on the swelling performance, the reaction layers between the U-Mo and the Al matrix induces a degradation of the thermal properties of the U-Mo/Al dispersion fuels. It is important to investigate the thermal behavior of U-Mo/Al dispersion fuel according to reaction between the fuel particles and the matrix with the burnup and linear power. In this study, a finite element analysis was used for the calculation of the temperature distribution of the U-Mo/Al dispersion fuel with a burnup and linear power. Kinetics data of the reaction layers such as the growth

  10. Development of very-high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snegrove, J.L.; Hofmann, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-01-01

    The RERTR (=Reduced Enrichment for Research and Test Reactors) program has begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place, and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and the first results should be available by the end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun. (author)

  11. Development of very high-density low-enriched uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Trybus, C.L.; Wiencek, T.C.

    1997-02-01

    The RERTR program has recently begun an aggressive effort to develop dispersion fuels for research and test reactors with uranium densities of 8 to 9 g U/cm 3 , based on the use of γ-stabilized uranium alloys. Fabrication development teams and facilities are being put into place and preparations for the first irradiation test are in progress. The first screening irradiations are expected to begin in late April 1997 and first results should be available by end of 1997. Discussions with potential international partners in fabrication development and irradiation testing have begun

  12. Scope and dissolution studies and characterization of irradiated nuclear fuel in Atalante Hot Cell Facilities (abstract and presentation slides)

    Energy Technology Data Exchange (ETDEWEB)

    Dancausse, Jean-Philippe; Reynier Tronche, Nathalie; Ferlay, Gilles; Herlet, Nathalie; Eysseric, Cathrine; Esbelin, Eric

    2005-01-01

    Since 1999, several studies on nuclear fuels were realised in C11/C12 Atalante Hot Cell. This paper presents firstly an overview of the apparatus used for fuel dissolution and characterisation like reactor design, gas trapping flask and solid/liquid separation. Then, the general methodology is described as a function of fuel, temperature, reagents, showing for each step, the reachable experimental data: Dissolution rate, chemical and radiochemical fuel composition including volatile LLRN, insoluble mass, composition, morphology, cladding chemical, radiochemical and physical characterisation using SIMS (made in Cadarache/LECA facilities), MEB. To conclude, some of the obtained results on 129I and 14C composition of oxide fuels, rate of dissolution and first results on dissolution studies of RERTR UMo fuel will be detailed. (Author)

  13. LEU WWR-M2 fuel assemblies burnable test

    International Nuclear Information System (INIS)

    Kirsanov, G.A.; Konoplev, K.A.; Pikulik, R.G.; Sajkov, Yu. P.; Tchmshkyan, D.V.; Tedoradze, L.V.; Zakharov, A.S.

    2000-01-01

    The results of in-pile irradiation tests of LEU WWR-M2 fuel assemblies with reduced enrichment of fuel are submitted in the report. The tests are made according to the Russian Program on Reduced Enrichment for Research and Test Reactors (RERTR). United States Department of Energy and the Ministry of Atomic Energy of Russian Federation jointly fund this Program. The irradiation tests of 5 WWR-M2 experimental assemblies are carried out at WWR-M reactor of the Petersburg Nuclear Physics Institute (PNPI). The information on assembly design and technique of irradiation tests is presented. In the irradiation tests the integrity of fuel assemblies is periodically measured. The report presents the data for the integrity maintained during the burnup of 5 fuel assemblies up to 45%. These results demonstrate the high reliability of the experimental fuel assemblies within the guaranteed burnup limits specified by the manufacturer. The tests are still in progress; it is planned to test and analyze the change in integrity for burnup of up to 70% - 75% or more. LEU WWR-M2 fuel assemblies are to be offered for export by their Novosibirsk manufacturer. Currently, HEU WWR-M2 fuel assemblies are used in Hungary, Ukraine and Vietnam. LEU WWR-M2 fuel assemblies were designed as a possible replacement for the HEU WWR-M2 fuel assemblies in those countries, but their use can be extended to other research reactors. (author)

  14. Conversion of research and test reactors to low enriched uranium fuel: technical overview and program status

    International Nuclear Information System (INIS)

    Roglans-Ribas, J.

    2008-01-01

    Many of the nuclear research and test reactors worldwide operate with high enriched uranium fuel. In response to worries over the potential use of HEU from research reactors in nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel by converting research reactors to low enriched uranium (LEU) fuel. The Reactor Conversion program is currently under the DOE's National Nuclear Security Administration's Global Threat Reduction Initiative (GTRI). 55 of the 129 reactors included in the scope have been already converted to LEU fuel or have shutdown prior to conversion. The major technical activities of the Conversion Program include: (1) the development of advanced LEU fuels; (2) conversion analysis and conversion support; and (3) technology development for the production of Molybdenum-99 (Mo 99 ) with LEU targets. The paper provides an overview of the status of the program, the technical challenges and accomplishments, and the role of international collaborations in the accomplishment of the Conversion Program objectives. Nuclear research and test reactors worldwide have been in operation for over 60 years. Many of these facilities operate with high enriched uranium fuel. In response to increased worries over the potential use of HEU from research reactors in the manufacturing of nuclear weapons, the U.S Department of Energy (DOE) initiated a program - the Reduced Enrichment for Research and Test Reactors (RERTR) - in 1978 to develop the technology necessary to reduce the use of HEU fuel in research reactors by converting them to low enriched uranium (LEU) fuel. The reactor conversion program was initially focused on U.S.-supplied reactors, but in the early 1990s it expanded and began to collaborate with Russian institutes with the objective of converting Russian supplied reactors to the use of LEU fuel.

  15. Shifting to non-explosive fuels for research reactors

    International Nuclear Information System (INIS)

    Gilinsky, Victor

    1984-01-01

    The RERTR program is not just an American program, it is an international program and it can succeed only with wide support and participation. Excellent work underway at research laboratories in several countries is making a vital contribution. The cooperative spirit shown by all participants is particularly gratifying. Some practical difficulties may be encountered with the safety regulatory agencies in different countries when modifying the reactors. The US NRC intends to demonstrate that conversion to low enriched fuel is not a difficult process by taking steps to enable domestic reactors operating in the United States to convert to low enriched fuels. A proposed regulation is being prepared limiting the use of highly enriched uranium in domestic reactors. In this connection, The US NRC will be prepared tp cooperate with reactor operators from other countries in the safety area as it relates to the conversion process

  16. Development of low enrichment technologies for high density fuels and for isotope production targets

    International Nuclear Information System (INIS)

    Taboada, Horacio; Gonzalez, Alfredo G.

    2005-01-01

    Since more than twenty years ago, CNEA has carried out RERTR activities. Main goals are to convert the RA 6 reactor core from HEU to LEU, to get a comprehensive understanding of U-Mo/Al compounds phase formation in dispersed and monolithic fuels, to develop possible solutions to VHD dispersed and monolithic fuels technical problems, and to optimize techniques to recover U from silicide scrap samples. The future plans include: 1) Completion the RA 6 reactor conversion to LEU; 2) Qualification by irradiation of the promising solutions found for the high density fuels; 3) Irradiation of mini plates and full scale fuel assemblies at the RA 3 reactor and at higher flux and temperature reactors; 4) Optimization of LEU target and radiochemical techniques for radioisotope production. (author) [es

  17. Application of the DART Code for the Assessment of Advanced Fuel Behavior

    International Nuclear Information System (INIS)

    Rest, J.; Totev, T.

    2007-01-01

    The Dispersion Analysis Research Tool (DART) code is a dispersion fuel analysis code that contains mechanistically-based fuel and reaction-product swelling models, a one dimensional heat transfer analysis, and mechanical deformation models. DART has been used to simulate the irradiation behavior of uranium oxide, uranium silicide, and uranium molybdenum aluminum dispersion fuels, as well as their monolithic counterparts. The thermal-mechanical DART code has been validated against RERTR tests performed in the ATR for irradiation data on interaction thickness, fuel, matrix, and reaction product volume fractions, and plate thickness changes. The DART fission gas behavior model has been validated against UO 2 fission gas release data as well as measured fission gas-bubble size distributions. Here DART is utilized to analyze various aspects of the observed bubble growth in U-Mo/Al interaction product. (authors)

  18. Neutron physics computation of CERCA fuel elements for Maria Reactor

    International Nuclear Information System (INIS)

    Andrzejewski, K.J.; Kulikowska, T.; Marcinkowska, Z.

    2008-01-01

    Neutron physics parameters of CERCA design fuel elements were calculated in the framework of the RERTR (Reduced Enrichment for Research and Test Reactors) program for Maria reactor. The analysis comprises burnup of experimental CERCA design fuel elements for 4 cycles in Maria Reactor To predict the behavior of the mixed core the differences between the CERCA fuel (485 g U-235 as U 3 Si 2 , 5 fuel tubes, low enrichment 19.75 % - LEU) and the presently used MR-6 fuel (430 g as UO 2 , 6 fuel tubes, high enrichment 36 % - HEU) had to be taken into account. The basic tool used in neutron-physics analysis of Maria reactor is program REBUS using in its dedicated libraries of effective microscopic cross sections. The cross sections were prepared using WIMS-ANL code, taking into account the actual structure, temperature and material composition of the fuel elements required preparation of new libraries.The problem is described in the first part of the present paper. In the second part the applicability of the new library is shown on the basis of the fuel core computational analysis. (author)

  19. Low-enriched uranium-molybdenum fuel plate development

    International Nuclear Information System (INIS)

    Wiencek, T.C.; Prokofiev, I.G.

    2000-01-01

    To examine the fabricability of low-enriched uranium-molybdenum powders, full-size 450 x 60 x 0.5-mm (17.7 x 2.4 x 0.020-in.) fuel zone test plates loaded to 6 g U/cm 3 were produced. U-10 wt.% Mo powders produced by two methods, centrifugal atomization and grinding, were tested. These powders were supplied at no cost to Argonne National Laboratory by the Korean Atomic Energy Research Institute and Atomic Energy of Canada Limited, respectively. Fuel homogeneity indicated that both of the powders produced acceptable fuel plates. Operator skill during loading of the powder into the compacting die and fuel powder morphology were found to be important when striving to achieve homogeneous fuel distribution. Smaller, 94 x 22 x 0.6-mm (3.7 x 0.87 x 0.025-in.) fuel zone, test plates were fabricated using U-10 wt.% Mo foil disks instead of a conventional powder metallurgy compact. Two fuel plates of this type are currently undergoing irradiation in the RERTR-4 high-density fuel experiment in the Advanced Test Reactor. (author)

  20. Relative neutronic performance of proposed high-density dispersion fuels in water-moderated and D2O-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.; Snelgrove, J.L.

    1996-01-01

    This paper provides an overview of the neutronic performance of an idealized research reactor using several high density LEU fuels that are being developed by the RERTR program. High-density LEU dispersion fuels are needed for new and existing high-performance research reactors and to extend the lifetime of fuel elements in other research reactors. This paper discusses the anticipated neutronic behavior of proposed advanced fuels containing dispersions of U 3 Si 2 , UN, U 2 Mo and several uranium alloys with Mo, or Zr and Nb. These advanced fuels are ranked based on the results of equilibrium depletion calculations for a simplified reactor model having a small H 2 O-cooled core and a D 2 O reflector. Plans have been developed to fabricate and irradiate several uranium alloy dispersion fuels in order to test their stability and compatibility with the matrix material and to establish practical loading limits

  1. Development of nuclear fuel materials for research reactor

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Park, H. D.; Kim, K. H.; Lee, J. T.; Ryu, W. S.; Hwang, W.; Kim, H. N.; Kim, H. I.; Kwon, H. I.; Park, C.; Lee, B. C.; Park, J. M.; Lee, C. S.; Chae, H. T.; Im, N. J.; Cho, M. S.; Im, I. C.; Nam, C.; Lee, D. B.; Goh, Y. M.; Kim, J. D.; Ahn, H. S.; Woo, Y. M.; Chang, S. J.; Cho, H. D.

    1997-09-01

    This project has aimed at the development of U 3 Si dispersion fuel for the localization of HANARO fuel and the application of atomization process to advanced RERTR fuel development. The design criteria were established through the modified computer codes. Design documents were prepared and issued. The acceptable co-extrusion cladding was achieved. The electron beam welding technology has been developed and the sealing of the end plug and cladding was accomplished without defects. The atomization fuel meats have about 200% higher elongation and about 20% higher than comminution fuel meats. The thermal compatibility test showed that atomization fuel have about 30% higher stability that the comminution fuel. The pressure drops of 18 rods fuel assembly and 36 rods fuel assembly were measured to have 213 kPa and 205 kPa respectively. Apparent wear was not found in endurance test. The irradiation fuel was designed and fabricated by using low enriched uranium metal following the developed Q/A system. The safety analysis of irradiation fuel assembly was performed through linear power calculation by using MCNP4A code and centerline temperature calculation by using DIFAIR code. The quality assurance system has been established. The quality inspection technologies were developed. By acquiring the license, low enriched uranium of 100 kg as well as depleted uranium can be used. U 3 Si 2 -Al fuel swelled less than comminution fuel irrespective of temperature and fuel fraction in a compatibility test. The atomized U-10wt.%Mo powder were found to have gamma phase of isotropic structure. Gamma structure remained with a little swelling without any structure change at 400 deg C for 100 hours. Irradiation miniplate and test rig were designed preliminary manufactured. Thermal hydraulic and linear power calculations were performed by using PLTEMP and MCNP4A computer codes respectively. The hydraulic test showed that the pressure drop met the HANARO requirement. The vibration

  2. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    International Nuclear Information System (INIS)

    Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes

    2011-01-01

    For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  3. Powder fabrication of U-Mo alloys for nuclear dispersion fuels

    Energy Technology Data Exchange (ETDEWEB)

    Durazzo, Michelangelo; Rocha, Claudio Jose da; Mestnik Filho, Jose; Leal Neto, Ricardo Mendes, E-mail: mdurazzo@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. That's why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-de hydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and gamma-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed. (author)

  4. Progress in developing very-high-density low-enriched-uranium fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Meyer, M.K.; Hayes, S.L.; Wiencek, T.C.; Strain, R.V.

    1999-01-01

    Preliminary results from the postirradiation examinations of microplates irradiated in the RERTR-1 and -2 experiments in the ATR have shown several binary and ternary U-Mo alloys to be promising candidates for use in aluminum-based dispersion fuels with uranium densities up to 8 to 9 g/cm 3 . Ternary alloys of uranium, niobium, and zirconium performed poorly, however, both in terms of fuel/matrix reaction and fission-gas-bubble behavior, and have been dropped from further study. Since irradiation temperatures achieved in the present experiments (approximately 70 deg. C) are considerably lower than might be experienced in a high-performance reactor, a new experiment is being planned with beginning-of-cycle temperatures greater than 200 deg. C in 8-g U/cm 3 fuel. (author)

  5. Transient behaviour study program of research reactors fuel elements at the Hydra Pulse Reactor

    International Nuclear Information System (INIS)

    Khvostionov, V.E.; Egorenkov, P.M.; Malankin, P.V.

    2004-01-01

    Program on behavior study of research reactor Fuel Elements (FE) under transient regimes initiated by excessive reactivity insertion is being presented. Program would be realized at HYDRA pulse reactor at Russian Research Center 'Kurchatov Institute' (RRC 'K1'). HYDRA uses aqueous solution of uranyl sulfate (UO 2 SO 4 ) as a fuel. Up to 30 MJ of energy can be released inside the core during the single pulse, effective power pulse width varying from 2 to 10 ms. Reactor facility allows to investigate behaviour of FE consisting of different types of fuel composition, being developed according to Russian RERTR. First part of program is aimed at transient behaviour studying of FE MR, IRT-3M, WWR-M5 types containing meats based on dioxide uranium in aluminum matrix. Mentioned FEs use 90% and 36% enriched uranium. (author)

  6. Designing the User Interface COBRET under Windows to Carry out Pre- and Post-Processing for the Programs COBRA-RERTR and PARET

    International Nuclear Information System (INIS)

    Ghazi, N.; Monther, A.; Hainoun, A.

    2004-01-01

    In the frame work of testing, evaluation and application of computer codes in the design and safety analysis of research reactors, the dynamic code PARET and the thermal hydraulic code COBRA-RERTR have been adopted. In order to run the codes under windows and to support the user by pre- and post processing, the user interface program COBRET has been developed in the programming language Visual Basic 6 and the data used by it are organized and stored in a relational database in MS Access, an integral part of the software package, MS Office. The interface works in the environment of the Windows operating system and utilizes its graphics as well as other possibilities. It consists of Pre and Post processor. The pre processor deals with the interactive preparation of the input files for PARET and COBRA codes. It supports the user with an automatic check in routine for detecting logical input errors in addition to many direct helps during the multi mode input process. This process includes an automatic branching according to the selected control parameters that depends on the simulation modes of the considered physical problem. The post processor supports the user with graphical tool to present the time and axial distribution of the system variables that consist of many neutronics and thermal hydraulic parameters of the reactor system like neutron flux, reactivity, temperatures, flow rate, pressure and void distribution. (authors)

  7. Designing the user interface COBRET under windows to carry out pre- and post-processing for the programs COBRA-RERTR and PARET

    International Nuclear Information System (INIS)

    Hainoun, A.; Monther, A.; Ghazi, N.

    2004-01-01

    In this framework of testing, evaluation and application of computer codes in the design studies and safety analysis of research reactors, the dynamic code PARET and the thermal hydraulic code COBRA-RERTR have been adopted. In order to run the codes under windows and to support the user by pre- and post processing. The user interface program COBRET has been developed in the programming language visual basic 6 and the data used by it are organized and stored in a relational database in MS Access, and integral port of the software package, MS Office. The interface works in the environment of the Windows operating system and utilizes its graphics as well as other possibilities. It consists of Pre and Post processor. The pre processor deals with the interactive preparation of the input files for PARET and COBRA codes. it supports the user with an automatic check in routine for detecting logical input errors in addition to many direct helps during the multi mode input process. This process includes an automatic branching according to the selected control parameters that depends on the simulation modes of the considered physical problem. The post processor supports the user with graphical tool to present the time and axial distribution of the system variables that consist of many neutronics and thermal hydraulic parameters o the reactor system like neutron flux, reactivity, temperatures, flow rate, pressure and void distribution (author)

  8. Microstructure of the irradiated U 3Si 2/Al silicide dispersion fuel

    Science.gov (United States)

    Gan, J.; Keiser, D. D.; Miller, B. D.; Jue, J.-F.; Robinson, A. B.; Madden, J. W.; Medvedev, P. G.; Wachs, D. M.

    2011-12-01

    The silicide dispersion fuel of U 3Si 2/Al is recognized as the best performance fuel for many nuclear research and test reactors with up to 4.8 gU/cm 3 fuel loading. An irradiated U 3Si 2/Al dispersion fuel ( 235U ˜ 75%) from the high-flux side of a fuel plate (U0R040) from the Reduced Enrichment for Research and Test Reactors (RERTR)-8 test was characterized using transmission electron microscopy (TEM). The fuel was irradiated in the Advanced Test Reactor (ATR) for 105 days. The average irradiation temperature and fission density of the U 3Si 2 fuel particles for the TEM sample are estimated to be approximately 110 °C and 5.4 × 10 27 f/m 3. The characterization was performed using a 200-kV TEM. The U/Si ratio for the fuel particle and (Si + Al)/U for the fuel-matrix-interaction layer are approximately 1.1 and 4-10, respectively. The estimated average diameter, number density and volume fraction for small bubbles (<1 μm) in the fuel particle are ˜94 nm, 1.05 × 10 20 m -3 and ˜11%, respectively. The results and their implication on the performance of the U 3Si 2/Al silicide dispersion fuel are discussed.

  9. Development of a treatment technology for diluting highly enriched AL-based DOE spent nuclear fuel: principles and practices

    International Nuclear Information System (INIS)

    Adams, T.M.; Duncan, A.J.; Peacock, H.B.; Fisher, D.L.

    2001-01-01

    The Savannah River Site is the U.S. Department of Energy's preferred site for return and treatment of all aluminum-base, spent, research and test reactor fuel assemblies. There are over 20,000 spent fuel assemblies now stored in different countries around the world, and by 2035 many will be returned to SRS for treatment and interim storage. Interim storage canisters at SRS will be sent to a Mined Geologic Repository for long-term geologic storage. Laboratory and plant layout of a full-scale process for irradiated fuel has been completed. Tests of the off gas absorption system have been initiated using both surrogate and irradiated RERTR mini fuel plates. An pilot-scale L-Area Experimental Facility (LEF) is planned to validate induction furnace operations, remote handling, and the off gas system for trapping volatile elements under plant operating conditions. (authors)

  10. The first critical experiment with a new type of fuel assemblies IRT-3M on the training reactor VR-I

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    1997-01-01

    The paper 'The first critical experiment with a new type of fuel assemblies IRT-3M on training reactor VR-1 presents basic information about the replacement of fuel on the reactor VR-1 run on FJFI CVUT in Prague. In spring 1997 the IRT-2M fuel type used till then was replaced by the IRT-3M type. When the fuel was replaced, no change in its enrichment was made, i.e. its level remained as 36% 235 U. The replacement itself was carried out in tight co-operation with the Nuclear Research Institute Rez plc., as related to the operation of the research reactor LVR-15. The fuel replacement on the VR-I reactor is a part of the international program RERTR (Reduced Enrichment for Research and Test Reactors) in which the Czech Republic participates. (author)

  11. An investigation on fuel meats extruded with atomized U-10wt% Mo powder for uranium high-density dispersion fuel

    International Nuclear Information System (INIS)

    Kim, Chang-Kyu; Kim, Ki-Hwan; Park, Jong-Man; Lee, Don-Bae; Sohn, Dong-Seong

    1997-01-01

    The RERTR program has been making an effort to develop dispersion fuels with uranium densities of 8 to 9 g U/cm3 for research and test reactors. Using atomized U-10wt%Mo powder, fuel meats have been fabricated successfully up to 55 volume % of fuel powder. The uranium density of an extruded meat with a 55 volume % of fuel powder was obtained to be 7.7 g/cm3. A relatively high porosity of 7.3% was formed due to cracking of particles, presumably induced by the impingement among agglomerated particles. Tensile test results indicated that the strength of fuel meats with 55% volume fraction decreased some and a little of ductility was maintained. Examination on the fracture surface revealed that some U-10%Mo particles appeared to be broken by the tensile force in brittle rupture mode. The increase of broken particles in high fuel fraction is considered to be induced mainly by the impingement among agglomerated particles. Uranium loading density is assumed to be improved through the development of the better homogeneous dispersion technology. (author)

  12. Reclamation and reuse of LEU silicide fuel from manufacturing scrap

    International Nuclear Information System (INIS)

    Gale, G.R.; Pace, B.W.; Evans, R.S.

    2004-01-01

    In order to provide an understanding of the organization which is the sole supplier of United States plate type research and test reactor fuel and LEU core conversions, a brief description of the structure and history is presented. Babcock and Wilcox (B and W) is a part of McDermott International, Inc. which is a large diversified corporation employing over 20,000 people primarily in engineering and construction for the off-shore oil and power generation industries throughout the world. B and W provides many energy related products requiring precision machining and high quality systems. This is accomplished by using state-of-the-art equipment, technology and highly skilled people. The RTRFE group within B and W has the ability to produce various complexly shaped fuel elements with a wide variety of fuels and enrichments. B and W RTRFE has fabricated over 200,000 plates since 1981 and gained the diversified experience necessary to satisfy many customer requirements. This accomplishment was possible with the support of McDermott International and all of its resources. B and W has always had a commitment to high quality and integrity. This is apparent by the success and longevity (125 years) of the company. A lower cost to convert cores to LEU provides direct support to RERTR and demonstrates Babcock and Wilcox's commitment to the program. As a supporter of RERTR reactor conversion from HEU to LEU, B and W has contributed a significant amount of R and D money to improve the silicide fuel process which ultimately lowers the LEU core costs. In the most recent R and D project, B and W is constructing a LEU silicide reclamation facility to re-use the unirradiated fuel scrap generated from the production process. Remanufacturing use of this fuel completes the fuel cycle and provides a contribution to LEU cores by reducing scrap inventory and handling costs, lowering initial purchase of fuel due to increasing the process yields, and lowering the replacement costs. This

  13. U.S. progress in the development of very high density low enrichment research reactor fuels

    International Nuclear Information System (INIS)

    Meyer, M. K.; Wachs, D. M.; Jue, J.-F.; Keiser, D. D.; Gan, J.; Rice, F.; Robinson, A.; Woolstenhulme, N. E.; Medvedev, P.; Hofman, G. L.; Kim, Y.-S.

    2012-01-01

    The effort to develop low-enriched fuels for high power research reactors began world-wide in 1996. Since that time, hundreds of fuel specimens have been tested to investigate the operational limits of many variations of U-Mo alloy dispersion and monolithic fuels. In the U.S., the fuel development program has focused on the development of monolithic fuel, and is currently transitioning from conducting research experiments to the demonstration of large scale, prototypic element assemblies. These larger scale, integral fuel performance demonstrations include the AFIP-7 test of full-sized, curved plates configured as an element, the RERTR-FE irradiation of hybrid fuel elements in the Advanced Test Reactor, reactor specific Design Demonstration Experiments, and a multi-element Base Fuel Demonstration. These tests are conducted alongside mini-plate tests designed to prove fuel stability over a wide range of operating conditions. Along with irradiation testing, work on collecting data on fuel plate mechanical integrity, thermal conductivity, fission product release, and microstructural stability is underway. (authors)

  14. High U-density nuclear fuel development with application of centrifugal atomization technology

    International Nuclear Information System (INIS)

    Kim, Chang Kyu; Kim, Ki Hwan; Lee, Don Bae

    1997-01-01

    In order to simplify the preparation process and improve the properties of uranium silicide fuels prepared by mechanical comminution, a fuel fabrication process applying rotating-disk centrifugal atomization technology was invented in KAERI in 1989. The major characteristic of atomized U 3 Si and U 3 Si 2 powders have been examined. The out-pile properties, including the thermal compatibility between atomized particle and aluminum matrix in uranium silicide dispersion fuels, have generally showed a superiority to the comminuted fuels. Moreover, the RERTR (reduced enrichment for research and test reactors) program, which recently begins to develop very-high-density uranium alloy fuels, including U-Mo fuels, requires the centrifugal atomization process to overcome the contaminations of impurities and the difficulties of the comminution process. In addition, a cooperation with ANL in the U.S. has been performed to develop high-density fuels with an application of atomization technology since December 1996. If the microplate and miniplate irradiation tests of atomized fuels, which have been performed with ANL, demonstrated the stability and improvement of in-reactor behaviors, nuclear fuel fabrication technology by centrifugal atomization could be most-promising to the production method of very-high-uranium-loading fuels. (author). 22 refs., 2 tabs., 12 figs

  15. The Microstructure of Multi-wire U-Mo Monolithic Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Sang; Park, Eun Kee; Cho, Woo Hyoung; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    In order to use low-enriched uranium (LEU) instead of highly enriched uranium (HEU) for high performance research reactors, the reduced enrichment for research and test reactors (RERTR) program is developing high uranium density fuel such as U-Mo/Al dispersion fuel. U-Mo alloys have an excellent irradiation performance when compared to other uranium alloys or compounds. But the results from the post-irradiation examination of the U-Mo/Al dispersion fuels indicate that an interaction between the U-Mo alloy fuel and the Al matrix phases occurs readily during an irradiation and it is sensitively dependent on the temperature. In order to lessen these severe interactions, a concept of a multi-wire type fuel was proposed. The fuel configuration is that three to six U-Mo fuel wires (1.5 mm {approx} 2 mm in diameter) are symmetrically arranged at the periphery side in the Al matrix as shown. This multi-wire fuels showed very good fuel performance during the KOMO-3 irradiation test. At the KOMO-3 test, the specimen of the multi-wire fuels were U-7Mo/Al and U-7Mo-1Si/Al. In this study we investigate the microstructure change of the U-7Mo and U-7Mo-1Ti with some variation of annealing conditions. In addition to this, we want to check the effect of adding Ti element to U-7Mo on the gamma phase stability

  16. Modeling of high-density U-MO dispersion fuel plate performance

    International Nuclear Information System (INIS)

    Hayes, S.L.; Meyer, M.K.; Hofman, G.L.; Rest, J.; Snelgrove, J.L.

    2002-01-01

    Results from postirradiation examinations (PIE) of highly loaded U-Mo/Al dispersion fuel plates over the past several years have shown that the interaction between the metallic fuel particles and the matrix aluminum can be extensive, reducing the volume of the high-conductivity matrix phase and producing a significant volume of low-conductivity reaction-product phase. This phenomenon results in a significant decrease in fuel meat thermal conductivity during irradiation. PIE has further shown that the fuel-matrix interaction rate is a sensitive function of irradiation temperature. The interplay between fuel temperature and fuel-matrix interaction makes the development of a simple empirical correlation between the two difficult. For this reason a comprehensive thermal model has been developed to calculate temperatures throughout the fuel plate over its lifetime, taking into account the changing volume fractions of fuel, matrix and reaction-product phases within the fuel meat owing to fuel-matrix interaction; this thermal model has been incorporated into the dispersion fuel performance code designated PLATE. Other phenomena important to fuel thermal performance that are also treated in PLATE include: gas generation and swelling in the fuel and reaction-product phases, incorporation of matrix aluminum into solid solution with the unreacted metallic fuel particles, matrix extrusion resulting from fuel swelling, and cladding corrosion. The phenomena modeled also make possible a prediction of fuel plate swelling. This paper presents a description of the models and empirical correlations employed within PLATE as well as validation of code predictions against fuel performance data for U-Mo experimental fuel plates from the RERTR-3 irradiation test. (author)

  17. Current and prospective fuel test programmes in the MIR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

    2007-07-01

    MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

  18. Development of technology of high density LEU dispersion fuel fabrication

    International Nuclear Information System (INIS)

    Wiencek, T.; Totev, T.

    2007-01-01

    Advanced Materials Fabrication Facilities at Argonne National Laboratory have been involved in development of LEU dispersion fuel for research and test reactors from the beginning of RERTR program. This paper presents development of technology of high density LEU dispersion fuel fabrication for full size plate type fuel elements. A brief description of Advanced Materials Fabrication Facilities where development of the technology was carried out is given. A flow diagram of the manufacturing process is presented. U-Mo powder was manufactured by the rotating electrode process. The atomization produced a U-Mo alloy powder with a relatively uniform size distribution and a nearly spherical shape. Test plates were fabricated using tungsten and depleted U-7 wt.% Mo alloy, 4043 Al and Al-2 wt% Si matrices with Al 6061 aluminum alloy for the cladding. During the development of the technology of manufacturing of full size high density LEU dispersion fuel plates special attention was paid to meet the required homogeneity, bonding, dimensions, fuel out of zone and other mechanical characteristics of the plates.

  19. Swelling Estimation of Multi-wire U-Mo Monolithic Fuel for HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon-Sang; Ryu, Ho-Jin; Park, Jong-Man; Oh, Jong-Myeong; Kim, Chang-Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    In order to use low-enriched uranium (LEU) instead of highly enriched uranium (HEU) for high performance research reactors, the reduced enrichment for research and test reactors (RERTR) program is developing high uranium density fuel such as U-Mo/Al dispersion fuel. U-Mo alloys have an excellent irradiation performance when compared to other uranium alloys or compounds. But the results from the post-irradiation examination of the U-Mo/Al dispersion fuels indicate that an interaction between the U-Mo alloy fuel and the Al matrix phases occurs readily during an irradiation and it is sensitively dependent on the temperature. In order to lessen these severe interactions, a concept of a multi-wire type fuel was proposed. The fuel configuration is that three to six U-Mo fuel wires (1.5 mm - 2 mm in diameter) are symmetrically arranged at the periphery side in the Al matrix. In this study temperature calculations and a swelling estimation of a multi-wire monolithic fuel were carried out. Also the results of a post irradiation analysis of this fuel will be introduced.

  20. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Science.gov (United States)

    Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

    2008-08-01

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  1. Oxidation of aluminum alloy cladding for research and test reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2008-08-31

    The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

  2. The Ford Nuclear Reactor demonstration project for the evaluation and analysis of low enrichment fuel

    International Nuclear Information System (INIS)

    Kerr, W.; King, J.S.; Lee, J.C.; Martin, W.R.; Wehe, D.K.

    1991-07-01

    The whole-core LEU fuel demonstration project at the University of Michigan was begun in 1979 as part of the Reduced Enrichment Research and Test Reactor (RERTR) Program at Argonne National Laboratory. An LEU fuel design was selected which would produce minimum perturbations in the neutronic, operations, and safety characteristics of the 2-MW Ford Nuclear Reactor (FNR). Initial criticality with a full LEU core on December 8, 1981, was followed by low- and full-power testing of the fresh LEU core, transitional operation with mixed HEU-LEU configurations, and establishment of full LEU equilibrium core operation. The transition from the HEU to the LEU configurations was achieved with negligible impact on experimental utilization and safe operation of the reactor. 78 refs., 74 figs., 84 tabs

  3. Conversion of Reactor LVR-15 in Czech Republic from HEU to LEU Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Broz, V.; Miletic, M.; Koleska, M.; Ernest, J.; Vins, M. [Research Reactors, Research Centre Rez Ltd., Husinec-Rez 130, CZ 250 68 (Czech Republic)

    2011-07-01

    Accordingly to the IAEA recommendations and RERTR program, the LVR-15 reactor started the process of conversion from fuel enriched to 36 % to fuel enriched up to 20 % U{sup 235}. As the most suitable fuel for the reactor was chosen the IRT-4M fuel enriched to 19.7% U{sup 235}, fabricated in NZCHK Novosibirsk. The most important requirements, the fuel had to fulfill, were attainability, constructional continuity with the old type of FAs and operational experiences. The conversion procedure began in January 2010 with testing irradiation of 3 IRT-4M FAs. Test irradiation took 9 reactor operation cycles. During this period were done visual inspection and sipping tests of FAs. An experiment with the aim to compare the influence of the fuel changing to neutron flux and the reactivity and to verify basic physical characteristics of the new fuel was performed at the start of test irradiation. The conversion itself is going on since January 2011. In this time, every cycle will be replaced 1 or 2 burned-up IRT-2M FAs with fresh IRT-4M FAs. This period will take 14 cycles and at the end should be in the core used only IRT-4M fuel. (author)

  4. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

  5. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    International Nuclear Information System (INIS)

    Travelli, A.

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  6. Proceedings of the international meeting on research and test reactor core conversions from HEU to LEU fuels

    Energy Technology Data Exchange (ETDEWEB)

    Travelli, A [Argonne National Laboratory, Argonne, IL (United States)

    1983-09-01

    Conversion of research and test reactor cores from the use of high enrichment uranium to the use of low enrichment uranium depends on the cooperation of many research organizations, reactor operators, and government agencies. At a technical level, it involves almost all aspects of the fuel cycle, including fuel development, testing, shipping and reprocessing; experiment performance; economics; and safety and licensing aspects. The reactors involved and the conversion activities are distributed among approximately 25 countries, making this a subject which is best dealt with on an international basis. To foster direct communication in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the fifth of a series which began in 1978. The previous meetings were held at Argonne (International Meeting of Research Reactor Fuel Designers, Developers, and Fabricators, Argonne National Laboratory, Argonne, U.S.A., November 910, 1978), at Saclay (IAEA Consultants' Meeting on Research Reactor Core Conversions from HEU to LEU, Centre d'etudes Nucleaires de Saclay, Saclay, France, December 12-14, 1979), at Argonne (International Meeting on Development, Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors, Argonne National Laboratory, Argonne, U.S.A., November 12-14, 1980) and at Juelich (Seminar on Research Reactor Operation and Use, Juelich Nuclear Research Center, Juelich, F.R.G., September 48, 1981). Proceedings from the two most recent previous meetings were published as ANL/RERTR/TM-3 (CONF-801144) and IAEA-SR-77. The spirit of this meeting differs slightly from that of the previous meetings. The advances which have been made and the growing maturity of the effort have caused a gradual shift of emphasis away from those topics which dominated the floor during the first meetings, such as fuel and methods development, and towards topics which concern more

  7. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    Science.gov (United States)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (RERTR-9B experiment. This paper discusses the TEM characterization results for this U-10Mo/Zr/Al6061 monolithic fuel plate (∼59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 °C, respectively. TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (>1 μm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ∼30 at% and ∼7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  8. Progress in qualifying low-enriched U-Mo dispersion fuels

    International Nuclear Information System (INIS)

    Snelgrove, J.L.; Hofman, G.L.; Hayes, S.L.; Meyer, M.K.

    2001-01-01

    The U.S. Reduced Enrichment for Research and Test Reactors program is working to qualify dispersions of U-Mo alloys in aluminum with fuel-meat densities of 8 to 9 gU cm -3 . Post irradiation examinations of the small fuel plates irradiated in the Advanced Test Reactor during the high-temperature RERTR-3 tests are virtually complete, and analysis of the large quantity of data obtained is underway. We have observed that the swelling of the fuel plates is stable and modest and that the swelling is dominated by the temperature-dependent interaction of the U-Mo fuel and the aluminum matrix. In order to extract detailed information about the behavior of these fuels from the data, a complex fuel-plate thermal model is being developed to account for the effects of the changing fission rate and thermal conductivity of the fuel meat during irradiation. This paper summarizes the empirical results of the post irradiation examinations and the preliminary results of the model development. In addition, the schedule for irradiation of full-sized elements in the HFR-Petten is briefly discussed. (author)

  9. Results of Microstructural Examinations of Irradiated LEU U-Mo Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, ID 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organization (Australia)

    2009-06-15

    Introduction: The RERTR program is responsible for converting research reactors that use high-enriched uranium fuels to ones that use low-enriched uranium fuels [1]. As part of the development of LEU fuels, a variety of irradiation experiments are being conducted using the Advanced Test Reactor. Based on the results of initial fuel plate testing, adjustments have been made to the characteristics of fuel plates to improve the stability of the fuel microstructure. One improvement has been to add Si to the matrix of a dispersion fuel. This material is also being added at the fuel/cladding interface of a monolithic fuel. This paper will discuss the irradiation performance of these fuels, in terms of the stability of their microstructures during irradiation. Results and discussion: The post-irradiation examinations of fuel plates are performed at the Idaho National Laboratory. These examinations consist of visual examinations of fuel plates, gamma scanning, thickness measurements, oxide thickness measurements, and optical metallographic examinations of the fuel plate microstructures. Microstructural analysis is also performed using scanning electron microscopy. Overall, U-7Mo and U-10Mo alloy fuels have displayed the best irradiation performance, particularly, when a Si-containing Al alloy is used as the dispersion fuel matrix. The benefit of using this type of matrix is that the commonly observed fuel/cladding interaction that occurs during irradiation is reduced and the interaction layer that forms exhibit stable behavior during irradiation. Monolithic-type fuels, which consist of a U-Mo foil encased in Al alloy cladding, are also being developed. These types of fuels are also showing promise and will continue to be developed. One challenge with this type of fuel is in trying to maximize the bond strength at the foil/cladding interface. Fuel/cladding interactions can affect the quality of the boding at this interface. Si is being added to improve the characteristics

  10. Reactivity feedback coefficients of a material test research reactor fueled with high-density U{sub 3}Si{sub 2} dispersion fuels

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

    2008-10-15

    The reactivity feedback coefficients of a material test research reactor fueled with high-density U{sub 3}Si{sub 2} dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U{sub 3}Si{sub 2} LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 deg. C to 100 deg. C, at the beginning of life, followed the relationships (in units of {delta}k/k x 10{sup -5} K{sup -1}) -2.116 - 0.118 {rho}{sub U}, 0.713 - 37.309/{rho}{sub U} and -12.765 - 34.309/{rho}{sub U}, respectively for 4.0 {<=} {rho}{sub U} (g/cm{sup 3}) {<=} 6.0.

  11. Fuel Exhaling Fuel Cell.

    Science.gov (United States)

    Manzoor Bhat, Zahid; Thimmappa, Ravikumar; Devendrachari, Mruthyunjayachari Chattanahalli; Kottaichamy, Alagar Raja; Shafi, Shahid Pottachola; Varhade, Swapnil; Gautam, Manu; Thotiyl, Musthafa Ottakam

    2018-01-18

    State-of-the-art proton exchange membrane fuel cells (PEMFCs) anodically inhale H 2 fuel and cathodically expel water molecules. We show an unprecedented fuel cell concept exhibiting cathodic fuel exhalation capability of anodically inhaled fuel, driven by the neutralization energy on decoupling the direct acid-base chemistry. The fuel exhaling fuel cell delivered a peak power density of 70 mW/cm 2 at a peak current density of 160 mA/cm 2 with a cathodic H 2 output of ∼80 mL in 1 h. We illustrate that the energy benefits from the same fuel stream can at least be doubled by directing it through proposed neutralization electrochemical cell prior to PEMFC in a tandem configuration.

  12. Progress in the development of very high density research and test reactor fuels

    Energy Technology Data Exchange (ETDEWEB)

    Wachs, D.M. [Idaho National Laboratory, P.O. Box 2528, Idaho Falls, Idaho 83415 (United States)

    2009-06-15

    New nuclear fuels are being developed to enable many of the most important research and test reactors worldwide to convert from high enriched uranium (HEU) fuels to low enriched uranium (LEU) fuels without significant loss in performance. The last decade of work has focused on the development of uranium-molybdenum alloy (U-Mo) based fuels and is an international effort that includes the active participation of more than ten national programs. The US RERTR program, under the NNSA's Global Threat Reduction Initiative (GTRI), is in the process of developing both dispersion and monolithic U-Mo fuel designs. While the U-Mo fuel alloy has behaved extremely well under irradiation, initial testing (circa 2003) revealed that the U-Mo fuels dispersed in aluminum had an unexpected tendency toward unstable swelling (pillowing) under high-power conditions. Technical investigations were initiated worldwide at this time by the partner programs to understand this behavior as well as to develop and test remedies. The behavior was corrected by modifying the chemistry of the U-Mo/Al interfaces in both fuel designs. In the dispersion fuel design, this was accomplished by the addition of small amounts of silicon to the aluminum matrix material. Two methods are under development for the monolithic fuel design, which include the application of a thin layer of silicon or a thin zirconium based diffusion barrier at the fuel/clad interface. This paper gives an overview of the current status of U-Mo fuel development, including basic research results, manufacturing aspects, results of the latest irradiations and post irradiation examinations, the approach to fuel performance qualification, and the scale-up and commercialization of fabrication technology. (authors)

  13. Milling uranium silicide powder for dispersion nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Vieira, E.; Silva, D.G.; Souza, J.A.B.; Durazzo, M. [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Riella, H.G. [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil)

    2009-07-01

    Full text: Uranium silicide (U3Si2) is presently considered the best fuel qualified so far in terms of uranium loading and performance. Stability of the U3Si2 fuel with uranium density of 4.8 g/cm3 was confirmed by burnup stability tests performed during the Reduced Enrichment for Research and Test Reactors (RERTR) program. This fuel was chosen to compose the first core of the new Brazilian Multipurpose Research Reactor (RMB), planned to be constructed in the next years. This new reactor will consume bigger quantities of U3Si2 powder, when compared with the small consumption of the IEA-R1 research reactor of IPEN-CNEN/SP, the unique MTR type research reactor operating in the country. At the present time, the milling operation of U3Si2 ingots is made manually. In order to increase the powder production capacity, the manual milling must be replaced by an automated procedure. This paper describes a new milling machine and procedure developed to produce U3Si2 powder with higher efficiency. (author)

  14. Analyses for inserting fresh LEU fuel assemblies instead of fresh HEU fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam

    International Nuclear Information System (INIS)

    Hanan, N. A.; Deen, J.R.; Matos, J.E.

    2005-01-01

    Analyses were performed by the RERTR Program to replace 36 burned HEU (36%) fuel assemblies in the Dalat Nuclear Research Reactor in Vietnam with either 36 fresh fuel assemblies currently on-hand at the reactor or with LEU fuel assemblies to be procured. The study concludes that the current HEU (36%) WWR-M2 fuel assemblies can be replaced with LEU WWR-M2 fuel assemblies that are fully-qualified and have been commercially available since 2001 from the Novosibirsk Chemical Concentrates Plant in Russia. The current reactor configuration using re-shuffled HEU fuel began in June 2004 and is expected to allow normal operation until around August 2006. If 36 HEU assemblies each with 40.2 g 235 U are inserted without fuel shuffling over the next five operating cycles, the core could operate for an additional 10 years until June 2016. Alternatively, inserting 36 LEU fuel assemblies each containing 49.7 g 235 U without fuel shuffling over five operating cycles would allow normal operation for about 14 years from August 2006 until October 2020. The main reason for the longer service life of the LEU fuel is that its 235 U content is higher than the 235 U content needed simply to match the service life of the HEU fuel. Fast neutron fluxes in the experiment regions would be very nearly the same in both the HEU and LEU cores. Thermal neutron fluxes in the experiment regions would be lower by 1-5%, depending on the experiment type and location. (author)

  15. LEU fuel development at CERCA. Status as of October 1997. Preliminary developments of MTR plates with UMo fuel

    International Nuclear Information System (INIS)

    Durand, J.P.; Lavastre, Y.; Grasse, M.

    1997-01-01

    UMo fuels are considered by the RERTR programme because of their higher density as compared to U 3 Si 2 . This paper is focused on the preliminary results about the manufacture feasibility of Uranium/Molybdenum fuel plates carried out by CERCA. A special procedure of casting and heat treatment has been developed in order to get an homogeneous gamma phase of UMo alloy Although U-5%Mo allows to reach densities up to 9.9 U/cm3 with the advanced process developed by CERCA for the high loaded plates, it is not a good candidate on the thermal stability point of view. U-9%Mo alloy seems to gather all the criteria for a good fuel alloy but it is a little less effective on the Uranium density point of view as compared to U-5%Mo alloy. In any case, the preliminary feasibility results are very much encouraging because UMo alloys seem to be compatible with the Aluminium matrix when taking special care while manufacturing. A good compromise could be an intermediate percentage of Molybdenum or the addition of metal traces in order to thermally stabilise 5%Mo. (author)

  16. Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

    Science.gov (United States)

    Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

    2018-04-01

    A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature- and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS). The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code.

  17. Development of PRIME for irradiation performance analysis of U-Mo/Al dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Gwan Yoon; Kim, Yeon Soo; Jeong, Yong Jin; Park, Jong Man; Sohn, Dong-Seong

    2018-04-01

    A prediction code for the thermo-mechanical performance of research reactor fuel (PRIME) has been developed with the implementation of developed models to analyze the irradiation behavior of U-Mo dispersion fuel. The code is capable of predicting the two-dimensional thermal and mechanical performance of U-Mo dispersion fuel during irradiation. A finite element method was employed to solve the governing equations for thermal and mechanical equilibria. Temperature-and burnup-dependent material properties of the fuel meat constituents and cladding were used. The numerical solution schemes in PRIME were verified by benchmarking solutions obtained using a commercial finite element analysis program (ABAQUS).The code was validated using irradiation data from RERTR, HAMP-1, and E-FUTURE tests. The measured irradiation data used in the validation were IL thickness, volume fractions of fuel meat constituents for the thermal analysis, and profiles of the plate thickness changes and fuel meat swelling for the mechanical analysis. The prediction results were in good agreement with the measurement data for both thermal and mechanical analyses, confirming the validity of the code. (c) 2018 Elsevier B.V. All rights reserved.

  18. Characterization of the Microstructure of Irradiated U-Mo Dispersion Fuel with a Matrix that Contains Si

    International Nuclear Information System (INIS)

    Keiser, Jr. D.D.; Robinson, A.B.; Jue, J.F.; Medvedev, P.; Finlay, M.R.

    2009-01-01

    RERTR U-Mo dispersion fuel plates are being developed for application in research reactors throughout the world. Of particular interest is the irradiation performance of U-Mo dispersion fuels with Si added to the Al matrix. Si is added to improve the performance of U-Mo dispersion fuels. Microstructural examinations have been performed on fuel plates with Al-2Si matrix after irradiation to around 50% LEU burnup. Si-rich layers were observed in many areas around the various U-7Mo fuel particles. In one local area of one of the samples, where the Si-rich layer had developed into a layer devoid of Si, relatively large fission gas bubbles were observed in the interaction phase. There may be a connection between the growth of these bubbles and the amount of Si present in the interaction layer. Overall, it was found that having Si-rich layers around the fuel particles after fuel plate fabrication positively impacted the overall performance of the fuel plate

  19. Mechanical Calculations on U-Mo Dispersion fuel plates with MAIA

    International Nuclear Information System (INIS)

    Marelle, V.; Huet, F.; Lemoine, P.

    2005-01-01

    CEA has developed a 2D thermo-mechanical code, called MAIA, for modelling the behaviour of U-Mo dispersion fuel. MAIA uses a finite element method for the resolution of the thermal and mechanical problems. Physical models, issued of the DOE-ANL code PLATE, evaluate the fission products swelling and the volume fraction of the interaction between U-Mo and Al. They allow establishing strains in the meat imposed as loading for the mechanical calculation. MAIA has been validated on the irradiations IRIS 1 and RERTR-3 and a rather good agreement is obtained with post irradiation examinations. MAIA is used to calculate the last irradiation of the French UMo group, IRIS 2. MAIA predicts a maximum temperature of 112 deg. C and meat swelling of 16%. Mechanical calculations are finally performed to evaluate the sensitivity to some mechanical hypotheses such as constitutive laws and the way the meat swelling is applied. (author)

  20. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-12-15

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60%{sup 235}U; the mini-rods were irradiated to an average burnup of ∼ 85%{sup 235}U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%.

  1. Comparative Analysis of Structural Changes In U-Mo Dispersed Fuel of Full-Size Fuel Elements And Mini-Rods Irradiated In The MIR Reactor

    International Nuclear Information System (INIS)

    Izhutov, Aleksey L.; Iakovlev, Valeriy V.; Novoselov, Andrey E. and others

    2013-01-01

    The paper summarizes the irradiation test and post-irradiation examination (PIE) data for the U-Mo low-enriched fuel that was irradiated in the MIR reactor under the RERTR Program. The PIE data were analyzed for both full-size fuel rods and mini-rods with atomized powder dispersed in Al matrix as well as with additions of 2%, 5% and 13% of silicon in the matrix and ZrN protective coating on the fuel particles. The full-size fuel rods were irradiated up to an average burnup of ∼ 60% 235 U; the mini-rods were irradiated to an average burnup of ∼ 85% 235 U. The presented data show a significant increase of the void fraction in the U-Mo alloy as the U-235 burnup rises from ∼ 40% up to ∼ 85%. The effect of irradiation test conditions and U-235 burnup were analyzed with regard to the formation of an interaction layer between the matrix and fuel particles as well as generation of porosity in the U-Mo alloy. Shown here are changes in distribution of U fission products as the U-235 burnup increases from ∼ 40% up to ∼ 85%

  2. United States and Russian Cooperation on Issues of Nuclear Nonproliferation

    Science.gov (United States)

    2005-06-01

    Reactors ( RERTR ) This project works with Russia to facilitate conversion of its research and test reactors from highly enriched uranium (HEU) fuel...reactor fuel purchase, accelerated RERTR activities, and accelerated Material Conversion and Consolidation implementation. 89 j. Fissile Materials

  3. KUCA critical experiments using MEU fuel (II)

    International Nuclear Information System (INIS)

    Kanda, Keiji; Hayashi, Masatoshi; Shiroya, Seiji; Kobayashi, Keiji; Fukui, Hiroshi; Mishima, Kaichiro; Shibata, Toshikazu

    1983-01-01

    Due to mutual concerns in the USA and Japan about the proliferation potential of highly-enriched uranium (HEU), a joint study program I was initiated between Argonne National Laboratory (ANL and Kyoto University Research Reactor Institute (KURRI) in 1978. In accordance with the reduced enrichment for research and test reactor (RERTR) program, the alternatives were studied for reducing the enrichment of the fuel to be used in the Kyoto University High Flux Reactor (KUHFR). The KUHFR has a distinct feature in its core configuration it is a coupled-core. Each annular shaped core is light-water-moderated and placed within a heavy water reflector with a certain distance between them. The phase A reports of the joint ANL-KURRI program independently prepared by two laboratories in February 1979, 3,4 concluded that the use of medium-enrichment uranium (MEU, 45%) in the KUHFR is feasible, pending results of the critical experiments in the Kyoto University Critical Assembly (KUCA) 5 and of the burnup test in the Oak Ridge Research Reactor 6 (ORR). An application of safety review (Reactor Installation License) for MEU fuel to be used in the KUCA was submitted to the Japanese Government in March 1980, and a license was issued in August 1980. Subsequently, the application for 'Authorization before Construction' was submitted and was authorized in September 1980. Fabrication of MEU fuel-elements for the KUCA experiments by CERCA in France was started in September 1980, and was completed in March 1981. The critical experiments in the KUCA with MEU fuel were started on a single-core in May 1981 as a first step. The first critical state of the core using MEU fuel was achieved at 312 p.m. in May 12, 1981. After that, the reactivity effects of the outer side-plates containing boron burnable poison were measured. At Munich Meeting in Sept., 1981, we presented a paper on critical mass and reactivity of burnable poison in the MEU core. Since then we carried out the following experiments

  4. KUCA critical experiments using MEU fuel (II)

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Hayashi, Masatoshi; Shiroya, Seiji; Kobayashi, Keiji; Fukui, Hiroshi; Mishima, Kaichiro; Shibata, Toshikazu [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka (Japan)

    1983-09-01

    Due to mutual concerns in the USA and Japan about the proliferation potential of highly-enriched uranium (HEU), a joint study program I was initiated between Argonne National Laboratory (ANL and Kyoto University Research Reactor Institute (KURRI) in 1978. In accordance with the reduced enrichment for research and test reactor (RERTR) program, the alternatives were studied for reducing the enrichment of the fuel to be used in the Kyoto University High Flux Reactor (KUHFR). The KUHFR has a distinct feature in its core configuration it is a coupled-core. Each annular shaped core is light-water-moderated and placed within a heavy water reflector with a certain distance between them. The phase A reports of the joint ANL-KURRI program independently prepared by two laboratories in February 1979, 3,4 concluded that the use of medium-enrichment uranium (MEU, 45%) in the KUHFR is feasible, pending results of the critical experiments in the Kyoto University Critical Assembly (KUCA) 5 and of the burnup test in the Oak Ridge Research Reactor 6 (ORR). An application of safety review (Reactor Installation License) for MEU fuel to be used in the KUCA was submitted to the Japanese Government in March 1980, and a license was issued in August 1980. Subsequently, the application for 'Authorization before Construction' was submitted and was authorized in September 1980. Fabrication of MEU fuel-elements for the KUCA experiments by CERCA in France was started in September 1980, and was completed in March 1981. The critical experiments in the KUCA with MEU fuel were started on a single-core in May 1981 as a first step. The first critical state of the core using MEU fuel was achieved at 312 p.m. in May 12, 1981. After that, the reactivity effects of the outer side-plates containing boron burnable poison were measured. At Munich Meeting in Sept., 1981, we presented a paper on critical mass and reactivity of burnable poison in the MEU core. Since then we carried out the following experiments

  5. Irradiated microstructure of U-10Mo monolithic fuel plate at very high fission density

    Energy Technology Data Exchange (ETDEWEB)

    Gan, J.; Miller, B. D.; Keiser, D. D.; Jue, J. F.; Madden, J. W.; Robinson, A. B.; Ozaltun, H.; Moore, G.; Meyer, M. K.

    2017-08-01

    Monolithic U-10Mo alloy fuel plates with Al-6061 cladding are being developed for use in research and test reactors as low enrichment fuel (< 20% U-235 enrichment) as a result of its high uranium loading capacity compared to that of U-7Mo dispersion fuel. These fuel plates contain a Zr diffusion barrier between the U-10Mo fuel and Al-6061 cladding that suppresses the interaction between the U-Mo fuel foil and Al alloy cladding that is known to be problematic under irradiation. This paper discusses the TEM results of the U-10Mo/Zr/Al6061 monolithic fuel plate (Plate ID: L1P09T, ~ 59% U-235 enrichment) irradiated in Advanced Test Reactor at Idaho National Laboratory as part of RERTR-9B irradiation campaign with an unprecedented high local fission density of 9.8E+21 fissions/cm3. The calculated fuel foil centerline temperature at the beginning of life and the end of life is 141 and 194 C, respectively. A total of 5 TEM lamellas were prepared using focus ion beam lift-out technique. The estimated U-Mo fuel swelling, based on the fuel foil thickness change from SEM, is approximately 76%. Large bubbles (> 1 µm) are distributed evenly in U-Mo and interlink of these bubbles is evident. The average size of subdivided grains at this fission density appears similar to that at 5.2E+21 fissions/cm3. The measured average Mo and Zr content in the fuel matrix is ~ 30 at% and ~ 7 at%, respectively, in general agreement with the calculated Mo and Zr from fission density.

  6. Fuel assemblies

    International Nuclear Information System (INIS)

    Mukai, Hideyuki

    1987-01-01

    Purpose: To prevent bending of fuel rods caused by the difference of irradiation growth between coupling fuel rods and standards fuel rods thereby maintain the fuel rod integrity. Constitution: The f value for a fuel can (the ratio of pole of zirconium crystals in the entire crystals along the axial direction of the fuel can) of a coupling fuel rod secured by upper and lower tie plates is made smaller than the f value for the fuel can of a standard fuel rod not secured by the upper and the lower tie plates. This can make the irradiation growth of the fuel can of the coupling fuel rod greater than the irradiation growth of the fuel can of the standard fuel rod and, accordingly, since the elongation of the standard fuel rod can always by made greater, bending of the standard fuel rod can be prevented. (Yoshihara, M.)

  7. Status of core conversion with LEU silicide fuel in JRR-4

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)

    1997-08-01

    Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10{sup 13}(n/cm{sup 2}/s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities.

  8. Study of diffusion bonding in 6061 aluminum and development of future high-density fuels fabrication

    International Nuclear Information System (INIS)

    Prokofiev, I.G.; Wiencek, T.C.; McGann, D.J.

    1997-01-01

    Powder metallurgy dispersions of uranium alloys and silicides in an aluminum matrix have been developed by the RERTR program as a new generation of proliferation-resistant fuels. Testing uses fuel miniplates to simulate standard fuel with cladding and matrix in plate-type configurations. In order to seal the dispersion fuel plates, a diffusion bond must be established between the aluminum cover plates that surround the fuel meat. Four different variations of the standard method for roll-bonding 6061 aluminum were studied: mechanical cleaning, addition of a getter material, modifications to the standard chemical etching, and modifications to welding. Aluminum test pieces were subjected to a bend test after each rolling pass. Results, based on 400 samples, indicate that a reduction in thickness of at least 70% is required to produce a diffusion bond with the standard roll-bonding method, versus a 60% reduction when using a method in which the assembly was 100% welded and contained empty 9 mm holes near the frame corners. (author)

  9. Status of core conversion with LEU silicide fuel in JRR-4

    International Nuclear Information System (INIS)

    Nakajima, Teruo; Ohnishi, Nobuaki; Shirai, Eiji

    1997-01-01

    Japan Research Reactor No.4 (JRR-4) is a light water moderated and cooled, 93% enriched uranium ETR-type fuel used and swimming pool type reactor with thermal output of 3.5MW. Since the first criticality was achieved on January 28, 1965, JRR-4 has been used for shielding experiments, radioisotope production, neutron activation analyses, training for reactor engineers and so on for about 30 years. Within the framework of the RERTR Program, the works for conversion to LEU fuel are now under way, and neutronic and thermal-hydraulic calculations emphasizing on safety and performance aspects are being carried out. The design and evaluation for the core conversion are based on the Guides for Safety Design and Evaluation of research and testing reactor facilities in Japan. These results show that the JRR-4 will be able to convert to use LEU fuel without any major design change of core and size of fuel element. LEU silicide fuel (19.75%) will be used and maximum neutron flux in irradiation hole would be slightly decreased from present neutron flux value of 7x10 13 (n/cm 2 /s). The conversion works are scheduled to complete in 1998, including with upgrade of the reactor building and utilization facilities

  10. Thermal compatibility of U-2wt.%Mo and U-10wt.%Mo fuel prepared by centrifugal atomization for high density research reactor fuels

    International Nuclear Information System (INIS)

    Kim Ki Hwan; Lee Don Bae; Kim Chang Kyu; Kuk Il Hyun; Hofman, G.E.

    1997-01-01

    Research on the intermetallic compounds of uranium was revived in 1978 with the decision by the international research reactor community to develop proliferation-resistant fuels. The reduction of 93% 235 U (HEU) to 20% 235 U (LEU) necessitates the use of higher U-loading fuels to accommodate the addition 238 U in the LEU fuels. While the vast majority of reactors can be satisfied with U 3 Si 2 -Al dispersion fuel, several high performance reactors require high loadings of up to 8-9 g U cm -3 . Consequently, in the renewed fuel development program of the Reduced Enrichment for Research and Test Reactors (RERTR) Program, attention has shifted to high density uranium alloys. Early irradiation experiments with uranium alloys showed promise of acceptable irradiation behavior, if these alloys can be maintained in their cubic γ-U crystal structure. It has been reported that high density atomized U-Mo powders prepared by rapid cooling have metastable isotropic γ-U phase saturated with molybdenum, and good γ-U phase stability, especially in U-10wt.%Mo alloy fuel. If the alloy has good thermal compatibility with aluminium, and this metastable gamma phase can be maintained during irradiation, U-Mo alloy would be a prime candidate for dispersion fuel for research reactors. In this paper, U-2w.%Mo and U-10w.%Mo alloy powder which have high density (above 15 g-U/cm 3 ), are prepared by centrifugal atomization. The U-Mo alloy fuel meats are made into rods extruding the atomized powders. The characteristics related to the thermal compatibility of U-2w.%Mo and U-10w.%Mo alloy fuel meat at 400 o C for time up to 2000 hours are examined. (author)

  11. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)], E-mail: mfarhan_73@yahoo.co.uk; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, P.O. Nilore, Islamabad 45650 (Pakistan)

    2008-09-15

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease.

  12. Effects of high density dispersion fuel loading on the kinetic parameters of a low enriched uranium fueled material test research reactor

    International Nuclear Information System (INIS)

    Muhammad, Farhan; Majid, Asad

    2008-01-01

    The effects of using high density low enriched uranium on the neutronic parameters of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density LEU fuels currently being developed under the RERTR program. Since the alloying elements have different cross-sections affecting the reactor in different ways, therefore fuels U-Mo (9 w/o) which contain the same elements in same ratio were selected for analysis. Simulations were carried out to calculate core excess reactivity, neutron flux spectrum, prompt neutron generation time, effective delayed neutron fraction and feedback coefficients including Doppler feedback coefficient, and reactivity coefficients for change of water density and temperature. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the excess reactivity at the beginning of life does not increase as the uranium density of fuel. Both the prompt neutron generation time and the effective delayed neutron fraction decrease as the uranium density increases. The absolute value of Doppler feedback coefficient increases while the absolute values of reactivity coefficients for change of water density and temperature decrease

  13. A model for recovery of scrap monolithic uranium molybdenum fuel by electrorefining

    Science.gov (United States)

    Van Kleeck, Melissa A.

    The goal of the Reduced Enrichment for Research and Test Reactors program (RERTR) is toreduce enrichment at research and test reactors, thereby decreasing proliferation risk at these facilities. A new fuel to accomplish this goal is being manufactured experimentally at the Y12 National Security Complex. This new fuel will require its own waste management procedure,namely for the recovery of scrap from its manufacture. The new fuel is a monolithic uraniummolybdenum alloy clad in zirconium. Feasibility tests were conducted in the Planar Electrode Electrorefiner using scrap U-8Mo fuel alloy. These tests proved that a uranium product could be recovered free of molybdenum from this scrap fuel by electrorefining. Tests were also conducted using U-10Mo Zr clad fuel, which confirmed that product could be recovered from a clad version of this scrap fuel at an engineering scale, though analytical results are pending for the behavior of Zr in the electrorefiner. A model was constructed for the simulation of electrorefining the scrap material produced in the manufacture of this fuel. The model was implemented on two platforms, Microsoft Excel and MatLab. Correlations, used in the model, were developed experimentally, describing area specific resistance behavior at each electrode. Experiments validating the model were conducted using scrap of U-10Mo Zr clad fuel in the Planar Electrode Electrorefiner. The results of model simulations on both platforms were compared to experimental results for the same fuel, salt and electrorefiner compositions and dimensions for two trials. In general, the model demonstrated behavior similar to experimental data but additional refinements are needed to improve its accuracy. These refinements consist of a function for surface area at anode and cathode based on charge passed. Several approximations were made in the model concerning areas of electrodes which should be replaced by a more accurate function describing these areas.

  14. Atomistic modeling and simulation of the role of Be and Bi in Al diffusion in U-Mo fuel

    Science.gov (United States)

    Hofman, G. L.; Bozzolo, G.; Mosca, H. O.; Yacout, A. M.

    2011-07-01

    Within the RERTR program, previous experimental and modeling studies identified Si as the alloying addition to the Al cladding responsible for inhibiting Al interdiffusion in the UMo fuel. However, difficulties with reprocessing have rendered this choice inappropriate, leading to the need to study alternative elements. In this work, we discuss the results of an atomistic modeling effort which allows for the systematic study of several possible alloying additions. Based on the behavior observed in the phase diagrams, beryllium or bismuth additions suggest themselves as possible options to replace Si. The results of temperature-dependent simulations using the Bozzolo-Ferrante-Smith (BFS) method for the energetics for varying concentrations of either element are shown, indicating that Be could have a substantial effect in stopping Al interdiffusion, while Bi does not. Details of the calculations and the dependence of the role of each alloying addition as a function of temperature and concentration (of beryllium or bismuth in Al) are shown.

  15. Atomistic modeling and simulation of the role of Be and Bi in Al diffusion in U-Mo fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hofman, G.L. [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States); Bozzolo, G., E-mail: guille_bozzolo@yahoo.com [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States); Mosca, H.O. [Gerencia de Investigaciones y Aplicaciones, CNEA, Av. Gral Paz 1499, B165KNA, San Martin, Buenos Aires (Argentina); Yacout, A.M. [Argonne National Laboratory, 9700 S. Cass Ave., Argonne, IL 60439 (United States)

    2011-07-15

    Within the RERTR program, previous experimental and modeling studies identified Si as the alloying addition to the Al cladding responsible for inhibiting Al interdiffusion in the UMo fuel. However, difficulties with reprocessing have rendered this choice inappropriate, leading to the need to study alternative elements. In this work, we discuss the results of an atomistic modeling effort which allows for the systematic study of several possible alloying additions. Based on the behavior observed in the phase diagrams, beryllium or bismuth additions suggest themselves as possible options to replace Si. The results of temperature-dependent simulations using the Bozzolo-Ferrante-Smith (BFS) method for the energetics for varying concentrations of either element are shown, indicating that Be could have a substantial effect in stopping Al interdiffusion, while Bi does not. Details of the calculations and the dependence of the role of each alloying addition as a function of temperature and concentration (of beryllium or bismuth in Al) are shown.

  16. SEM and TEM Characterization of As-Fabricated U-7Mo Disperson Fuel Plates

    International Nuclear Information System (INIS)

    Keiser, D.D. Jr.; Yao, B.; Perez, E.; Sohn, Y.H.

    2009-01-01

    The starting microstructure of a dispersion fuel plate can have a dramatic impact on the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of dispersion fuel plates, SEM and TEM analysis have been performed on RERTR-9A archive fuel plates, which went through an additional hot isostatic procsssing (HIP) step during fabrication. The fuel plates had depleted U-7Mo fuel particles dispersed in either Al-2Si or 4043 Al alloy matrix. For the characterized samples, it was observed that a large fraction of the ?-phase U-7Mo alloy particles had decomposed during fabrication, and in areas near the fuel/matrix interface where the transformation products were present significant fuel/matrix interaction had occurred. Relatively thin Si-rich interaction layers were also observed around the U-7Mo particles. In the thick interaction layers, (U)(Al,Si)3 and U6Mo4Al43 were identified, and in the thin interaction layers U(Al,Si)3, U3Si3Al2, U3Si5, and USi1.88-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this work, exposure of dispersion fuel plates to relatively high temperatures during fabrication impacts the overall microstructure, particularly the nature of the interaction layers around the fuel particles. The time and temperature of fabrication should be carefully controlled in order to produce the most uniform Si-rich layers around the U-7Mo particles.

  17. Alternative Fuels

    Science.gov (United States)

    Alternative fuels include gaseous fuels such as hydrogen, natural gas, and propane; alcohols such as ethanol, methanol, and butanol; vegetable and waste-derived oils; and electricity. Overview of alternative fuels is here.

  18. From high enriched to low enriched uranium fuel in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Van Den Berghe, S.; Leenaers, A.; Koonen, E.; Moons, F.; Sannen, L. [Nuclear Materials Science Institute, SCK.CEN, Boeretang 200, B-2400 Mol (Belgium)

    2010-07-01

    Since the 1970's, global efforts have been going on to replace the high-enriched (>90% {sup 235}U), low-density UAlx research reactor fuel with high-density, low enriched (<20% {sup 235}U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U{sub 3}Si{sub 2} dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U{sub 3}Si{sub 2} (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development. (authors)

  19. From high enriched to low enriched uranium fuel in research reactors

    International Nuclear Information System (INIS)

    Van Den Berghe, S.; Leenaers, A.; Koonen, E.; Moons, F.; Sannen, L.

    2010-01-01

    Since the 1970's, global efforts have been going on to replace the high-enriched (>90% 235 U), low-density UAlx research reactor fuel with high-density, low enriched ( 235 U) replacements. This search is driven by the attempt to reduce the civil use of high-enriched material because of proliferation risks and terrorist threats. American initiatives, such as the Global Threat Reduction Initiative (GTRI) and the Reduced Enrichment for Research and Test Reactors (RERTR) program have triggered the development of reliable low-enriched fuel types for these reactors, which can replace the high enriched ones without loss of performance. Most success has presently been obtained with U 3 Si 2 dispersion fuel, which is currently used in many research reactors in the world. However, efforts to search for a replacement with even higher density, which will also allow the conversion of some high flux research reactors that currently cannot change to U 3 Si 2 (eg. BR2 in Belgium), have continued and are for the moment mainly directed towards the U(Mo) alloy fuel (7-10 w% Mo). This paper provides an overview of the past efforts and presents the current status of the U(Mo) development. (authors)

  20. Fuel assembly

    International Nuclear Information System (INIS)

    Chaki, Masao; Nishida, Koji; Karasawa, Hidetoshi; Kanazawa, Toru; Orii, Akihito; Nagayoshi, Takuji; Kashiwai, Shin-ichi; Masuhara, Yasuhiro

    1998-01-01

    The present invention concerns a fuel assembly, for a BWR type nuclear reactor, comprising fuel rods in 9 x 9 matrix. The inner width of the channel box is about 132mm and the length of the fuel rods which are not short fuel rods is about 4m. Two water rods having a circular cross section are arranged on a diagonal line in a portion of 3 x 3 matrix at the center of the fuel assembly, and two fuel rods are disposed at vacant spaces, and the number of fuel rods is 74. Eight fuel rods are determined as short fuel rods among 74 fuel rods. Assuming the fuel inventory in the short fuel rod as X(kg), and the fuel inventory in the fuel rods other than the short fuel rods as Y(kg), X and Y satisfy the relation: X + Y ≥ 173m, Y ≤ - 9.7X + 292, Y ≤ - 0.3X + 203 and X > 0. Then, even when the short fuel rods are used, the fuel inventory is increased and fuel economy can be improved. (I.N.)

  1. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  2. Fuel Management at the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham, V.L.; Nguyen, N.D.; Luong, B.V.; Le, V.V.; Huynh, T.N.; Nguyen, K.C. [Nuclear Research Institute, 01 Nguyen Tu Luc Street, Dalat City (Viet Nam)

    2011-07-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the old 250 kW TRIGA-MARK II reactor. The spent fuel storage was newly designed and installed in the place of the old thermalizing column for biological irradiation. The core was loaded by Russian WWR-M2 fuel assemblies (FAs) with 36% enrichment. The reconstructed reactor reached its initial criticality in November 1983 and attained it nominal power of 500 kW in February 1984. The first fuel reloading was executed in April 1994 after more than 10 years of operation with 89 highly enriched uranium (HEU) FAs. The third fuel reloading by shuffling of HEU FAs was executed in June 2004. After the shuffling the working configuration of reactor core kept unchanged of 104 HEU FAs. The fourth fuel reloading was executed in November 2006. The 2 new HEU FAs were loaded in the core periphery, at previous locations of wet irradiation channel and dry irradiation channel. After reloading the working configuration of reactor core consisted of 106 HEU FAs. Contracts for reactor core conversion between USA, Russia, Vietnam and the International Atomic Energy Agency for Nuclear fuel manufacture and supply for DNRR and Return of Russian-origin non-irradiated highly enriched uranium fuel to the Russian Federation have been realized in 2007. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory and Vietnam Atomic Energy Institute the mixed core configurations of irradiated HEU and new low enriched uranium (LEU) FAs has been created on 12 September, 2007 and on 20 July, 2009. After reloading in 2009, the 14 HEU FAs with highest burnup were removed from the core and put in the interim storage in reactor pool. The works on full core conversion for the DNRR are being realized in cooperation with the organizations, DOE and IAEA. Contract for Nuclear fuel manufacture and supply of 66 LEU FAs for DNRR

  3. Nuclear fuels

    International Nuclear Information System (INIS)

    Gangwani, Saloni; Chakrabortty, Sumita

    2011-01-01

    Nuclear fuel is a material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned for energy. Nuclear fuels are the most dense sources of energy available. Nuclear fuel in a nuclear fuel cycle can refer to the fuel itself, or to physical objects (for example bundles composed of fuel rods) composed of the fuel material, mixed with structural, neutron moderating, or neutron reflecting materials. Long-lived radioactive waste from the back end of the fuel cycle is especially relevant when designing a complete waste management plan for SNF. When looking at long-term radioactive decay, the actinides in the SNF have a significant influence due to their characteristically long half-lives. Depending on what a nuclear reactor is fueled with, the actinide composition in the SNF will be different. The following paper will also include the uses. advancements, advantages, disadvantages, various processes and behavior of nuclear fuels

  4. Fuel and nuclear fuel cycle

    International Nuclear Information System (INIS)

    Prunier, C.

    1998-01-01

    The nuclear fuel is studied in detail, the best choice and why in relation with the type of reactor, the properties of the fuel cans, the choice of fuel materials. An important part is granted to the fuel assembly of PWR type reactor and the performances of nuclear fuels are tackled. The different subjects for research and development are discussed and this article ends with the particular situation of mixed oxide fuels ( materials, behavior, efficiency). (N.C.)

  5. The analysis of effect of gaseous fission products on service-ability of various types of fuel elements used in research reactors

    International Nuclear Information System (INIS)

    Vatulin, A.A.; Kulakov, G.V.; Kosaurov, A.A.; Dobrikova, I.V.; Morozov, A.A.

    2005-01-01

    The analysis of the work to license U-Mo dispersion fuel that has been carried out by program RERTR shows that at high burnups the U-Mo particles interacts with an Al matrix to form porosity in the central areas of the fuel meat. This results leads to change the loading of cladding due to the formation of long-length defects in the fuel meat. As a result, the pressure of gases leads to a larger shape change (pillowing) of fuels and in some instances to their failure. The paper presents the results of modeling pillowing process proceeding in Russia's research reactor fuels of different types (tubular and rod fuels) that is effected by the accumulation and pressure of fission gas in meat porosity. The problem has been resolved in the viscous-plasticity statement using the program MARC for finite element analysis. The behavior of fission gas (FG) in open porosity has been analyzed to reveal the influence of the fuel design on their resistance to the pressure of FG. (author)

  6. Study of diffusion bond development in 6061 aluminum and its relationship to future high density fuels fabrication.

    Energy Technology Data Exchange (ETDEWEB)

    Prokofiev, I.; Wiencek, T.; McGann, D.

    1997-10-07

    Powder metallurgy dispersions of uranium alloys and silicides in an aluminum matrix have been developed by the RERTR program as a new generation of proliferation-resistant fuels. Testing is done with miniplate-type fuel plates to simulate standard fuel with cladding and matrix in plate-type configurations. In order to seal the dispersion fuel plates, a diffusion bond must exist between the aluminum coverplates surrounding the fuel meat. Four different variations in the standard method for roll-bonding 6061 aluminum were studied. They included mechanical cleaning, addition of a getter material, modifications to the standard chemical etching, and welding methods. Aluminum test pieces were subjected to a bend test after each rolling pass. Results, based on 400 samples, indicate that at least a 70% reduction in thickness is required to produce a diffusion bond using the standard rollbonding method versus a 60% reduction using the Type II method in which the assembly was welded 100% and contained open 9mm holes at frame corners.

  7. Fuel assembly

    International Nuclear Information System (INIS)

    Sakuyama, Tadashi; Mukai, Hideyuki.

    1988-01-01

    Purpose: To prevent the bending of a fuel rod caused by the difference in the elongation between a joined fuel rod and a standard fuel rod thereby maintain the fuel rod integrity. Constitution: A joined fuel rod is in a thread engagement at its lower end plug thereof with a lower plate, while passed through at its upper end plug into an upper tie plate and secured with a nut. Further, a standard fuel rod is engaged at its upper end plug and lower end plug with the upper tie plate and the lower tie plate respectively. Expansion springs are mounted to the upper end plugs of these bonded fuel rods and the standard fuel rods for preventing this lifting. Each of the fuel rods comprises a plurality of sintered pellets of nuclear fuel materials laminated in a zircaloy fuel can. The content of the alloy ingredient in the fuel can of the bonded fuel rod is made greater than that of the alloy ingredient of the standard fuel rod. this can increase the elongation for the bonded fuel rod, and the spring of the standard fuel rod is tightly bonded to prevent the bending. (Yoshino, Y.)

  8. Fuel processing

    International Nuclear Information System (INIS)

    Allardice, R.H.

    1990-01-01

    The technical and economic viability of the fast breeder reactor as an electricity generating system depends not only upon the reactor performance but also on a capability to recycle plutonium efficiently, reliably and economically through the reactor and fuel cycle facilities. Thus the fuel cycle is an integral and essential part of the system. Fuel cycle research and development has focused on demonstrating that the challenging technical requirements of processing plutonium fuel could be met and that the sometimes conflicting requirements of the fuel developer, fuel fabricator and fuel reprocessor could be reconciled. Pilot plant operation and development and design studies have established both the technical and economic feasibility of the fuel cycle but scope for further improvement exists through process intensification and flowsheet optimization. These objectives and the increasing processing demands made by the continuing improvement to fuel design and irradiation performance provide an incentive for continuing fuel cycle development work. (author)

  9. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    International Nuclear Information System (INIS)

    Sinha, V.P.; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P.

    2009-01-01

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and γ-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes

  10. Development, preparation and characterization of uranium molybdenum alloys for dispersion fuel application

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, V.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)], E-mail: vedsinha@barc.gov.in; Prasad, G.J.; Hegde, P.V.; Keswani, R.; Basak, C.B.; Pal, S.; Mishra, G.P. [Metallic Fuels Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2009-04-03

    Most of the research and test reactors worldwide have undergone core conversion from high enriched uranium base fuel to low enriched uranium base fuel under the Reduced Enrichment for Research and Test Reactor (RERTR) program, which was launched in the late 1970s to reduce the risk of nuclear proliferation. To realize this goal, high density uranium compounds and {gamma}-stabilized uranium alloy powder were identified. In Metallic Fuels Division of BARC, R and D efforts are on to develop these high density uranium base alloys. This paper describes the preparation flow sheet for different compositions of Uranium and molybdenum alloys by an innovative powder processing route with uranium and molybdenum metal powders as starting materials. The same composition of U-Mo alloys were also fabricated by conventional method i.e. ingot metallurgy route. The U-Mo alloys prepared by both the methods were then characterized by XRD for phase analysis. The photomicrographs of alloys with different compositions prepared by powder metallurgy and ingot metallurgy routes are also included in the paper. The paper also covers the comparison of properties of the alloys prepared by powder metallurgy and ingot metallurgy routes.

  11. Nuclear fuels

    International Nuclear Information System (INIS)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F.

    2009-01-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO 2 pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO 2 and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under irradiation

  12. Nuclear fuels

    Energy Technology Data Exchange (ETDEWEB)

    Beauvy, M.; Berthoud, G.; Defranceschi, M.; Ducros, G.; Guerin, Y.; Limoge, Y.; Madic, Ch.; Santarini, G.; Seiler, J.M.; Sollogoub, P.; Vernaz, E.; Guillet, J.L.; Ballagny, A.; Bechade, J.L.; Bonin, B.; Brachet, J.Ch.; Delpech, M.; Dubois, S.; Ferry, C.; Freyss, M.; Gilbon, D.; Grouiller, J.P.; Iracane, D.; Lansiart, S.; Lemoine, P.; Lenain, R.; Marsault, Ph.; Michel, B.; Noirot, J.; Parrat, D.; Pelletier, M.; Perrais, Ch.; Phelip, M.; Pillon, S.; Poinssot, Ch.; Vallory, J.; Valot, C.; Pradel, Ph.; Bonin, B.; Bouquin, B.; Dozol, M.; Lecomte, M.; Vallee, A.; Bazile, F.; Parisot, J.F.; Finot, P.; Roberts, J.F

    2009-07-01

    Fuel is one of the essential components in a reactor. It is within that fuel that nuclear reactions take place, i.e. fission of heavy atoms, uranium and plutonium. Fuel is at the core of the reactor, but equally at the core of the nuclear system as a whole. Fuel design and properties influence reactor behavior, performance, and safety. Even though it only accounts for a small part of the cost per kilowatt-hour of power provided by current nuclear power plants, good utilization of fuel is a major economic issue. Major advances have yet to be achieved, to ensure longer in-reactor dwell-time, thus enabling fuel to yield more energy; and improve ruggedness. Aside from economics, and safety, such strategic issues as use of plutonium, conservation of resources, and nuclear waste management have to be addressed, and true technological challenges arise. This Monograph surveys current knowledge regarding in-reactor behavior, operating limits, and avenues for R and D. It also provides illustrations of ongoing research work, setting out a few noteworthy results recently achieved. Content: 1 - Introduction; 2 - Water reactor fuel: What are the features of water reactor fuel? 9 (What is the purpose of a nuclear fuel?, Ceramic fuel, Fuel rods, PWR fuel assemblies, BWR fuel assemblies); Fabrication of water reactor fuels (Fabrication of UO{sub 2} pellets, Fabrication of MOX (mixed uranium-plutonium oxide) pellets, Fabrication of claddings); In-reactor behavior of UO{sub 2} and MOX fuels (Irradiation conditions during nominal operation, Heat generation, and removal, The processes involved at the start of irradiation, Fission gas behavior, Microstructural changes); Water reactor fuel behavior in loss of tightness conditions (Cladding, the first containment barrier, Causes of failure, Consequences of a failure); Microscopic morphology of fuel ceramic and its evolution under irradiation; Migration and localization of fission products in UOX and MOX matrices (The ceramic under

  13. Studsvik's implementation of the DOE spent fuel return program: The US perspective

    International Nuclear Information System (INIS)

    Grover, Stephen; Browser, Rita C.

    1996-01-01

    The government of Sweden has long been a proponent of nuclear nonproliferation policies. This includes laws governing the conduct of the Studsvik R2 Reactor in Nykoeping, Sweden. Studsvik became a participant in the RERTR program early on and was one of the first research reactors to convert from Highly Enriched Uranium (HEU) to Low Enriched Uranium (LEU) fuel. Since Studsvik operates the reactor on a commercial basis, it is important to fully comply with all Swedish laws and regulations for the operation of a nuclear reactor. This includes development of a proper long term program for the final disposition for the irradiated fuel. For this reason it was very important to Studsvik to be involved in the process for development of the DOE Spent Fuel Return Program as early as possible. This process included input into the Environmental Impact Statement process, early contractual negotiations after the issuance of the Record of Decision, participation in litigation involving the Spent Fuel Return Program, and plans for return of HEU fuel on the first European shipment. Not only was program involvement important, but actual program implementation was necessary for Studsvik to maintain its reactor license. Studsvik needed to have fuel included in the first shipment in the return program not only for licensing considerations, but also to reduce the proliferation risk associated with the storage of non-self protecting fuel. To ensure participation in the first shipment, Studsvik's activities included the early development of Appendix A's, coordinating the shipment, obtaining approvals and authorizations, and other technical aspects. (author)

  14. Conceptual designs parameters for MURR LEU U-Mo fuel conversion design demonstration experiment. Revision 1

    International Nuclear Information System (INIS)

    Stillman, J.; Feldman, E.; Stevens, J.

    2013-01-01

    The design parameters for the conceptual design of a fuel assembly containing U-10Mo fuel foils with low-enriched uranium (LEU) for the University of Missouri Research Reactor (MURR) are described. The Design Demonstration Experiment (MURR-DDE) will use a prototypic MURR-LEU element manufactured according to the parameters specified here. Also provided are calculated performance parameters for the LEU element in the MURR, and a set of goals for the MURR-DDE related to those parameters. The conversion objectives are to develop a fuel element design that will ensure safe reactor operations, as well as maintaining existing performance. The element was designed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. A set of manufacturing assumptions were provided by the Fuel Development (FD) and Fuel Fabrication Capability (FFC) pillars of the GTRI Reduced Enrichment for Research and Test Reactors (RERTR) program to reliably manufacture the fuel plates. The proposed LEU fuel element has an overall design and exterior dimensions that are similar to those of the current highly-enriched uranium (HEU) fuel elements. There are 23 fuel plates in the LEU design. The overall thickness of each plate is 44 mil, except for the exterior plate that is furthest from the center flux trap (plate 23), which is 49 mil thick. The proposed LEU fuel plates have U-10Mo monolithic fuel foils with a 235U enrichment of 19.75% varying from 9 mil to 20 mil thick, and clad with Al-6061 aluminum. A thin layer of zirconium exists between the fuel foils and the aluminum as a diffusion barrier. The thinnest nominal combined zirconium and aluminum clad thickness on each side of the fuel plates is 12 mil. The LEU U-10Mo monolithic fuel is not yet qualified as driver fuel in research reactors, but is under intense development under the auspices of the GTRI FD and FFC programs.

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1970-01-01

    Herein disclosed is a fuel assembly in which a fuel rod bundle is easily detachable by rotating a fuel rod fastener rotatably mounted to the upper surface of an upper tie-plate supporting a fuel bundle therebelow. A locking portion at the leading end of each fuel rod protrudes through the upper tie-plate and is engaged with or separated from the tie-plate by the rotation of the fastener. The removal of a desired fuel rod can therefore be remotely accomplished without the necessity of handling pawls, locking washers and nuts. (Owens, K.J.)

  16. Nuclear fuel

    International Nuclear Information System (INIS)

    D Hondt, P.

    1998-01-01

    The research and development programme on nuclear fuel at the Belgian Nuclear Research Centre SCK/CEN is described. The objective of this programme is to enhance the quantitative prediction of the operational limits of nuclear fuel and to assess the behaviour of fuel under incidental and accidental conditions. Progress is described in different domains including the modelling of fission gas release in LWR fuel, thermal conductivity, basic physical phenomena, post-irradiation examination for fuel performance assessment, and conceptual studies of incidental and accidental fuel experiments

  17. MTR spent fuel back-end - Cogema's long-term commitment

    International Nuclear Information System (INIS)

    Thomasson, J.

    1998-01-01

    MTR spent fuel back end has been subject to many reversal and uncertainties in the past 10 years. Until the end of 1988, US obligated materials were subject to the Off site Fuels Policy (OFP). Under this policy, spent fuels were returned to USA, and were reprocessed there. This OFP took end the 31th of December 1988, and Research Reactor's operators had to implement others solutions: On site storage or Reprocessing in Europe. Meanwhile the RERTR Program was leading to a new LEU fuel to replace HEU aluminide. This new silicide fuel has one main drawback: it cannot be reprocessed in working plants without some process main line modifications. Fortunately, a new Research Reactors spent fuels return policy has been set up by the US in the early 1996. This new policy applies to all reactors converted or that have agreed to convert to LEU, and reactors operating with HEU for which no suitable LEU is available. It covers all the spent fuels discharged until 2006/05/12. But after that period of time, each reactor will be fully responsible for its spent fuels. Since the end of 1996, COGEMA is proposing reprocessing services for Aluminides spent fuels, based on the La Hague capability. This COGEMA answer is for the long term, as the La Hague plant has a good load for the coming years, including the first decade of the next century. Further, this activity benefits from a strong R and D support, that allowed fulfilling the evolutive needs of our customers, and gives us the ability to adapt the plant to the future market. Taking advantage of this flexibility, COGEMA offers Research Reactors' operators a long-term commitment. Already two reactors' operators have chosen to contract with COGEMA for the whole life of their reactors. The contracts execution is under progress and the first transportation will take place soon. Beside today's services, COGEMA is involved in R and D activities to support new fuels development enhancing present LEU performances and having the ability to

  18. Fuel management

    International Nuclear Information System (INIS)

    Schwarz, E.R.

    1975-01-01

    Description of the operation of power plants and the respective procurement of fuel to fulfil the needs of the grid. The operation of the plants shall be optimised with respect to the fuel cost. (orig./RW) [de

  19. Fuel gases

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    This paper gives a brief presentation of the context, perspectives of production, specificities, and the conditions required for the development of NGV (Natural Gas for Vehicle) and LPG-f (Liquefied Petroleum Gas fuel) alternative fuels. After an historical presentation of 80 years of LPG evolution in vehicle fuels, a first part describes the economical and environmental advantages of gaseous alternative fuels (cleaner combustion, longer engines life, reduced noise pollution, greater natural gas reserves, lower political-economical petroleum dependence..). The second part gives a comparative cost and environmental evaluation between the available alternative fuels: bio-fuels, electric power and fuel gases, taking into account the processes and constraints involved in the production of these fuels. (J.S.)

  20. Fuel cycles

    International Nuclear Information System (INIS)

    Hawley, N.J.

    1983-05-01

    AECL publications, from the open literature, on fuels and fuel cycles used in CANDU reactors are listed in this bibliography. The accompanying index is by subject. The bibliography will be brought up to date periodically

  1. Nuclear fuels

    International Nuclear Information System (INIS)

    2008-01-01

    The nuclear fuel is one of the key component of a nuclear reactor. Inside it, the fission reactions of heavy atoms, uranium and plutonium, take place. It is located in the core of the reactor, but also in the core of the whole nuclear system. Its design and properties influence the behaviour, the efficiency and the safety of the reactor. Even if it represents a weak share of the generated electricity cost, its proper use represents an important economic stake. Important improvements remain to be made to increase its residence time inside the reactor, to supply more energy, and to improve its robustness. Beyond the economical and safety considerations, strategical questions have to find an answer, like the use of plutonium, the management of resources and the management of nuclear wastes and real technological challenges have to be taken up. This monograph summarizes the existing knowledge about the nuclear fuel, its behaviour inside the reactor, its limits of use, and its R and D tracks. It illustrates also the researches in progress and presents some key results obtained recently. Content: 1 - Introduction; 2 - The fuel of water-cooled reactors: aspect, fabrication, behaviour of UO 2 and MOX fuels inside the reactor, behaviour in loss of tightness situation, microscopic morphology of fuel ceramics and evolution under irradiation - migration and localisation of fission products in UOX and MOX matrices, modeling of fuels behaviour - modeling of defects and fission products in the UO 2 ceramics by ab initio calculations, cladding and assembly materials, pellet-cladding interaction, advanced UO 2 and MOX ceramics, mechanical behaviour of the fuel assembly, fuel during a loss of coolant accident, fuel during a reactivity accident, fuel during a serious accident, fuel management inside reactor cores, fuel cycle materials balance, long-term behaviour of the spent fuel, fuel of boiling water reactors; 3 - the fuel of liquid metal fast reactors: fast neutrons radiation

  2. Fuel pellet

    International Nuclear Information System (INIS)

    Hayashi, K.

    1980-01-01

    Fuel pellet for insertion into a cladding tube in order to form a fuel element or a fuel rod. The fuel pellet has got a belt-like projection around its essentially cylindrical lateral circumferential surface. The upper and lower edges in vertical direction of this belt-like projection are wave-shaped. The projection is made of the same material as the bulk pellet. Both are made in one piece. (orig.) [de

  3. Fossil Fuels.

    Science.gov (United States)

    Crank, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fossil fuels. Some topics covered are historic facts, development of fuels, history of oil production, current and future trends of the oil industry, refining fossil fuels, and environmental problems. Material in each unit may…

  4. Fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A new fuel can with a loose bottom and head is described. The fuel bar is attached to the loose bottom and head with two grid poles keeping the distance between bottom and head. A bow-shaped handle is attached to the head so that the fuel bar can be lifted from the can

  5. LPG fuel

    International Nuclear Information System (INIS)

    Dagnas, F.X.; Jeuland, N.; Fouquet, J.P.; Lauraire, S.; Coroller, P.

    2005-01-01

    LPG fuel has become frequently used through a distribution network with 2 000 service stations over the French territory. LPG fuel ranks number 3 world-wide given that it can be used on individual vehicles, professional fleets, or public transport. What is the environmental benefit of LPG fuel? What is the technology used for these engines? What is the current regulation? Government commitment and dedication on support to promote LPG fuel? Car makers projects? Actions to favour the use of LPG fuel? This article gathers 5 presentations about this topic given at the gas conference

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto; Ogiya, Shunsuke.

    1989-01-01

    For improving the economy of a BWR type reactor by making the operation cycle longer, the fuel enrichment degree has to be increased further. However, this makes the subcriticality shallower in the upper portion of the reactor core, to bring about a possibility that the reactor shutdown becomes impossible. In the present invention, a portion of fuel rod is constituted as partial length fuel rods (P-fuel rods) in which the entire stack length in the effective portion is made shorter by reducing the concentration of fissionable materials in the axial portion. A plurality of moderator rods are disposed at least on one diagonal line of a fuel assembly and P-fuel rods are arranged at a position put between the moderator rods. This makes it possible to reactor shutdown and makes the axial power distribution satisfactory even if the fuel enrichment degree is increased. (T.M.)

  7. Fuel Services

    International Nuclear Information System (INIS)

    Silberstein, A.

    1982-09-01

    FRAGEMA has developed most types of inspection equipments to work on irradiated fuel assemblies and on single fuel rods during reactor outages with an efficiency compatible with the utilities operating priorities. In order to illustrate this statement, two specific examples of inspection equipments are shortly described: the on-site removable fuel rod assembly examination stand, and the fuel assembly multiple examination device. FRAGEMA has developed techniques for the identifiction of the leaking fuel rods in the fuel assembly and the tooling necessary to perform the replacement of the faulted element. These examples of methods, techniques and equipments described and the experience accumulated through their use allow FRAGEMA to qualify for offering the supply of the corresponding software, hardware or both whenever an accurate understanding of the fuel behaviour is necessary and whenever direct intervention on the assembly and associated components is necessary due to safety, operating or economical reasons

  8. Fuel assembly

    International Nuclear Information System (INIS)

    Watanabe, Shoichi; Hirano, Yasushi.

    1998-01-01

    A one-half or more of entire fuel rods in a fuel assembly comprises MOX fuel rods containing less than 1wt% of burnable poisons, and at least a portion of the burnable poisons comprises gadolinium. Then, surplus reactivity at an initial stage of operation cycle is controlled to eliminate burnable poisons remained unburnt at a final stage, as well as increase thermal reactivity. In addition, the content of fission plutonium is determined to greater than the content of uranium 235, and fuel rods at corner portions are made not to incorporate burnable poisons. Fuel rods not containing burnable poisons are disposed at positions in adjacent with fuel rods facing to a water rod at one or two directions. Local power at radial center of the fuel assembly is increased to flatten the distortion of radial power distribution. (N.H.)

  9. Neutronic performance of high-density LEU fuels in water-moderated and water-reflected research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Matos, J.E.

    1996-01-01

    At the Reduced Enrichment for Research and Test Reactors (RERTR) meeting in September 1994, Durand reported that the maximum uranium loading attainable with U 3 Si 2 fuel is about 6.0 g U/cm 3 . The French Commissariat a l'Energie Atomique (CEA) plan to perform irradiation tests with 5 plates at this loading. Compagnie pour L'Etude et La Realisation de Combustibles Atomiques (CERCA) has also fabricated a few uranium nitride (UN) plates with a uranium density in the fuel meat of 7.0 g/cm 3 and found that UN is compatible with the aluminum matrix at temperatures below 500 C. High density dispersion fuels proposed for development include U-Zr(4 wt%)-Nb(2 wt%), U-Mo(5 wt%), and U-Mo(9 wt%). The purpose of this note is to examine the relative neutronic behavior of these high density fuels in a typical light water-reflected and water-moderated MTR-type research reactor. The results show that a dispersion of the U-Zr-Nb alloy has the most favorable neutronic properties and offers the potential for uranium densities greater than 8.0 g/cm 3 . On the other hand, UN is the least reactive fuel because of the relatively large 14 N(n,p) cross section. For a fixed value of k eff , the required 235 U loading per fuel element is least for the U-Zr-Nb fuel and steadily increases for the U-Mo(5%), U-Mo(9%), and UN fuels. Because of volume fraction limitations, the UO 2 dispersions are only useful for uranium densities below 5.0 g/cm 3 . In this density range, however, UO 2 is more reactive than U 3 Si 2

  10. Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

    Energy Technology Data Exchange (ETDEWEB)

    Perez, E.; Yao, B. [Advanced Materials Processing and Analysis Center, Department of Mechanical, Materials and Aerospace Engineering, University of Central Florida, Orlando, FL 32816 (United States); Keiser, D.D. [Nuclear Fuels and Materials Division, Idaho National Laboratory, Scoville, ID 83415 (United States); Sohn, Y.H., E-mail: ysohn@mail.ucf.ed [Advanced Materials Processing and Analysis Center, Department of Mechanical, Materials and Aerospace Engineering, University of Central Florida, Orlando, FL 32816 (United States)

    2010-07-01

    For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr{sub 2}, {gamma}-UZr, Zr solid-solution and Mo{sub 2}Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si){sub 2}Zr, (Al, Si)Zr{sub 3} (Al, Si){sub 3}Zr, and AlSi{sub 4}Zr{sub 5}. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

  11. Microstructural analysis of as-processed U-10 wt.%Mo monolithic fuel plate in AA6061 matrix with Zr diffusion barrier

    Science.gov (United States)

    Perez, E.; Yao, B.; Keiser, D. D., Jr.; Sohn, Y. H.

    2010-07-01

    For higher U-loading in low-enriched U-10 wt.%Mo fuels, monolithic fuel plate clad in AA6061 is being developed as a part of Reduced Enrichment for Research and Test Reactor (RERTR) program. This paper reports the first characterization results from a monolithic U-10 wt.%Mo fuel plate with a Zr diffusion barrier that was fabricated as part of a plate fabrication campaign for irradiation testing in the Advanced Test Reactor (ATR). Both scanning and transmission electron microscopy (SEM and TEM) were employed for analysis. At the interface between the Zr barrier and U-10 wt.%Mo, going from Zr to U(Mo), UZr 2, γ-UZr, Zr solid-solution and Mo 2Zr phases were observed. The interface between AA6061 cladding and Zr barrier plate consisted of four layers, going from Al to Zr, (Al, Si) 2Zr, (Al, Si)Zr 3 (Al, Si) 3Zr, and AlSi 4Zr 5. Irradiation behavior of these intermetallic phases is discussed based on their constituents. Characterization of as-fabricated phase constituents and microstructure would help understand the irradiation behavior of these fuel plates, interpret post-irradiation examination, and optimize the processing parameters of monolithic fuel system.

  12. Neutronic feasibility studies using U-Mo dispersion fuel (9 Wt % Mo, 5.0 gU/cm3) for LEU conversion of the MARIA (Poland), IR-8 (Russia), and WWR-SM (Uzbekistan) research reactors

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Deen, J.R.; Hanan, N.A.; Matos, E.

    2000-01-01

    U-Mo alloys dispersed in an Al matrix offer the potential for high-density uranium fuels needed for the LEU conversion of many research reactors. On-going fuel qualification tests by the US RERTR Program show good irradiation properties of U-Mo alloy dispersion fuel containing 7-10 weight percent molybdenum. For the neutronic studies in this paper the alloy was assumed to contain 9 wt % Mo (U-9Mo) with a uranium density in the fuel meat of 5.00 gU/cm 3 which corresponds to 32.5 volume % U-9Mo. Fuels containing U-9Mo have been used in Russian reactors since the 1950's. For the three research reactors analyzed here, LEU fuel element thicknesses are the same as those for the Russian-fabricated HEU reference fuel elements. Relative to the reference fuels containing 80-90% enriched uranium, LEU U-9Mo Al-dispersion fuel with 5.00 gU/cm 3 doubles the cycle length of the MARIA reactor and increases the IR-8 cycle length by about 11%. For the WWR-SM reactor, the cycle length, and thus the number of fuel assemblies used per year, is nearly unchanged. To match the cycle length of the 36% enriched fuel currently used in the WWR-SM reactor will require a uranium density in the LEU U-9Mo Al-dispersion fuel of about 5.4 gU/cm 3 . The 5.00 gU/cm 3 LEU fuel causes thermal neutron fluxes in water holes near the edge of the core to decrease by (6-8)% for all three reactors. (author)

  13. Fuel assembly

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi; Matsuzuka, Ryuji.

    1976-01-01

    Object: To provide a fuel assembly which can decrease pressure loss of coolant to uniform temperature. Structure: A sectional area of a flow passage in the vicinity of an inner peripheral surface of a wrapper tube is limited over the entire length to prevent the temperature of a fuel element in the outermost peripheral portion from being excessively decreased to thereby flatten temperature distribution. To this end, a plurality of pincture-frame-like sheet metals constituting a spacer for supporting a fuel assembly, which has a plurality of fuel elements planted lengthwise and in given spaced relation within the wrapper tube, is disposed in longitudinal grooves and in stacked fashion to form a substantially honeycomb-like space in cross section. The fuel elements are inserted and supported in the space to form a fuel assembly. (Kamimura, M.)

  14. Fuel assemblies

    International Nuclear Information System (INIS)

    Nagano, Mamoru; Yoshioka, Ritsuo

    1983-01-01

    Purpose: To effectively utilize nuclear fuels by increasing the reactivity of a fuel assembly and reduce the concentration at the central region thereof upon completion of the burning. Constitution: A fuel assembly is bisected into a central region and a peripheral region by disposing an inner channel box within a channel box. The flow rate of coolants passing through the central region is made greater than that in the peripheral region. The concentration of uranium 235 of the fuel rods in the central region is made higher. In such a structure, since the moderating effect in the central region is improved, the reactivity of the fuel assembly is increased and the uranium concentration in the central region upon completion of the burning can be reduced, fuel economy and effective utilization of uranium can be attained. (Kamimura, M.)

  15. Fuel assembly

    International Nuclear Information System (INIS)

    Bando, Masaru.

    1993-01-01

    As neutron irradiation progresses on a fuel assembly of an FBR type reactor, a strong force is exerted to cause ruptures if the arrangement of fuel elements is not displaced, whereas the fuel elements may be brought into direct contact with each other not by way of spacers to cause burning damages if the arrangement is displaced. In the present invention, the circumference of fuel elements arranged in a normal triangle lattice is surrounded by a wrapper tube having a hexagonal cross section, wire spacers are wound therearound, and deformable spacers are distributed to optional positions for fuel elements in the wrapper tube. Interaction between the fuel elements caused by irradiation is effectively absorbed, thereby enabling to delay the occurrence of the rupture and burning damages of the elements. (N.H.)

  16. Fuel assembly

    International Nuclear Information System (INIS)

    Yokota, Tokunobu.

    1990-01-01

    A fuel assembly used in a FBR type nuclear reactor comprises a plurality of fuel rods and a moderator guide member (water rod). A moderator exit opening/closing mechanism is formed at the upper portion of the moderator guide member for opening and closing a moderator exit. In the initial fuel charging operation cycle to the reactor, the moderator exit is closed by the moderator exit opening/closing mechanism. Then, voids are accumulated at the inner upper portion of the moderator guide member to harden spectrum and a great amount of plutonium is generated and accumulated in the fuel assembly. Further, in the fuel re-charging operation cycle, the moderator guide member is used having the moderator exit opened. In this case, voids are discharged from the moderator guide member to decrease the ratio, and the plutonium accumulated in the initial charging operation cycle is burnt. In this way, the fuel economy can be improved. (I.N.)

  17. Fuel spacer

    International Nuclear Information System (INIS)

    Nishida, Koji; Yokomizo, Osamu; Kanazawa, Toru; Kashiwai, Shin-ichi; Orii, Akihito.

    1992-01-01

    The present invention concerns a fuel spacer for a fuel assembly of a BWR type reactor and a PTR type reactor. Springs each having a vane are disposed on the side surface of a circular cell which supports a fuel rods. A vortex streams having a vertical component are formed by the vanes in the flowing direction of a flowing channel between adjacent cylindrical cells. Liquid droplets carried by streams are deposited on liquid membrane streams flowing along the fuel rod at the downstream of the spacer by the vortex streams. In view of the above, the liquid droplets can be deposited to the fuel rod without increasing the amount of metal of the spacer. Accordingly, the thermal margin of the fuel assembly can be improved without losing neutron economy. (I.N.)

  18. Fuel Cells

    DEFF Research Database (Denmark)

    Smith, Anders; Pedersen, Allan Schrøder

    2014-01-01

    Fuel cells have been the subject of intense research and development efforts for the past decades. Even so, the technology has not had its commercial breakthrough yet. This entry gives an overview of the technological challenges and status of fuel cells and discusses the most promising applications...... of the different types of fuel cells. Finally, their role in a future energy supply with a large share of fluctuating sustainable power sources, e.g., solar or wind, is surveyed....

  19. Fuel cycle

    International Nuclear Information System (INIS)

    Bahm, W.

    1989-01-01

    The situation of the nuclear fuel cycle for LWR type reactors in France and in the Federal Republic of Germany was presented in 14 lectures with the aim to compare the state-of-the-art in both countries. In addition to the momentarily changing fuilds of fuel element development and fueling strategies, the situation of reprocessing, made interesting by some recent developmnts, was portrayed and differences in ultimate waste disposal elucidated. (orig.) [de

  20. Nuclear fuel

    International Nuclear Information System (INIS)

    Azevedo, J.B.L. de.

    1980-01-01

    All stages of nuclear fuel cycle are analysed with respect to the present situation and future perspectives of supply and demand of services; the prices and the unitary cost estimation of these stages for the international fuel market are also mentioned. From the world resources and projections of uranium consumption, medium-and long term analyses are made of fuel availability for several strategies of use of different reactor types. Finally, the cost of nuclear fuel in the generation of electric energy is calculated to be used in the energetic planning of the electric sector. (M.A.) [pt

  1. Fuel assembly

    International Nuclear Information System (INIS)

    Nomata, Terumitsu.

    1993-01-01

    Among fuel pellets to be loaded to fuel cans of a fuel assembly, fuel pellets having a small thermal power are charged in a region from the end of each of spacers up to about 50mm on the upstream of coolants that flow vertically at the periphery of fuel rods. Coolants at the periphery of fuel rods are heated by the heat generation, to result in voids. However, since cooling effect on the upstream of the spacers is low due to influences of the spacers. Further, since the fuel pellets disposed in the upstream region have small thermal power, a void coefficient is not increased. Even if a thermal power exceeding cooling performance should be generated, there is no worry of causing burnout in the upstream region. Even if burnout should be caused, safety margin and reliability relative to burnout are improved, to increase an allowable thermal power, thereby enabling to improve integrity and reliability of fuel rods and fuel assemblies. (N.H.)

  2. Fuel assembly

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Bassler, E.A.; Huckestein, E.A.; Salton, R.B.; Tower, S.N.

    1988-01-01

    A fuel assembly adapted for use with a pressurized water nuclear reactor having capabilities for fluid moderator spectral shift control is described comprising: parallel arranged elongated nuclear fuel elements; means for providing for axial support of the fuel elements and for arranging the fuel elements in a spaced array; thimbles interspersed among the fuel elements adapted for insertion of a rod control cluster therewithin; means for structurally joining the fuel elements and the guide thimbles; fluid moderator control means for providing a volume of low neutron absorbing fluid within the fuel assembly and for removing a substantially equivalent volume of reactor coolant water therefrom, a first flow manifold at one end of the fuel assembly sealingly connected to a first end of the moderator control tubes whereby the first ends are commonly flow connected; and a second flow manifold, having an inlet passage and an outlet passage therein, sealingly connected to a second end of the moderator control tubes at a second end of the fuel assembly

  3. Analytical analyses of startup measurements associated with the first use of LEU fuel in Romania's 14-MW TRIGA reactor

    International Nuclear Information System (INIS)

    Bretscher, M.M.; Snelgrove, J.L.; Ciocanescu, M.

    1992-01-01

    The 14-MW TRIGA steady state reactor (SSR) is located in Pitesti, Romania. Beginning with an HEU core (10 wt% U), the reactor first went critical in November 1979 but was shut down ten years later because of insufficient excess reactivity. Last November the Institute for Nuclear Research (INR), which operates the SSR, received from the ANL RERTR program a shipment of 125 LEU pins fabricated by General Atomics and of the same geometry as the original fuel but with an enrichment of 19.7% 235U and a loading of 45 wt% U. Using 100 of these pins, four LEU clusters, each containing a 5 x 5 square array of fuel rods, were assembled. These four LEU clusters replaced the four most highly burned HEU elements in the SSR. The reactor resumed operations last February with a 35-element mixed HEU/LEU core configuration. In preparation for full power operation of the SSR with this mixed HEU/LEU core, a number of measurements were made. These included control rod calibrations, excess reactivity determinations, worths of experiment facilities, reaction rate distributions, and themocouple measurements of fuel temperatures as a function of reactor power. This paper deals with a comparison of some of these measured reactor parameters with corresponding analytical calculations

  4. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Energy Technology Data Exchange (ETDEWEB)

    Rest, J. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)], E-mail: jrest@anl.gov; Hofman, G.L.; Kim, Yeon Soo [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2009-04-15

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than {approx}7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  5. Analysis of intergranular fission-gas bubble-size distributions in irradiated uranium-molybdenum alloy fuel

    Science.gov (United States)

    Rest, J.; Hofman, G. L.; Kim, Yeon Soo

    2009-04-01

    An analytical model for the nucleation and growth of intra and intergranular fission-gas bubbles is used to characterize fission-gas bubble development in low-enriched U-Mo alloy fuel irradiated in the advanced test reactor in Idaho as part of the Reduced Enrichment for Research and Test Reactor (RERTR) program. Fuel burnup was limited to less than ˜7.8 at.% U in order to capture the fuel-swelling stage prior to irradiation-induced recrystallization. The model couples the calculation of the time evolution of the average intergranular bubble radius and number density to the calculation of the intergranular bubble-size distribution based on differential growth rate and sputtering coalescence processes. Recent results on TEM analysis of intragranular bubbles in U-Mo were used to set the irradiation-induced diffusivity and re-solution rate in the bubble-swelling model. Using these values, good agreement was obtained for intergranular bubble distribution compared against measured post-irradiation examination (PIE) data using grain-boundary diffusion enhancement factors of 15-125, depending on the Mo concentration. This range of enhancement factors is consistent with values obtained in the literature.

  6. Effects of high density dispersion fuel loading on the uncontrolled reactivity insertion transients of a low enriched uranium fueled material test research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Muhammad, Farhan [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)], E-mail: farhan73@hotmail.com; Majid, Asad [Department of Nuclear Engineering, Pakistan Institute of Engineering and Applied Sciences, Nilore, Islamabad 45650 (Pakistan)

    2009-08-15

    The effects of using high density low enriched uranium on the uncontrolled reactivity insertion transients of a material test research reactor were studied. For this purpose, the low density LEU fuel of an MTR was replaced with high density U-Mo (9w/o) LEU fuels currently being developed under the RERTR program having uranium densities of 6.57 gU/cm{sup 3}, 7.74 gU/cm{sup 3} and 8.57 gU/cm{sup 3}. Simulations were carried out to determine the reactor performance under reactivity insertion transients with totally failed control rods. Ramp reactivities of 0.25$/0.5 s and 1.35$/0.5 s were inserted with reactor operating at full power level of 10 MW. Nuclear reactor analysis code PARET was employed to carry out these calculations. It was observed that when reactivity insertion was 0.25$/0.5 s, the new power level attained increased by 5.8% as uranium density increases from 6.57 gU/cm{sup 3} to 8.90 gU/cm{sup 3}. This results in increased maximum temperatures of fuel, clad and coolant outlet, achieved at the new power level, by 4.7 K, 4.4 K and 2.4 K, respectively. When reactivity insertion was 1.35$/0.5 s, the feedback reactivities were unable to control the reactor which resulted in the bulk boiling of the coolant; the one with the highest fuel density was the first to reach the boiling point.

  7. A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

    Energy Technology Data Exchange (ETDEWEB)

    Drera, Saleem S., E-mail: saleem.drera@gmail.com [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); Hofman, Gerard L. [Argonne National Laboratory, Chicago, IL 60439 (United States); Kee, Robert J. [Mechanical Engineering, Colorado School of Mines, Golden, CO 80401 (United States); King, Jeffrey C. [Metallurgical and Materials Engineering, Colorado School of Mines, Golden, CO 80401 (United States)

    2014-10-15

    Highlights: • This article presents a cellular automata (CA) algorithm to synthesize the growth of intermetallic interaction layers in U–Mo/Al dispersion fuel. • The method utilizes a 3D representation of the fuel, which is discretized into separate voxels that can change identy based on derived CA rules. • The CA model is compared to ILT measurements for RERTR experimental data. • The primary objective of the model is to synthesize three-dimensional microstructures that can be used in subsequent thermal and mechanical modeling. • The CA model can be used for predictive analysis. For example, it can be used to study the dependence of temperature on interaction layer growth. - Abstract: Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium–molybdenum (U–Mo) particles within an aluminum matrix. Fresh U–Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction–diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

  8. Feasibility study for LEU conversion of the WWR-K reactor at the Institute of Nuclear Physics in Kazakhstan using a 5-tube fuel assembly

    International Nuclear Information System (INIS)

    Hanan, N.A.; Liaw, J.R.; Matos, J.E.

    2005-01-01

    A feasibility study by the RERTR program for possible LEU conversion of the 6 MW WWR-K reactor concludes that conversion is feasible using an LEU 5-tube Russian fuel assembly design. This 5-tube design is one of several LEU fuel assembly designs being studied (Ref. 1) for possible use in this reactor. The 5-tube assembly contains 200 g 235 U with an enrichment of 19.7% in four cylindrical inner tubes and an outer hexagonal tube with the same external dimensions as the current HEU (36%) 5-tube fuel assembly, which contains 112.5 g 235 U. The fuel meat material, LEU UO 2 -Al dispersion fuel with ∼ 2.5 g U/cm 3 , has been extensively irradiation tested in a number of reactors with uranium enrichments of 36% and 19.7%. Since the 235 U loading of the LEU assemblies is much larger than the HEU assemblies, a smaller LEU core with five rows of fuel assemblies is possible (instead of six rows of fuel assemblies in the HEU core). This smaller LEU core would consume about 60% as many fuel assemblies per year as the current HEU core and provide thermal neutron fluxes in the inner irradiation channels that are ∼ 17% larger than with the present HEU core. The current 21 day cycle length would be maintained and the average discharge burnup would be ∼ 42%. Neutron fluxes in the five outer irradiation channels would be smaller in the LEU core unless these channels can be moved closer to the LEU fuel assemblies. Results show that the smaller LEU core would meet the reactor's shutdown margin requirements and would have an adequate thermal-hydraulic safety margin to onset of nucleate boiling. (author)

  9. Fuel element

    International Nuclear Information System (INIS)

    Kennedy, S.T.

    1982-01-01

    A nuclear reactor fuel element wherein a stack of nuclear fuel is prevented from displacement within its sheath by a retainer comprising a tube member which is radially expanded into frictional contact with the sheath by means of a captive ball within a tapered bore. (author)

  10. Nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Nakano, H [Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan)

    1976-10-01

    It is expected that nuclear power generation will reach 49 million kW in 1985 and 129 million kW in 1995, and the nuclear fuel having to be supplied and processed will increase in proportion to these values. The technical problems concerning nuclear fuel are presented on the basis of the balance between the benefit for human beings and the burden on the human beings. Recently, especially the downstream of nuclear fuel attracts public attention. Enriched uranium as the raw material for light water reactor fuel is almost monopolized by the U.S., and the technical information has not been published for fear of the diversion to nuclear weapons. In this paper, the present situations of uranium enrichment, fuel fabrication, transportation, reprocessing and waste disposal and the future problems are described according to the path of nuclear fuel cycle. The demand and supply of enriched uranium in Japan will be balanced up to about 1988, but afterwards, the supply must rely upon the early establishment of the domestic technology by centrifugal separation method. No problem remains in the fabrication of light water reactor fuel, but for the fabrication of mixed oxide fuel, the mechanization of the production facility and labor saving are necessary. The solution of the capital risk for the construction of the second reprocessing plant is the main problem. Japan must develop waste disposal techniques with all-out efforts.

  11. Fuel cells

    NARCIS (Netherlands)

    Veen, van J.A.R.; Janssen, F.J.J.G.; Santen, van R.A.

    1999-01-01

    The principles and present-day embodiments of fuel cells are discussed. Nearly all cells are hydrogen/oxygen ones, where the hydrogen fuel is usually obtained on-site from the reforming of methane or methanol. There exists a tension between the promise of high efficiency in the conversion of

  12. Fuel assembly

    International Nuclear Information System (INIS)

    Nakajima, Akiyoshi; Bessho, Yasunori; Aoyama, Motoo; Koyama, Jun-ichi; Hirakawa, Hiromasa; Yamashita, Jun-ichi; Hayashi, Tatsuo

    1998-01-01

    In a fuel assembly of a BWR type reactor in which a water rod of a large diameter is disposed at the central portion, the cross sectional area perpendicular to the axial direction comprises a region a of a fuel rod group facing to a wide gap water region to which a control rod is inserted, a region b of a fuel rod group disposed on the side of the wide gap water region other than the region a, a region d of a fuel rod group facing to a narrow gap water region and a region c of a fuel rod group disposed on the side of the narrow gap water region other than the region d. When comparing an amount of fission products contained in the four regions relative to that in the entire regions and average enrichment degrees of fuel rods for the four regions, the relative amount and the average enrichment degree of the fuel rod group of the region a is minimized, and the relative amount and the average enrichment degree of the fuel rod group in the region b is maximized. Then, reactor shut down margin during cold operation can be improved while flattening the power in the cross section perpendicular to the axial direction. (N.H.)

  13. Nuclear fuel

    International Nuclear Information System (INIS)

    Quinauk, J.P.

    1990-01-01

    Since 1985, Fragema has been marketing and selling the Advanced Fuel Assemby AFA whose main features are its zircaloy grids and removable top and bottom nozzles. It is this product, which exists for several different fuel assembly arrays and heights, that will be employed in the reactors at Daya Bay. Fragema employs gadolinium as the consumable poison to enable highperformance fuel management. More recently, the company has supplied fuel assemblies of the mixed-oxide(MOX) and enriched reprocessed uranium type. The reliability level of the fuel sold by Fragema is one of the highest in the world, thanks in particular to the excellence of the quality assurance and quality control programs that have been implemented at all stages of its design and manufacture

  14. Fuel assemblies

    International Nuclear Information System (INIS)

    Echigoya, Hironori; Nomata, Terumitsu.

    1983-01-01

    Purpose: To render the axial distribution relatively flat. Constitution: First nuclear element comprises a fuel can made of zircalloy i.e., the metal with less neutron absorption, which is filled with a plurality of UO 2 pellets and sealed by using a lower end plug, a plenum spring and an upper end plug by means of welding. Second fuel element is formed by substituting a part of the UO 2 pellets with a water tube which is sealed with water and has a space for allowing the heat expansion. The nuclear fuel assembly is constituted by using the first and second fuel elements together. In such a structure, since water reflects neutrons and decrease their leakage to increase the temperature, reactivity is added at the upper portion of the fuel assembly to thereby flatten the axial power distribution. Accordingly, stable operation is possible only by means of deep control rods while requiring no shallow control rods. (Sekiya, K.)

  15. Microstructural Characterization of a Mg Matrix U-Mo Dispersion Fuel Plate Irradiated in the Advanced Test Reactor to High Fission Density: SEM Results

    Science.gov (United States)

    Keiser, Dennis D.; Jue, Jan-Fong; Miller, Brandon D.; Gan, Jian; Robinson, Adam B.; Medvedev, Pavel G.; Madden, James W.; Moore, Glenn A.

    2016-06-01

    Low-enriched (U-235 RERTR-8 experiment at high temperature, high fission rate, and high power, up to high fission density. This paper describes the results of the scanning electron microscopy (SEM) analysis of an irradiated fuel plate using polished samples and those produced with a focused ion beam. A follow-up paper will discuss the results of transmission electron microscopy (TEM) analysis. Using SEM, it was observed that even at very aggressive irradiation conditions, negligible chemical interaction occurred between the irradiated U-7Mo fuel particles and Mg matrix; no interconnection of fission gas bubbles from fuel particle to fuel particle was observed; the interconnected fission gas bubbles that were observed in the irradiated U-7Mo particles resulted in some transport of solid fission products to the U-7Mo/Mg interface; the presence of microstructural pathways in some U-9.1 Mo particles that could allow for transport of fission gases did not result in the apparent presence of large porosity at the U-7Mo/Mg interface; and, the Mg-Al interaction layers that were present at the Mg matrix/Al 6061 cladding interface exhibited good radiation stability, i.e. no large pores.

  16. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David [ORNL; Chandler, David [ORNL; Cook, David [ORNL; Ilas, Germina [ORNL; Jain, Prashant [ORNL; Valentine, Jennifer [ORNL

    2014-10-30

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy’s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the “complex” aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The

  17. Preliminary Evaluation of Alternate Designs for HFIR Low-Enriched Uranium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Renfro, David G [ORNL; Chandler, David [ORNL; Cook, David Howard [ORNL; Ilas, Germina [ORNL; Jain, Prashant K [ORNL; Valentine, Jennifer R [ORNL

    2014-11-01

    Engineering design studies of the feasibility of conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel are ongoing at Oak Ridge National Laboratory (ORNL) as part of an effort sponsored by the U.S. Department of Energy s Global Threat Reduction Initiative (GTRI)/Reduced Enrichment for Research and Test Reactors (RERTR) program. The fuel type selected by the program for the conversion of the five high-power research reactors in the U.S. that still use HEU fuel is a new U-Mo monolithic fuel. Studies by ORNL have previously indicated that HFIR can be successfully converted using the new fuel provided (1) the reactor power can be increased from 85 MW to 100 MW and (2) the fuel can be fabricated to a specific reference design. Fabrication techniques for the new fuel are under development by the program but are still immature, especially for the complex aspects of the HFIR fuel design. In FY 2012, the program underwent a major shift in focus to emphasize developing and qualifying processes for the fabrication of reliable and affordable LEU fuel. In support of this new focus and in an effort to ensure that the HFIR fuel design is as suitable for reliable fabrication as possible, ORNL undertook the present study to propose and evaluate several alternative design features. These features include (1) eliminating the fuel zone axial contouring in the previous reference design by substituting a permanent neutron absorber in the lower unfueled region of all of the fuel plates, (2) relocating the burnable neutron absorber from the fuel plates of the inner fuel element to the side plates of the inner fuel element (the fuel plates of the outer fuel element do not contain a burnable absorber), (3) relocating the fuel zone inside the fuel plate to be centered on the centerline of the depth of the plate, and (4) reshaping the radial contour of the relocated fuel zone to be symmetric about this centerline. The present

  18. Fuel cells:

    DEFF Research Database (Denmark)

    Sørensen, Bent

    2013-01-01

    A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil and nucl......A brief overview of the progress in fuel cell applications and basic technology development is presented, as a backdrop for discussing readiness for penetration into the marketplace as a solution to problems of depletion, safety, climate or environmental impact from currently used fossil...... and nuclear fuel-based energy technologies....

  19. Fuel assemblies

    International Nuclear Information System (INIS)

    Nakatsuka, Masafumi.

    1979-01-01

    Purpose: To prevent scattering of gaseous fission products released from fuel assemblies stored in an fbr type reactor. Constitution; A cap provided with means capable of storing gas is adapted to amount to the assembly handling head, for example, by way of threading in a storage rack of spent fuel assemblies consisting of a bottom plate, a top plate and an assembly support mechanism. By previously eliminating the gas inside of the assembly and the cap in the storage rack, gaseous fission products upon loading, if released from fuel rods during storage, are stored in the cap and do not scatter in the storage rack. (Horiuchi, T.)

  20. Fuel behaviour

    International Nuclear Information System (INIS)

    Fodor, M.; Matus, L.; Vigassy, J.

    1987-11-01

    A short summary of the main critical points in fuel performance of nuclear power reactors from chemical and mechanical point of view is given. A schedule for a limited research program is included. (author) 17 refs

  1. Fuel cells

    International Nuclear Information System (INIS)

    Niederdoeckl, J.

    2001-01-01

    Europe has at present big hopes on the fuel cells technology, in comparison with other energy conversion technologies, this technology has important advantages, for example: high efficiency, very low pollution and parallel use of electric and thermal energy. Preliminary works for fuel cells developing and its commercial exploitation are at full speed; until now the European Union has invested approx. 1.7 billion Schillings, 60 relevant projects are being executed. The Austrian industry is interested in applying this technique to drives, thermal power stations and the miniature fuel cells as replacement of batteries in electronic products (Notebooks, mobile telephones, etc.). A general description of the historic development of fuel cells including the main types is given as well as what is the situation in Austria. (nevyjel)

  2. Fuel assembly

    International Nuclear Information System (INIS)

    Ishibashi, Yoko; Aoyama, Motoo; Oyama, Jun-ichi.

    1995-01-01

    Burnable poison-incorporating fuel rods of a first group are disposed in a region in adjacent with a water rod having a large diameter (neutron moderator rod) disposed to the central portion of a fuel assembly. Burnable poison-incorporating fuel rods of a second group are disposed to a region other than peripheral zone in adjacent with a channel box and corners positioned at an inner zone, in adjacent with the channel box. The average concentration of burnable poisons of the burnable poison-incorporating fuel rods of the first group is made greater than that of the second group. With such a constitution, when the burnable poisons of the first group are burnt out, the burnable poisons of the second group are also burnt out at the same time. Accordingly, an amount of burnable poisons left unburnt at the final stage of the operation cycle is reduced, to improve the reactivity. This can improve the economical property. (I.N.)

  3. Fuel element

    International Nuclear Information System (INIS)

    Armijo, J.S.

    1976-01-01

    A fuel element for nuclear reactors is proposed which has a higher corrosion resisting quality in reactor operations. The zirconium alloy coating around the fuel element (uranium or plutonium compound) has on its inside a protection layer of metal which is metallurgically bound to the substance of the coating. As materials are namned: Alluminium, copper, niobium, stainless steel, and iron. This protective metallic layer has another inner layer, also metallurgically bound to its surface, which consists usually of a zirconium alloy. (UWI) [de

  4. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Sei; Ando, Ryohei; Mitsutake, Toru.

    1995-01-01

    The present invention concerns a fuel assembly suitable to a BWR-type reactor and improved especially with the nuclear characteristic, heat performance, hydraulic performance, dismantling or assembling performance and economical property. A part of poison rods are formed as a large-diameter/multi-region poison rods having a larger diameter than a fuel rod. A large number of fuel rods are disposed surrounding a large diameter water rod and a group of the large-diameter/multi-region poison rods in adjacent with the water rod. The large-diameter water rod has a burnable poison at the tube wall portion. At least a portion of the large-diameter poison rods has a coolant circulation portion allowing coolants to circulate therethrough. Since the large-diameter poison rods are disposed at a position of high neutron fluxes, a large neutron multiplication factor suppression effect can be provided, thereby enabling to reduce the number of burnable poison rods relative to fuels. As a result, power peaking in the fuel assembly is moderated and a greater amount of plutonium can be loaded. In addition the flow of cooling water which tends to gather around the large diameter water rod can be controlled to improve cooling performance of fuels. (N.H.)

  5. A model to predict failure of irradiated U–Mo dispersion fuel

    Energy Technology Data Exchange (ETDEWEB)

    Burkes, Douglas E., E-mail: Douglas.Burkes@pnnl.gov; Senor, David J.; Casella, Andrew M.

    2016-12-15

    Highlights: • Simple model to predict failure of dispersion fuel meat designs. • Evaluated as a function of fabrication parameters and irradiation conditions. • Predictions compare well with experimental measurements of miniature fuel plates. • Interaction layer formation reduces matrix strength and increases temperature. • Si additions to the matrix appear effective only at moderate heat flux and burnup. - Abstract: Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium–molybdenum (U–Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO{sub 2}-stainless steel dispersion fuels and uses currently available thermal–mechanical property information for the materials of interest in the currently proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as onset of pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the {sup 235}U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of

  6. Characterization of intergranular fission gas bubbles in U-Mo fuel

    International Nuclear Information System (INIS)

    Kim, Y. S.; Hofman, G.; Rest, J.; Shevlyakov, G. V.

    2008-01-01

    This report can be divided into two parts: the first part, which is composed of sections 1, 2, and 3, is devoted to report the analyses of fission gas bubbles; the second part, which is in section 4, is allocated to describe the mechanistic model development. Swelling data of irradiated U-Mo alloy typically show that the kinetics of fission gas bubbles is composed of two different rates: lower initially and higher later. The transition corresponds to a burnup of ∼0 at% U-235 (LEU) or a fission density of ∼3 x 10 21 fissions/cm 3 . Scanning electron microscopy (SEM) shows that gas bubbles appear only on the grain boundaries in the pretransition regime. At intermediate burnup where the transition begins, gas bubbles are observed to spread into the intragranular regions. At high burnup, they are uniformly distributed throughout fuel. In highly irradiated U-Mo alloy fuel large-scale gas bubbles form on some fuel particle peripheries. In some cases, these bubbles appear to be interconnected and occupy the interface region between fuel and the aluminum matrix for dispersion fuel, and fuel and cladding for monolithic fuel, respectively. This is a potential performance limit for U-Mo alloy fuel. Microscopic characterization of the evolution of fission gas bubbles is necessary to understand the underlying phenomena of the macroscopic behavior of fission gas swelling that can lead to a counter measure to potential performance limit. The microscopic characterization data, particularly in the pre-transition regime, can also be used in developing a mechanistic model that predicts fission gas bubble behavior as a function of burnup and helps identify critical physical properties for the future tests. Analyses of grain and grain boundary morphology were performed. Optical micrographs and scanning electron micrographs of irradiated fuel from RERTR-1, 2, 3 and 5 tests were used. Micrographic comparisons between as-fabricated and as-irradiated fuel revealed that the site of

  7. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  8. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  9. Fuel assembly

    International Nuclear Information System (INIS)

    Abe, Hideaki; Sakai, Takao; Ishida, Tomio; Yokota, Norikatsu.

    1992-01-01

    The lower ends of a plurality of plate-like shape memory alloys are secured at the periphery of the upper inside of the handling head of a fuel assembly. As the shape memory alloy, a Cu-Zn alloy, a Ti-Pd alloy or a Fe-Ni alloy is used. When high temperature coolants flow out to the handling head, the shape memory alloy deforms by warping to the outer side more greatly toward the upper portion thereof with the temperature increase of the coolants. As the result, the shape of the flow channel of the coolants is changed so as to enlarge at the exit of the upper end of the fuel assembly. Then, the pressure loss of the coolants in the fuel assembly is decreased by the enlargement. Accordingly, the flow rate of the coolants in the fuel assembly is increased to lower the temperature of the coolants. Further, high temperature coolants and low temperature coolants are mixed sufficiently just above the fuel assembly. This can suppress the temperature fluctuation of the mixed coolants in the upper portion of the reactor core, thereby enabling to decrease a fatigue and failures of the structural components in the upper portion of the reactor core. (I.N.)

  10. Fuel assembly

    International Nuclear Information System (INIS)

    Sano, Hiroki; Fushimi, Atsushi; Tominaga, Kenji; Aoyama, Motoo; Ishii, Kazuya.

    1997-01-01

    In burnable poison-incorporated uranium fuels of a BWR type reactor, the compositional ratio of isotopes of the burnable poisons is changed so as to increase the amount of those having a large neutron absorbing cross sectional area. For example, if the ratio of Gd-157 at the same burnable poison enrichment degree is made greater than the natural ratio, this gives the same effect as the increase of the enrichment degree per one fuel rod, thereby providing an effect of reducing a surplus reactivity. Gadolinium, hafnium and europium as burnable poisons have an absorbing cross sectional area being greater in odd numbered nuclei than in even numbered nuclei, on the contrary, boron has a cross section being greater in even numbered nucleus than odd numbered nuclei. Accordingly, if the ratio of isotopes having greater cross section at the same burnable poison enrichment degree is made greater than the natural ratio, surplus reactivity at the initial stage of the burning can be reduced without greatly increasing the amount of burnable poison-incorporated uranium fuels, fuel loading amount is not reduced and the fuel economy is not worsened. (N.H.)

  11. Canadian power reactor fuel

    International Nuclear Information System (INIS)

    Page, R.D.

    1976-03-01

    The following subjects are covered: the basic CANDU fuel design, the history of the bundle design, the significant differences between CANDU and LWR fuel, bundle manufacture, fissile and structural materials and coolants used in the CANDU fuel program, fuel and material behaviour, and performance under irradiation, fuel physics and management, booster rods and reactivity mechanisms, fuel procurement, organization and industry, and fuel costs. (author)

  12. 2010 national progress report on R and D on LEU fuel and target technology in Argentina

    International Nuclear Information System (INIS)

    Balart, S.; Blaumann, H.; Cristini, P.; Gonzalez, A.G.; Gonzalez, R.; Hermida, J.D.; Lopez, M.; Mirandou, M.; Taboada, H.

    2010-01-01

    Since last RRFM meeting, CNEA has deployed several related tasks. The RA-6 MTR type reactor, converted its core from HEU to a new LEU silicide one is scaling up the power, according to a protocol requested by the national regulatory body, ARN. CNEA is deploying an intense R and D activity to fabricate both dispersed U-Mo (Al-Si matrix and Al cladding) and monolithic (Zry-4 cladding) miniplates to develop possible solutions to VHD dispersed and monolithic fuels technical problems. Some monolithic 58% enrichment U8%Mo and U10%Mo are being delivered to INL-DoE to be irradiated in ATR reactor core. A conscientious study on compound interphase formation in both cases is being carried out. CNEA, a worldwide leader on LEU technology for fission radioisotope production is providing Brazil with these radiopharmaceutical products and Egypt and Australia with the technology through INVAP SE. CNEA is also committed to improve the diffusion of LEU target and radiochemical technology for radioisotope production and target and process optimization. Future plans include: 1) Fabrication of a LEU dispersed U-Mo fuel prototype following the recommendations of the IAEA's Good Practices document, to be irradiated in a high flux reactor in the frame of the ARG/4/092 IAEA's Technical Cooperation project. 2) Development of LEU very high density monolithic and dispersed U-Mo fuel plates with Zry-4 or Al cladding as a part of the RERTR program. 3) Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

  13. CANDU fuel

    International Nuclear Information System (INIS)

    MacEwan, J.R.; Notley, M.J.F.; Wood, J.C.; Gacesa, M.

    1982-09-01

    The direction of CANDU fuel development was set in 1957 with the decision to build pressure tube reactors. Short - 50 cm long - rodded bundles of natural UO 2 clad in Zircaloy were adopted to facilitate on-power fuelling to improve uranium utilization. Progressive improvements were made during 25 years of development, involving 650 man years and 180 million dollars. Today's CANDU bundle is based on the knowledge gained from extensive irradiation testing and experience in power reactors. The main thrust of future development is to demonstrate that the present bundle is suitable, with minor modifications, for thorium fuels

  14. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    International Nuclear Information System (INIS)

    1985-07-01

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses

  15. Proceedings of the 1984 international meeting on Reduced Enrichment for Research and Test Reactors. Base technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    More than 40 papers were presented at this RERTR Meeting during the following sessions: Status of RERTR programs and licensing procedures; LEU fuel element development; fuel fabrication and testing; economics; mixed reactor cores; and applications, i.e. neutronics and thermal hydraulics design of upgraded reactors, with new LEU fuel, fuel cycle studies, feasibility and safety analyses.

  16. Fuels characterization studies. [jet fuels

    Science.gov (United States)

    Seng, G. T.; Antoine, A. C.; Flores, F. J.

    1980-01-01

    Current analytical techniques used in the characterization of broadened properties fuels are briefly described. Included are liquid chromatography, gas chromatography, and nuclear magnetic resonance spectroscopy. High performance liquid chromatographic ground-type methods development is being approached from several directions, including aromatic fraction standards development and the elimination of standards through removal or partial removal of the alkene and aromatic fractions or through the use of whole fuel refractive index values. More sensitive methods for alkene determinations using an ultraviolet-visible detector are also being pursued. Some of the more successful gas chromatographic physical property determinations for petroleum derived fuels are the distillation curve (simulated distillation), heat of combustion, hydrogen content, API gravity, viscosity, flash point, and (to a lesser extent) freezing point.

  17. Alternative Fuels Data Center: Ethanol Fueling Stations

    Science.gov (United States)

    ... More in this section... Ethanol Basics Benefits & Considerations Stations Locations Infrastructure fueling stations by location or along a route. Infrastructure Development Learn about ethanol fueling infrastructure; codes, standards, and safety; and ethanol equipment options. Maps & Data E85 Fueling Station

  18. Alternative Fuels Data Center: Biodiesel Fueling Stations

    Science.gov (United States)

    Locations Infrastructure Development Vehicles Laws & Incentives Biodiesel Fueling Stations Photo of a location or along a route. Infrastructure Development Learn about biodiesel fueling infrastructure codes Case Studies California Ramps Up Biofuels Infrastructure Green Fueling Station Powers Fleets in Upstate

  19. Fuels processing for transportation fuel cell systems

    Science.gov (United States)

    Kumar, R.; Ahmed, S.

    Fuel cells primarily use hydrogen as the fuel. This hydrogen must be produced from other fuels such as natural gas or methanol. The fuel processor requirements are affected by the fuel to be converted, the type of fuel cell to be supplied, and the fuel cell application. The conventional fuel processing technology has been reexamined to determine how it must be adapted for use in demanding applications such as transportation. The two major fuel conversion processes are steam reforming and partial oxidation reforming. The former is established practice for stationary applications; the latter offers certain advantages for mobile systems and is presently in various stages of development. This paper discusses these fuel processing technologies and the more recent developments for fuel cell systems used in transportation. The need for new materials in fuels processing, particularly in the area of reforming catalysis and hydrogen purification, is discussed.

  20. Fuel assembly

    International Nuclear Information System (INIS)

    Kurihara, Kunitoshi; Azekura, Kazuo.

    1992-01-01

    In a reactor core of a heavy water moderated light water cooled pressure tube type reactor, no sufficient effects have been obtained for the transfer width to a negative side of void reactivity change in a region of a great void coefficient. Then, a moderation region divided into upper and lower two regions is disposed at the central portion of a fuel assembly. Coolants flown into the lower region can be discharged to the cooling region from an opening disposed at the upper end portion of the lower region. Light water flows from the lower region of the moderator region to the cooling region of the reactor core upper portion, to lower the void coefficient. As a result, the reactivity performance at low void coefficient, i.e., a void reaction rate is transferred to the negative side. Thus, this flattens the power distribution in the fuel assembly, increases the thermal margin and enables rapid operaiton and control of the reactor core, as well as contributes to the increase of fuel burnup ratio and reduction of the fuel cycle cost. (N.H.)

  1. Fuel assembly

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Shimada, Hidemitsu; Aoyama, Motoo; Nakajima, Junjiro

    1998-01-01

    In a fuel assembly for an n x n lattice-like BWR type reactor, n is determined to 9 or greater, and the enrichment degree of plutonium is determined to 4.4% by weight or less. Alternatively, n is determined to 10 or greater, and the enrichment degree of plutonium is determined to 5.2% by weight or less. An average take-out burnup degree is determined to 39GWd/t or less, and the matrix is determined to 9 x 9 or more, or the average take-out burnup degree is determined to 51GWd/t, and the matrix is determined to 10 x 10 or more and the increase of the margin of the maximum power density obtained thereby is utilized for the compensation of the increase of distortion of power distribution due to decrease of the kinds of plutonium enrichment degree, thereby enabling to reduce the kind of the enrichment degree of MOX fuel rods to one. As a result, the manufacturing step for fuel pellets can be simplified to reduce the manufacturing cost for MOX fuel assemblies. (N.H.)

  2. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  3. Transport fuel

    DEFF Research Database (Denmark)

    Ronsse, Frederik; Jørgensen, Henning; Schüßler, Ingmar

    2014-01-01

    Worldwide, the use of transport fuel derived from biomass increased four-fold between 2003 and 2012. Mainly based on food resources, these conventional biofuels did not achieve the expected emission savings and contributed to higher prices for food commod - ities, especially maize and oilseeds...

  4. Thorium fuel cycle management

    International Nuclear Information System (INIS)

    Zajac, R.; Darilek, P.; Breza, J.; Necas, V.

    2010-01-01

    In this presentation author deals with the thorium fuel cycle management. Description of the thorium fuels and thorium fuel cycle benefits and challenges as well as thorium fuel calculations performed by the computer code HELIOS are presented.

  5. Repairing fuel for reinsertion

    International Nuclear Information System (INIS)

    Krukshenk, A.

    1986-01-01

    Eqiupment for nuclear reactor fuel assembly repairing produced by Westinghouse and Brawn Bovery companies is described. Repair of failed fuel assemblies replacement of defect fuel elements gives a noticeable economical effect. Thus if the cost of a new fuel assembly is 450-500 thousand dollars, the replacement of one fuel element in it costs approximately 40-60 thousand dollars. In simple cases repairing includes either removal of failed fuel elements from a fuel assembly and its reinsertion with the rest of fuel elements into the reactor core (reactor refueling), or replacement of unfailed fuel elements from one fuel assembly to a new one (fuel assembly overhaul and reconditioning)

  6. Nuclear power fuel cycle

    International Nuclear Information System (INIS)

    Havelka, S.; Jakesova, L.

    1982-01-01

    Economic problems are discussed of the fuel cycle (cost of the individual parts of the fuel cycle and the share of the fuel cycle in the price of 1 kWh), the technological problems of the fuel cycle (uranium ore mining and processing, uranium isotope enrichment, the manufacture of fuel elements, the building of long-term storage sites for spent fuel, spent fuel reprocessing, liquid and gaseous waste processing), and the ecologic aspects of the fuel cycle. (H.S.)

  7. A Prediction Study on Oxidation of Aluminum Alloy Cladding of U{sub 3}Si{sub 2}-Al Fuel Plate

    Energy Technology Data Exchange (ETDEWEB)

    Tahk, Y.W.; Lee, B.H.; Oh, J.Y.; Park, J.H.; Yim, J.S. [Research Reactor Design and Engineering Div., Korea Atomic Energy Research Institute, 1045 Daedeokdaero, Yuseong, Daejeon 305-353 (Korea, Republic of)

    2011-07-01

    U{sub 3}Si{sub 2}-Al dispersion fuel with aluminum alloy cladding will be used for the Jordan Research and Training Reactor (JRTR). Aluminum alloy cladding undergoes corrosion at slow rates under operational status. This causes thinning of the cladding walls and impairs heat transfer to the coolant. Predictions of the aluminum oxide thickness of the fuel cladding and the maximum temperature difference across the oxide film are needed for reliability evaluation based on the design criteria and limits which prohibit spallation of oxide film. In this work, several oxide thickness prediction models were compared with the measured data of in-pile test results from RERTR program. Moreover, specific parametric studies and a preliminary prediction of the aluminum alloy oxidation using the latest model were performed for JRTR fuel. According to the current JRTR fuel management scheme and operation strategy for 5 MW power, fresh fuel is discharged after 900 effective full power days (EFPD), which is too long a span to predict oxidation properly without an elaborate model. The latest model developed by Kim et al. is in good agreement with the recent in-pile test data as well as with the out-of-pile test data available in the literature, and is one of the best predictors for the oxidation of aluminum alloy cladding in various operating condition. Accordingly, this model was chosen for estimating the oxide film thickness. Through the preliminarily evaluation, water pH level is to be controlled lower than 6.2 for the conservativeness in the case of including the effect of anticipated operational occurrences and the spent fuel residence time in the storage rack after discharging. (author)

  8. Fuel trading

    International Nuclear Information System (INIS)

    2015-01-01

    A first part of this report proposes an overview of trends and predictions. After a synthesis on the sector changes and trends, it indicates and comments the most recent predictions for the consumption of refined oil products and for the turnover of the fuel wholesale market, reports the main highlights concerning the sector's life, and gives a dashboard of the sector activity. The second part proposes the annual report on trends and competition. It presents the main operator profiles and fuel categories, the main determining factors of the activity, the evolution of the sector context between 2005 and 2015 (consumptions, prices, temperature evolution). It analyses the evolution of the sector activity and indicators (sales, turnovers, prices, imports). Financial performances of enterprises are presented. The economic structure of the sector is described (evolution of the economic fabric, structural characteristics, French foreign trade). Actors are then presented and ranked in terms of turnover, of added value, and of result

  9. Fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Makoto.

    1991-01-01

    In a fuel assembly in which spectral shift type moderator guide members are arranged, the moderator guide member has a flow channel resistance member, that provides flow resistance against the moderators, in the upstream of a moderator flowing channel, by which the ratio of removing coolants is set greater at the upstream than downstream. With such a constitution, the void distribution increasing upward in the channel box except for the portion of the moderator guide member is moderated by the increase of the area of the void region that expands downward in the guide member. Accordingly, the axial power distribution is flattened throughout the operation cycle and excess distortion is eliminated to improve the fuel integrity. (T.M.)

  10. Fuel element

    International Nuclear Information System (INIS)

    Hirose, Yasuo.

    1982-01-01

    Purpose: To increase the plenum space in a fuel element used for a liquid metal cooled reactor. Constitution: A fuel pellet is secured at one end with an end plug and at the other with a coil spring in a tubular container. A mechanism for fixing the coil spring composed of a tubular unit is mounted by friction with the inner surface of the tubular container. Accordingly, the recoiling force of the coil spring can be retained by fixing mechanism with a small volume, and since a large amount of plenum space can be obtained, the internal pressure rise in the cladding tube can be suppressed even if large quantities of fission products are discharged. (Kamimura, M.)

  11. Fuel assembly

    International Nuclear Information System (INIS)

    Kawai, Mitsuo.

    1988-01-01

    Purpose: To reduce the corrosion rate and suppress the increase of radioactive corrosion products in reactor water of nuclear fuel assemblies for use in BWR type reactors having spacer springs made of nickel based deposition reinforced type alloys. Constitution: Spacer rings made of nickel based deposition reinforced type alloy are incorporated and used as fuel assemblies after applying treatment of dipping and maintaining at high temperature water followed by heating in steams. Since this can remove the nickel leaching into reactor water at the initial stage, Co-58 as the radioactive corrosion products in the reactor water can be reduced, and the operation at in-service inspection or repairement can be facilitated to improve the working efficiency of the nuclear power plant. The dipping time is desirably more than 10 hours and more desirably more than 30 hours. (Horiuchi, T. )

  12. Fuel assemblies

    International Nuclear Information System (INIS)

    Yoshioka, Ritsuo.

    1983-01-01

    Purpose: To improve the operation performance of a BWR type reactor by improving the distribution of the uranium enrichment and the incorporation amount of burnable poisons in fuel assemblies. Constitution: The average enrichment of uranium 235 is increased in the upper portion as compared with that in the lower portion, while the incorporation amount of burnable poisons is increased in an upper portion as compared with that in the lower portion. The difference in the incorporation amount of the burnable poisons between the upper and lower portions is attained by charging two kinds of fuel rods; the ones incorporated with the burnable poisons over the entire length and the others incorporated with the burnable poisons only in the upper portions. (Seki, T.)

  13. Fuel assembly

    International Nuclear Information System (INIS)

    Hirukawa, Koji; Sakurada, Koichi.

    1992-01-01

    In a fuel assembly for a BWR type reactor, water rods or water crosses are disposed between fuel rods, and a value with a spring is disposed at the top of the coolant flow channel thereof, which opens a discharge port when pressure is increased to greater than a predetermined value. Further, a control element for the amount of coolant flow rate is inserted retractable to a control element guide tube formed at the lower portion of the water rod or the water cross. When the amount of control elements inserted to the control element guide tube is small and the inflown coolant flow rate is great, the void coefficient at the inside of the water rod is less than 5%. On the other hand, when the control elements are inserted, the flow resistance is increased, so that the void coefficient in the water rod is greater than 80%. When the pressure in the water rod is increased, the valve with the spring is raised to escape water or steams. Then, since the variation range of the change of the void coefficient can be controlled reliably by the amount of the control elements inserted, and nuclear fuel materials can be utilized effectively. (N.H.)

  14. Solid TRU fuels and fuel cycle technology

    International Nuclear Information System (INIS)

    Ogawa, Toru; Suzuki, Yasufumi

    1997-01-01

    Alloys and nitrides are candidate solid fuels for transmutation. However, the nitride fuels are preferred to the alloys because they have more favorable thermal properties which allows to apply a cold-fuel concept. The nitride fuel cycle technology is briefly presented

  15. Used fuel packing plant for CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Menzies, I.; Thayer, B.; Bains, N., E-mail: imenzies@atsautomation.com [ATS Automation, Cambridge, ON (Canada); Murchison, A., E-mail: amurchison@nwmo.ca [NWMO, Toronto, ON (Canada)

    2015-07-01

    Large forgings have been selected to containerize Light Water Reactor used nuclear fuel. CANDU fuel, which is significantly smaller in size, allows novel approaches for containerization. For example, by utilizing commercially available extruded ASME pipe a conceptual design of a Used Fuel Packing Plant for containerization of used CANDU fuel in a long lived metallic container has been developed. The design adopts a modular approach with multiple independent work cells to transfer and containerize the used fuel. Based on current technologies and concepts from proven industrial systems, the Used Fuel Packing Plant can assemble twelve used fuel containers per day considering conservative levels of process availability. (author)

  16. A continuing success - The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Mustin, Tracy P.; Clapper, Maureen; Reilly, Jill E.

    2000-01-01

    The United States Department of Energy, in consultation with the Department of State, adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel in May 1996. To date, the Foreign Research Reactor (FRR) Spent Nuclear Fuel (SNF) Acceptance Program, established under this policy, has completed 16 spent fuel shipments. 2,651 material test reactor (MTR) assemblies, one Slowpoke core containing less than 1 kilogram of U.S.-origin enriched uranium, 824 Training, Research, Isotope, General Atomic (TRIGA) rods, and 267 TRIGA pins from research reactors around the world have been shipped to the United States so far under this program. As the FRR SNF Acceptance Program progresses into the fifth year of implementation, a second U.S. cross country shipment has been completed, as well as a second overland truck shipment from Canada. Both the cross country shipment and the Canadian shipment were safely and successfully completed, increasing our knowledge and experience in these types of shipments. In addition, two other shipments were completed since last year's RERTR meeting. Other program activities since the last meeting included: taking pre-emptive steps to avoid license amendment pitfalls/showstoppers for spent fuel casks, publication of a revision to the Record of Decision allowing up to 16 casks per ocean going vessel, and the issuance of a cable to 16 of the 41 eligible countries reminding their governments and the reactor operators that the U.S.-origin uranium in their research reactors may be eligible for return to the United States under the Acceptance Program and urging them to begin discussions on shipping schedules. The FRR SNF program has also supported the Department's implementation of the competitive pricing policy for uranium and resumption of shipments of fresh uranium for fabrication into assemblies for research reactors. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues

  17. Moving into the 21st century - The United States' Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Huizenga, David G.; Mustin, Tracy P.; Saris, Elizabeth C.; Reilly, Jill E.

    1999-01-01

    Since 1996, when the United States Department of Energy and the Department of State jointly adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, twelve shipments totaling 2,985 MTR and TRIGA spent nuclear fuel assemblies from research reactors around the world have been accepted into the United States. These shipments have contained approximately 1.7 metric tons of HEU and 0.6 metric tons of LEU. Foreign research reactor operators played a significant role in this success. A new milestone in the acceptance program occurred during the summer of 1999 with the arrival of TRIGA spent nuclear fuel from Europe through the Charleston Naval Weapons Station via the Savannah River Site to the Idaho National Engineering and Environmental Laboratory. This shipment consisted of five casks of TRIGA spent nuclear fuel from research reactors in Germany, Italy, Slovenia, and Romania. These casks were transported by truck approximately 2,400 miles across the United States (one cask packaged in an ISO container per truck). Drawing upon lessons learned in previous shipments, significant technical, legal, and political challenges were addressed to complete this cross-country shipment. Other program activities since the last RERTR meeting have included: formulation of a methodology to determine the quantity of spent nuclear fuel in a damaged condition that may be transported in a particular cask (containment analysis for transportation casks); publication of clarification of the fee policy; and continued planning for the outyears of the acceptance policy including review of reactors and eligible material quantities. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues to demonstrate success due to the continuing commitment between the United States and the research reactor community to make this program work. We strongly encourage all eligible research reactors to decide as soon as possible to

  18. Nuclear fuel preheating system

    International Nuclear Information System (INIS)

    Andrea, C.

    1975-01-01

    A nuclear reactor new fuel handling system which conveys new fuel from a fuel preparation room into the reactor containment boundary is described. The handling system is provided with a fuel preheating station which is adaptd to heat the new fuel to reactor refueling temperatures in such a way that the fuel is heated from the top down so that fuel element cladding failure due to thermal expansions is avoided. (U.S.)

  19. Fuel element loading system

    International Nuclear Information System (INIS)

    Arya, S.P; s.

    1978-01-01

    A nuclear fuel element loading system is described which conveys a plurality of fuel rods to longitudinal passages in fuel elements. Conveyor means successively position the fuel rods above the longitudinal passages in axial alignment therewith and adapter means guide the fuel rods from the conveyor means into the longitudinal passages. The fuel elements are vibrated to cause the fuel rods to fall into the longitudinal passages through the adapter means

  20. Experiences and Trends of Manufacturing Technology of Advanced Nuclear Fuels

    International Nuclear Information System (INIS)

    2012-08-01

    for sodium cooled fast reactors (SFRs) on the basis of their higher breeding ratio and higher thermal conductivity. The advanced SFR fuels have so far been prepared only on a pilot plant scale in a very few countries, and MA bearing metal, oxide, carbide, nitride and inert matrix fuels are being prepared only on a laboratory scale for property evaluation and irradiation testing. Likewise, thorium based MOX has been manufactured on a pilot plant scale and utilized in power reactors in a few countries in a limited way, but there is no industrial facility for manufacturing these fuels. Until the end of the 1970s, non-power research reactors and their fuels were mostly supplied worldwide by the USA and the former USSR. These reactors used high enriched uranium (HEU) fuel, containing >80% 235 U in the form of aluminium matrix, UAl x or U3O 8 dispersion fuel of low uranium density (1.3-1.7 g/cm 3 ). In 1978, the USA launched the Reduced Enrichment for Research and Test Reactors (RERTR) Programme, with the objective of converting the HEU core to a low enriched uranium (LEU) core without affecting the operation and performance of the reactors. Later, the Russian Federation also joined this programme. As part of this programme, LEU based Al-U 3 Si 2 dispersion fuel with a uranium density of 4.8 g/cm 3 emerged and has been qualified. Efforts are under way to develop advanced fuels such as U-Mo alloy either as Al matrix dispersion fuel with a density of 6-8.5 g/cm 3 or as monolithic fuel with a uranium density of 15-16 g/cm 3 . The IAEA has been fostering information exchange and collaborative R and D among Member States for the development, manufacture and performance evaluation of nuclear fuels for both nuclear power and research reactors, and has published a number of technical reports on nuclear fuels. The purpose of this report is to summarize the current status of nuclear fuel manufacturing technology worldwide for both power reactors and research reactors, and to

  1. Nuclear Safeguards and the International Atomic Energy Agency

    Science.gov (United States)

    1995-04-01

    Hexapar- reactors were originally designed to use HEU, the tite safeguards agreement already in place for cen- RERTR (Reduced Enrichment for Research...See discussion below on the RERTR program.) partite agreement. Such short notice is possible in Some reactors, however, have yet to be converted, Europe...the fuel and compares them to those of 1994.21 alternate policies. The RERTR program has made significant Of the some 22,700 fuel elements slated for

  2. Artificial fuel

    Energy Technology Data Exchange (ETDEWEB)

    Hamon, L L.W.

    1918-08-20

    Lignite, peat, sud, leaf-mold, or shale, or two or more of these raw carbonaceous materials are mixed with cellulose material, such as sawdust, silica, alkali, and tar or pitch, or residues from tar or pitch, or residues from the distillation of oils, and the mixture is molded into blocks. Other carbonaceous materials, such as graphite, anthracite, or coal-dust, coke, breeze, or culm, and mineral substances, such as iron and manganese ores, may be added. A smokeless fuel can be obtained by coking the blocks in the usual way in retorts.

  3. Nuclear liability and research reactor fuel. A plant supplier's view

    International Nuclear Information System (INIS)

    Roegler, H.-J.; Hetzmann, A.

    2000-01-01

    compensation than of the operator's country. As there are many potential conflicts around the issue and as there is the need to protect the supplier from, there has been established already early in nuclear history an international regime on how to deal with such claims: 1960 - the Paris Convention (amended 1964/1982) plus the 1963 - the Brussels Supplementary Convention; 1963 - the Vienna Convention; 1988 - the Joint Protocol linking the application of the Vienna and the Paris Convention. But as many states have joined neither the Vienna nor the Paris Convention (nor the Joint Protocol), there have often been established bilateral agreements on how to deal with the nuclear liability in terms of a specific project. The main issue with such bilateral agreements is the backing by that state the owner of the plant belongs to for the owner's obligation out of the nuclear liability. Often this cannot be reached due to the need of a state's decree or modification of the law for that very purpose. This is refused sometimes as well due to lack of insight in its need for a nuclear facility as small as a research reactor. All the above issues have to be taken into considerations during the contract negotiation between the plant supplier and the plant owner when contracting for its construction. All the issues may influence also whether the plant supplier find subsuppliers which accept or can live with the results of such negotiations for their subsupplies. It is inherent that the fuel suppliers are especially keen in solutions which indemnify them from any such risk out of nuclear events; so are organisations which perform safety evaluations. The fuel suppliers especially may be the more interested in high protection the more they are urged to deliver advanced fuel with less tests prior to application, e.g. as consequence of the steady announcement of extremely ambitious development progress by the RERTR program. This contribution to the RERTR-meeting will detail and give examples in

  4. Status report on the irradiation testing and post-irradiation examination of low-enriched U3O8-Al and UAlsub(x)-Al fuel element by the Netherlands Energy Research Foundation (ECN)

    International Nuclear Information System (INIS)

    Pruimboom, H.; Lijbrink, E.; Otterdijk, K. von; Swanenburg de Veye, R.J.

    1984-01-01

    Within the framework of the RERTR-programme four low-enriched (20%) MTR-type fuel elements have been irradiated in the High Flux Reactor at Petten (The Netherlands) and are presently subjected to postirradiation examination. Two of the elements contain UAlsub(x)-Al and two contain U 3 O 8 -Al fuel. The test irradiation has been completed up to the target burn-up values of 50% and 75% respectively. An extensive surveillance programme carried out during the test period has confirmed the excellent in-reactor behaviour of both types. Post-irradiation examination of the 50% burn-up test elements, comprising of dimensional measurements, burn-up determination, fuel metallography and blister testing, has sofar confirmed the irradiation experiences. Good agreement between calculated and measured power and burn-up characteristics has been found. A survey of the test element characteristics, their irradiation history, the irradiation tests and the preliminary PIE results is given in the paper. (author)

  5. Nuclear Fuel elements

    International Nuclear Information System (INIS)

    Hirakawa, Hiromasa.

    1979-01-01

    Purpose: To reduce the stress gradient resulted in the fuel can in fuel rods adapted to control the axial power distribution by the combination of fuel pellets having different linear power densities. Constitution: In a fuel rod comprising a first fuel pellet of a relatively low linear power density and a second fuel pellet of a relatively high linear power density, the second fuel pellet is cut at its both end faces by an amount corresponding to the heat expansion of the pellet due to the difference in the linear power density to the adjacent first fuel pellet. Thus, the second fuel pellet takes a smaller space than the first fuel pellet in the fuel can. This can reduce the stress produced in the portion of the fuel can corresponding to the boundary between the adjacent fuel pellets. (Kawakami, Y.)

  6. Fuel assembly

    International Nuclear Information System (INIS)

    Hiraiwa, Koji; Ueda, Makoto

    1989-01-01

    In a fuel assembly used for a light water cooled reactor such as a BWR type reactor, a water rod is divided axially into an upper outer tube and a lower outer tube by means of a plug disposed from the lower end of a water rod to a position 1/4 - 1/2 of the entire length for the water rod. Inlet apertures and exit apertures for moderators are respectively perforated for the divided outer tube and upper and lower portions. Further, an upper inner tube with less neutron irradiation growing amount than the outer tube is perforated on the plug in the outer tube, while a lower inner tube with greater neutron irradiation growing amount than the outer tube is suspended from the lower surface of the plug in the outer tube. Then, the opening area for the exit apertures disposed to the upper outer tube and the lower outer tube is controlled depending on the difference of the neutron irradiation growing amount between the upper inner tube and the upper outer tube, and the difference of the neutron irradiation growing amount between the lower inner tube and the lower outer tube. This enables effective spectral shift operation and improve the fuel economy. (T.M.)

  7. Fuel Burn Estimation Model

    Science.gov (United States)

    Chatterji, Gano

    2011-01-01

    Conclusions: Validated the fuel estimation procedure using flight test data. A good fuel model can be created if weight and fuel data are available. Error in assumed takeoff weight results in similar amount of error in the fuel estimate. Fuel estimation error bounds can be determined.

  8. Constant strength fuel-fuel cell

    International Nuclear Information System (INIS)

    Vaseen, V.A.

    1980-01-01

    A fuel cell is an electrochemical apparatus composed of both a nonconsumable anode and cathode; and electrolyte, fuel oxidant and controls. This invention guarantees the constant transfer of hydrogen atoms and their respective electrons, thus a constant flow of power by submergence of the negative electrode in a constant strength hydrogen furnishing fuel; when said fuel is an aqueous absorbed hydrocarbon, such as and similar to ethanol or methnol. The objective is accomplished by recirculation of the liquid fuel, as depleted in the cell through specific type membranes which pass water molecules and reject the fuel molecules; thus concentrating them for recycle use

  9. KMRR fuel design

    International Nuclear Information System (INIS)

    Son, D.S.; Sim, B.S.; Kim, T.R.; Hwang, W.; Kim, B.G.; Ku, Y.H.; Lee, C.B.; Lim, I.C.

    1992-06-01

    KMRR fuel rod design criteria on fuel swelling, blistering and oxide spallation have been reexamined. Fuel centerline temperature limit of 250deg C in normal operation condition and fuel swelling limit of 12 % at the end of life have been proposed to prevent fuel failure due to excessive fuel swelling. Fuel temperature limit of 485deg C has been proposed to exclude the possibility of fuel failures during transients or under accident condition. Further analyses are needed to decide the fuel cladding temperature limit to preclude the oxide spallation. Design changes in fuel assembly structure and their effects on related systems have been reviewed from a structural integrity viewpoint. The remained works in fuel mechanical design area have been identified and further efforts of fuel design group will be focused on these aspects. (Author)

  10. Fuel Property Blend Model

    Energy Technology Data Exchange (ETDEWEB)

    Pitz, William J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Mehl, Marco [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wagnon, Scott J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Zhang, Kuiwen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kukkadapu, Goutham [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Westbrook, Charles K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-01-12

    The object of this project is to develop chemical models and associated correlations to predict the blending behavior of bio-derived fuels when mixed with conventional fuels like gasoline and diesel fuels.

  11. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    The Air Base Technologies Division of the Air Force Research Laboratory has developed a logistic fuel processor that removes the sulfur content of the fuel and in the process converts logistic fuel...

  12. Fuel pellet loading apparatus

    International Nuclear Information System (INIS)

    1980-01-01

    Apparatus is described for loading a predetermined amount of nuclear fuel pellets into nuclear fuel elements and particularly for the automatic loading of fuel pellets from within a sealed compartment. (author)

  13. Fuel assembly

    International Nuclear Information System (INIS)

    Wataumi, Kazutoshi; Tajiri, Hiroshi.

    1992-01-01

    In a fuel assembly of a BWR type reactor, a pellet to be loaded comprises an external layer of fissile materials containing burnable poisons and an internal layer of fissile materials not containing burnable poison. For example, there is provided a dual type pellet comprising an external layer made of UO 2 incorporated with Gd 2 O 3 at a predetermined concentration as the burnable poisons and an internal layer made of UO 2 not containing Gd 2 O 3 . The amount of the burnable poisons required for predetermined places is controlled by the thickness of the ring of the external layer. This can dissipate an unnecessary poisoning effect at the final stage of the combustion cycle. Further, since only one or a few kinds of powder mixture of the burnable poisons and the fissile materials is necessary, production and product control can be facilitated. (I.N.)

  14. Fuel storage

    International Nuclear Information System (INIS)

    Palacios, C.; Alvarez-Miranda, A.

    2009-01-01

    ENSA is a well known manufacturer of multi-system primary components for the nuclear industry and is totally prepared to satisfy future market requirements in this industry. At the same time that ENSA has been gaining a reputation world wider for the supply of primary components, has been strengthening its commitment and experience in supplying spent fuel components, either pool racks or storage and transportation casks, and offers not only fabrication but also design capabilities for its products. ENSA has supplied Spent Fuel Pool Racks, in spain, Finland, Taiwan, Korea, China, and currently it is in the process of licensing its own rack design in the United States of America for the ESBWR along with Ge-Hitachi. ENSA has supplied racks for 20 pools and 22 different reactors and it has also manufactured racks under all available technologies and developed a design known as Interlock Cell Matrix whose main features are outlined in this article. Another ENSA achievement in rack technology is the use of remote control for re-racking activities instead of using divers, which improves the ALARA requirements. Regarding casks for storage and transportation, ENSA also has al leading worldwide position, with exports prevailing over the Spanish market where ENSA has supplied 16 storage and transportation casks to the Spanish nuclear power Trillo. In some cases, ENSA acts as subcontractor for other clients. Foreign markets are still a major challenge for ENSA. ENSA-is well known for its manufacturing capabilities in the nuclear industry, but has been always involved in design activities through its engineering division, which carries out different tasks: components Design; Tooling Design; Engineering and Documentation; Project Engineering; Calculations, Design and Development Engineering. (Author)

  15. Nuclear fuel replacement device

    International Nuclear Information System (INIS)

    Ritz, W.C.; Robey, R.M.; Wett, J.F.

    1984-01-01

    A fuel handling arrangement for a liquid metal cooled nuclear reactor having a single rotating plug eccentric to the fuel core and a fuel handling machine radially movable along a slot in the plug with a transfer station disposed outside the fuel core but covered by the eccentric plug and within range of movement of said fuel handling machine to permit transfer of fuel assemblies between the core and the transfer station. (author)

  16. CANDU fuel performance

    International Nuclear Information System (INIS)

    Ivanoff, N.V.; Bazeley, E.G.; Hastings, I.J.

    1982-01-01

    CANDU fuel has operated successfully in Ontario Hydro's power reactors since 1962. In the 19 years of experience, about 99.9% of all fuel bundles have performed as designed. Most defects occurred before 1979 and subsequent changes in fuel design, fuel management, reactor control, and manufacturing quality control have reduced the current defect rate to near zero. Loss of power production due to defective fuel has been negligible. The outstanding performance continues while maintaining a low unit energy cost for fuel

  17. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2001-01-01

    .... The focus during the subject period was directed to understanding the pyrolysis and combustion of endothermic fuels under subcritical conditions and the pyrolysis of these fuels under supercritical conditions...

  18. Fuels Combustion Research: Supercritical Fuel Pyrolysis

    National Research Council Canada - National Science Library

    Glassman, Irvin

    2000-01-01

    .... The focus during the subject period was directed to understanding the pyrolysis and combustion of endothermic fuels under subcritical conditions and the pyrolysis of these fuels under supercritical conditions...

  19. GSPEL - Fuel Cell Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The Fuel Cell Lab (FCL)Established to investigate, integrate, testand verifyperformance and technology readiness offuel cell systems and fuel reformers for use with...

  20. Fuel performance experience

    International Nuclear Information System (INIS)

    Sofer, G.A.

    1986-01-01

    The history of LWR fuel supply has been characterized by a wide range of design developments and fuel cycle cost improvements. Exxon Nuclear Company, Inc. has pursued an aggressive fuel research and development program aimed at improved fuel performance. Exxon Nuclear has introduced many design innovations which have improved fuel cycle economics and operating flexibility while fuel failures remain at very low levels. The removable upper tie plate feature of Exxon Nuclear assemblies has helped accelerate this development, enabling repeated inspections during successive plant outages. Also, this design feature has made it possible to repair damaged fuel assemblies during refueling outages, thereby minimizing the economic impact of fuel failure from all causes

  1. Catalytic Fuel Conversion Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility enables unique catalysis research related to power and energy applications using military jet fuels and alternative fuels. It is equipped with research...

  2. HTGR fuel reprocessing technology

    International Nuclear Information System (INIS)

    Brooks, L.H.; Heath, C.A.; Shefcik, J.J.

    1976-01-01

    The following aspects of HTGR reprocessing technology are discussed: characteristics of HTGR fuels, criteria for a fuel reprocessing flowsheet; selection of a reference reprocessing flowsheet, and waste treatment

  3. Characterization of dispersed type fuel miniplates based in alloy UMo by evaluation of changes volumetrics and thermal conductivity

    International Nuclear Information System (INIS)

    Salinas Valero, Pablo Ignacio

    2016-01-01

    The development of new technologies in the nuclear area is extremely important to achieve greater efficiency and security in the production of electrical energy in the case of power reactors and for the production of radioisotopes and neutrons in research reactors. Throughout history, uranium-based nuclear fuels evolved in parallel with the requirements of nuclear reactors, this emphasis was increased when the RERTR program was created, which restricts the use of fuels with a maximum enrichment of 20% of the isotope U 235 (fissile isotope), which makes it necessary to increase the mass of uranium to compensate the amount of fissile material to maintain a neutron flux necessary for the reactors to operate with the same power. The search for new nuclear fuels has reached the UMo alloy with which densities of 18 gU/cm 3 are achieved in type fuels and 8 gU/cm 3 in dispersed type fuels, properties under irradiation due to their cubic crystalline structure. This type of fuel, when used dispersed in an aluminum matrix, becomes thermodynamically unstable by increasing the fission temperature of the U 235 isotope, due to this, compounds of lower density are formed, which causes an increase in volume (swelling). ). This swelling is studied throughout the present work, to relate the changes of UMo-Al / 4% volume of thermally induced miniecography in thermal treatments, with the purpose of evaluating changes in the thermal conductivity of the material. In this study it was detected that the swelling in miniplates is related in some way to the reduction of thermal conductivity, it was also recorded that the volume of change is irregular increasing and decreasing its volume according to the hours of induced swelling. The purpose of this work is to contribute to the development of dispersed fuels based on the UMo alloy in order to control the variables and reduce the probability of faults and possible accidents, such as fission products, or an increase in temperature in the core

  4. Nuclear fuel production

    International Nuclear Information System (INIS)

    Randol, A.G.

    1985-01-01

    The production of new fuel for a power plant reactor and its disposition following discharge from the power plant is usually referred to as the ''nuclear fuel cycle.'' The processing of fuel is cyclic in nature since sometime during a power plant's operation old or ''depleted'' fuel must be removed and new fuel inserted. For light water reactors this step typically occurs once every 12-18 months. Since the time required for mining of the raw ore to recovery of reusable fuel materials from discharged materials can span up to 8 years, the management of fuel to assure continuous power plant operation requires simultaneous handling of various aspects of several fuel cycles, for example, material is being mined for fuel to be inserted in a power plant 2 years into the future at the same time fuel is being reprocessed from a discharge 5 years prior. Important aspects of each step in the fuel production process are discussed

  5. Fuel manufacturing and utilization

    International Nuclear Information System (INIS)

    2005-01-01

    The efficient utilisation of nuclear fuel requires manufacturing facilities capable of making advanced fuel types, with appropriate quality control. Once made, the use of such fuels requires a proper understanding of their behaviour in the reactor environment, so that safe operation for the design life can be achieved. The International Atomic Energy Agency supports Member States to improve in-pile fuel performance and management of materials; and to develop advanced fuel technologies for ensuring reliability and economic efficiency of the nuclear fuel cycle. It provides assistance to Member States to support fuel-manufacturing capability, including quality assurance techniques, optimization of manufacturing parameters and radiation protection. The IAEA supports the development fuel modelling expertise in Member States, covering both normal operation and postulated and severe accident conditions. It provides information and support for the operation of Nuclear Power Plant to ensure that the environment and water chemistry is appropriate for fuel operation. The IAEA supports fuel failure investigations, including equipment for failed fuel detection and for post-irradiation examination and inspection, as well as fuel repair, it provides information and support research into the basic properties of fuel materials, including UO 2 , MOX and zirconium alloys. It further offers guidance on the relationship with back-end requirement (interim storage, transport, reprocessing, disposal), fuel utilization and management, MOX fuels, alternative fuels and advanced fuel technology

  6. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1993-01-01

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have led to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  7. Instrumentation of fuel elements and fuel plates

    International Nuclear Information System (INIS)

    Durand, J.P.; Fanjas, Y.

    1994-01-01

    When controlling the behaviour of a reactor or developing a new fuel concept, it is of utmost interest to have the possibility to confirm the thermohydraulic calculations by actual measurements in the fuel elements or in the fuel plates. For years, CERCA has developed the technology and supplied its customers with fuel elements equipped with pressure or temperature measuring devices according to the requirements. Recent customer projects have lead to the development of a new method to introduce thermocouples directly into the fuel plate meat instead of the cladding. The purpose of this paper is to review the various instrumentation possibilities available at CERCA. (author)

  8. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Coobs, J.H.

    1976-08-01

    The status of fuel and fuel cycle technology for high-temperature gas-cooled reactors (HTGRs) is reviewed. The all-ceramic core of the HTGRs permits high temperatures compared with other reactors. Core outlet temperatures of 740 0 C are now available for the steam cycle. For advanced HTGRs such as are required for direct-cycle power generation and for high-temperature process heat, coolant temperatures as high as 1000 0 C may be expected. The paper discusses the variations of HTGR fuel designs that meet the performance requirements and the requirements of the isotopes to be used in the fuel cycle. Also discussed are the fuel cycle possibilities, which include the low-enrichment cycle, the Th- 233 U cycle, and plutonium utilization in either cycle. The status of fuel and fuel cycle development is summarized

  9. HTGR fuel and fuel cycle technology

    International Nuclear Information System (INIS)

    Lotts, A.L.; Homan, F.J.; Balthesen, E.; Turner, R.F.

    1977-01-01

    Significant advances have occurred in the development of HTGR fuel and fuel cycle. These accomplishments permit a wide choice of fuel designs, reactor concepts, and fuel cycles. Fuels capable of providing helium outlet temperatures of 750 0 C are available, and fuels capable of 1000 0 C outlet temperatures may be expected from extension of present technology. Fuels have been developed for two basic HTGR designs, one using a spherical (pebble bed) element and the other a prismatic element. Within each concept a number of variations of geometry, fuel composition, and structural materials are permitted. Potential fuel cycles include both low-enriched and high-enriched Th- 235 U, recycle Th- 233 U, and Th-Pu or U-Pu cycles. This flexibility offered by the HTGR is of great practical benefit considering the rapidly changing economics of power production. The inflation of ore prices has increased optimum conversion ratios, and increased the necessity of fuel recycle at an early date. Fuel element makeup is very similar for prismatic and spherical designs. Both use spherical fissile and fertile particles coated with combinations of pyrolytic carbon and silicon carbide. Both use carbonaceous binder materials, and graphite as the structural material. Weak-acid resin (WAR) UO 2 -UC 2 fissile fuels and sol-gel-derived ThO 2 fertile fuels have been selected for the Th- 233 U cycle in the prismatic design. Sol-gel-derived UO 2 UC 2 is the reference fissile fuel for the low-enriched pebble bed design. Both the United States and Federal Republic of Germany are developing technology for fuel cycle operations including fabrication, reprocessing, refabrication, and waste handling. Feasibility of basic processes has been established and designs developed for full-scale equipment. Fuel and fuel cycle technology provide the basis for a broad range of applications of the HTGR. Extension of the fuels to higher operating temperatures and development and commercial demonstration of fuel

  10. Treat upgrade fuel fabrication

    International Nuclear Information System (INIS)

    Davidson, K.V.; Schell, D.H.

    1979-01-01

    An extrusion and thermal treatment process was developed to produce graphite fuel rods containing a dispersion of enriched UO 2 . These rods will be used in an upgraded version of the Transient Reactor Test Facility (TREAT). The improved fuel provides a higher graphite matrix density, better fuel dispersion and higher thermal capabilities than the existing fuel

  11. Integrated fuel processor development

    International Nuclear Information System (INIS)

    Ahmed, S.; Pereira, C.; Lee, S. H. D.; Krumpelt, M.

    2001-01-01

    The Department of Energy's Office of Advanced Automotive Technologies has been supporting the development of fuel-flexible fuel processors at Argonne National Laboratory. These fuel processors will enable fuel cell vehicles to operate on fuels available through the existing infrastructure. The constraints of on-board space and weight require that these fuel processors be designed to be compact and lightweight, while meeting the performance targets for efficiency and gas quality needed for the fuel cell. This paper discusses the performance of a prototype fuel processor that has been designed and fabricated to operate with liquid fuels, such as gasoline, ethanol, methanol, etc. Rated for a capacity of 10 kWe (one-fifth of that needed for a car), the prototype fuel processor integrates the unit operations (vaporization, heat exchange, etc.) and processes (reforming, water-gas shift, preferential oxidation reactions, etc.) necessary to produce the hydrogen-rich gas (reformate) that will fuel the polymer electrolyte fuel cell stacks. The fuel processor work is being complemented by analytical and fundamental research. With the ultimate objective of meeting on-board fuel processor goals, these studies include: modeling fuel cell systems to identify design and operating features; evaluating alternative fuel processing options; and developing appropriate catalysts and materials. Issues and outstanding challenges that need to be overcome in order to develop practical, on-board devices are discussed

  12. Methanol Fuel Cell

    Science.gov (United States)

    Voecks, G. E.

    1985-01-01

    In proposed fuel-cell system, methanol converted to hydrogen in two places. External fuel processor converts only part of methanol. Remaining methanol converted in fuel cell itself, in reaction at anode. As result, size of fuel processor reduced, system efficiency increased, and cost lowered.

  13. Reactor fueling system

    International Nuclear Information System (INIS)

    Hattori, Noriaki; Hirano, Haruyoshi.

    1983-01-01

    Purpose: To optimally position a fuel catcher by mounting a television camera to a fuel catching portion and judging video images by the use of a computer or the like. Constitution: A television camera is mounted to the lower end of a fuel catching mechanism for handling nuclear fuels and a fuel assembly disposed within a reactor core or a fuel storage pool is observed directly from above to judge the position for the fuel assembly by means of video signals. Then, the relative deviation between the actual position of the fuel catcher and that set in a memory device is determined and the positional correction is carried out automatically so as to reduce the determined deviation to zero. This enables to catch the fuel assembly without failure and improves the efficiency for the fuel exchange operation. (Moriyama, K.)

  14. Nuclear fuel elements

    International Nuclear Information System (INIS)

    Nakai, Keiichi

    1983-01-01

    Purpose: To decrease the tensile stresses resulted in a fuel can as well as prevent decladding of fuel pellets into the bore holes by decreasing the inner pressure within the nuclear fuel element. Constitution: A fuel can is filled with hollow fuel pellets, inserted with a spring for retaining the hollow fuel pellets with an appropriate force and, thereafter, closely sealed at the both ends with end plugs. A cylindrical body is disposed into the bore holes of the hollow fuel pellets. Since initial sealing gases and/or gaseous nuclear fission products can thus be excluded from the bore holes where the temperature is at the highest level, the inner pressure of the nuclear fuel element can be reduced to decrease the tensile strength resulted to the fuel can. Furthermore, decladding of fuel pellets into the bore holes can be prevented. (Moriyama, K.)

  15. Failed fuel rod detector

    Energy Technology Data Exchange (ETDEWEB)

    Uchida, Katsuya; Matsuda, Yasuhiko

    1984-05-02

    The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.

  16. 77 FR 13009 - Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel...

    Science.gov (United States)

    2012-03-05

    ... Regulation of Fuels and Fuel Additives: Identification of Additional Qualifying Renewable Fuel Pathways Under the Renewable Fuel Standard Program AGENCY: Environmental Protection Agency (EPA). ACTION: Withdrawal... Renewable Fuel Standard program regulations. Because EPA received adverse comment, we are withdrawing the...

  17. Reduced enrichment for research and test reactors: proceedings

    International Nuclear Information System (INIS)

    1985-07-01

    Separate abstracts are presented for each of the papers included in the data base concerning RERTR programs and licensing; fuel development; plate-type fuel fabrication; fuel demonstration; economics; mixed cores; and applications

  18. Materials for fuel cells

    OpenAIRE

    Haile, Sossina M

    2003-01-01

    Because of their potential to reduce the environmental impact and geopolitical consequences of the use of fossil fuels, fuel cells have emerged as tantalizing alternatives to combustion engines. Like a combustion engine, a fuel cell uses some sort of chemical fuel as its energy source but, like a battery, the chemical energy is directly converted to electrical energy, without an often messy and relatively inefficient combustion step. In addition to high efficiency and low emissions, fuel cell...

  19. Advanced fuels safety comparisons

    International Nuclear Information System (INIS)

    Grolmes, M.A.

    1977-01-01

    The safety considerations of advanced fuels are described relative to the present understanding of the safety of oxide fueled Liquid Metal Fast Breeder Reactors (LMFBR). Safety considerations important for the successful implementation of advanced fueled reactors must early on focus on the accident energetics issues of fuel coolant interactions and recriticality associated with core disruptive accidents. It is in these areas where the thermal physical property differences of the advanced fuel have the greatest significance

  20. Note on current position regarding the development by the UKAEA of Reduced Enrichment fuels for Research and Test Reactors

    International Nuclear Information System (INIS)

    Hickey, B.

    1983-01-01

    The United Kingdom Atomic Energy Authority have an MTR fuel fabrication plant located at Dounreay on the north coast of Scotland. The prime function of the plant is to manufacture fuel elements for the UKAEA's own DIDO and PLUTO heavy water reactors located at their research establishment at Harwell. The plant, which has a capacity of about 1000 fuel elements per annum, also manufactures fuel elements, on a commercial basis, for university reactors in the United Kingdom and for a number of customers in overseas countries. The UKAEA have been manufacturing MTR fuel elements of a wide range of designs for over twenty-five years. Following the initiative of the US Government's RERTR programme, the UKAEA have embarked on a modest programme of MTR fuel manufacturing development., irradiation and post-irradiation examination to establish the techniques required to manufacture fuel elements containing uranium of a significantly lower enrichment than that in the fuel elements they currently manufacture. In the first instance this work is being directed towards the production of fuel elements containing uranium of 45% enrichment. After an initial analysis it was recognised that although a satisfactory 45% enriched version of certain of the designs of fuel elements currently manufactured could probably be produced using established U/Al alloy technology, it would be necessary to utilise powder technology for other elements in order to achieve the higher uranium density required. Studies of published information and consideration of the technology and facilities already available at Dounreay prompted the decision to concentrate on the development Of U 3 O 8 /Al cermet type fuel elements of similar geometry to those currently manufactured. Some of the fuel element designs currently manufactured by the UKAEA are listed: Concentric (Extruded) 74% enriched; Concentric Plates 80% enriched with densities 0.60 and 0.53 g U/ cm 3 ; Flat Plate (Swaged) 80% enriched and Flat Plate

  1. The Proliferation Security Initiative: A Means to an End for the Operational Commander

    Science.gov (United States)

    2009-05-04

    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. The RERTR Program was initiated by the U.S. Department of...processes have been developed for producing radioisotopes with LEU targets. The RERTR Program is managed by the Office of Nuclear Material Threat

  2. Nuclear fuel storage facility

    International Nuclear Information System (INIS)

    Matsumoto, Takashi; Isaka, Shinji.

    1987-01-01

    Purpose: To increase the spent fuel storage capacity and reduce the installation cost in a nuclear fuel storage facility. Constitution: Fuels handled in the nuclear fuel storage device of the present invention include the following four types: (1) fresh fuels, (2) 100 % reactor core charged fuels, (3) spent fuels just after taking out and (4) fuels after a certain period (for example one half-year) from taking out of the reactor. Reactivity is high for the fuels (1), and some of fuels (2), while low in the fuels (3) (4), Source intensity is strong for the fuels (3) and some of the fuels (2), while it is low for the fuels (1) and (4). Taking notice of the fact that the reactivity, radioactive source intensity and generated after heat are different in the respective fuels, the size of the pool and the storage capacity are increased by the divided storage control. While on the other hand, since the division is made in one identical pool, the control method becomes important, and the working range is restricted by means of a template, interlock, etc., the operation mode of the handling machine is divided into four, etc. for preventing errors. (Kamimura, M.)

  3. Fuel pattern recognition device

    International Nuclear Information System (INIS)

    Sato, Tomomi.

    1995-01-01

    The device of the present invention monitors normal fuel exchange upon fuel exchanging operation carried out in a reactor of a nuclear power plant. Namely, a fuel exchanger is movably disposed to the upper portion of the reactor and exchanges fuels. An exclusive computer receives operation signals of the fuel exchanger during operation as inputs, and outputs reactor core fuel pattern information signals to a fuel arrangement diagnosis device. An underwater television camera outputs image signals of a fuel pattern in the reactor core to an image processing device. If there is any change in the image signals for the fuel pattern as a result of the fuel exchange operation of the fuel exchanger, the image processing device outputs the change as image signals to the fuel pattern diagnosis device. The fuel pattern diagnosis device compares the pattern information signals from the exclusive computer with the image signals from the image processing device, to diagnose the result of the fuel exchange operation performed by the fuel exchanger and inform the diagnosis by means of an image display. (I.S.)

  4. BWR fuel performance

    International Nuclear Information System (INIS)

    Baily, W.E.; Armijo, J.S.; Jacobson, J.; Proebstle, R.A.

    1979-01-01

    The General Electric experience base on BWR fuel includes over 29,000 fuel assemblies which contain 1,600,000 fuel rods. Over the last five years, design, process and operating changes have been introduced which have had major effects in improving fuel performance. Monitoring this fuel performance in BWRs has been accomplished through cooperative programs between GE and utilities. Activities such as plant fission product monitoring, fuel sipping and fuel and channel surveillance programs have jointly contributed to the value of this extensive experience base. The systematic evaluation of this data has established well-defined fuel performance trends which provide the assurance and confidence in fuel reliability that only actual operating experience can provide

  5. Dual Tank Fuel System

    Science.gov (United States)

    Wagner, Richard William; Burkhard, James Frank; Dauer, Kenneth John

    1999-11-16

    A dual tank fuel system has primary and secondary fuel tanks, with the primary tank including a filler pipe to receive fuel and a discharge line to deliver fuel to an engine, and with a balance pipe interconnecting the primary tank and the secondary tank. The balance pipe opens close to the bottom of each tank to direct fuel from the primary tank to the secondary tank as the primary tank is filled, and to direct fuel from the secondary tank to the primary tank as fuel is discharged from the primary tank through the discharge line. A vent line has branches connected to each tank to direct fuel vapor from the tanks as the tanks are filled, and to admit air to the tanks as fuel is delivered to the engine.

  6. HTGR Fuel performance basis

    International Nuclear Information System (INIS)

    Shamasundar, B.I.; Stansfield, O.M.; Jensen, D.D.

    1982-05-01

    The safety characteristics of the high-temperature gas-cooled reactor (HTGR) during normal and accident conditions are determined in part by HTGR fuel performance. During normal operation, less than 0.1% fuel failure occurs, primarily from defective particles. This low fuel failure fraction limits circulating activity to acceptable levels. During severe accidents, the radiological consequence is influenced by high-temperature fuel particle behavior. An empirical fuel failure model, supported by recent experimental data, is presented. The onset of significant fuel particle failure occurs at temperatures in excess of 1600 0 C, and complete fuel failure occurs at 2660 0 C. This indicates that the fuel is more retentive at higher temperatures than previously assumed. The more retentive nature of the fuel coupled with the high thermal capacitance of the core results in slow release of fission products from the core during severe accidents

  7. Elongated fuel road

    International Nuclear Information System (INIS)

    Williams, A.E.; Linkison, W.S.

    1977-01-01

    A fuel rod is proposed where a reorientation of the fuel in case of a considerable temperature increase, causing the melting of the densified fuel powder, will be avoided. For this purpose, in longitudinal direction of the fuel rod, a number of diameter reductions of the can are applied of certain distances. In the reduction zone the cross-sectional area of the fuel is reduced, as compared to the one of the remaining fuel material in the regions without diameter reduction, but not the density of the fuel. The recess is chosen to that in case of melting of the fuel in the center of the not contracted zone the fuel in the center of the narrowed area will remain solid and keep the molten material in position. (HR) [de

  8. Argentine activities related to the development of low enriched fuel elements

    International Nuclear Information System (INIS)

    Giorsetti, Domingo R.; Perez, Edmundo E.

    1983-01-01

    Within the framework of the RERTR Program and supported by the technical cooperation work agreed upon between the U.S.A. and Argentina in May 1979, the CNEA Nuclear Fuel Department - Low Enriched Fuel Elements Project (ECBE Project), has carried on its own program for developing fuels with low enrichment for research and test reactors. Up to the present, its main objective has been to replace the highly enriched fuel used in its only reactor (RA-3) for research, development and radioisotopes production. The basic stages of the Argentine Program are shown in Table 1. At a meeting held in Vienna in March, 1980, the CNEA stated that its development of fuels with low enrichment would be in two fuel lines: UAl x -Al and U 3 O 8 -Al, and that its aim would be to reach uranium densities of 18-2.2 g/cm 3 for the UAI x -Al line and 2.4-3.0 g/cm 3 for the U 3 O 8 line. At the international meeting held at ANL in November, 1980, and after having received depleted uranium and uranium with 20% and 45% enrichment (purchased from the U.S.A. for manufacturing miniplates and possible standard fuels) to carry on the proposed development, CNEA anticipated -- after its first tests -- that the conditions were satisfactory for reaching uranium densities of 2.4-3.0 g/cm 3 in U 3 O 8 -Al fuel and of 2.4 g/cm 3 in UAI x -Al fuel. In February 1981, after Argentina accepted the obligation of paying for the irradiation service, authorization was obtained for irradiating miniplates in the Oak Ridge Reactor within the RERTR Program. In June 1981, the first set of miniplates was sent to Oak Ridge National Laboratory (ORNL). The maximum actual densities reached at that time were 3.12 g/cm 3 with U 3 O 8 -Al and 2.52 g/cm 3 with UAl x -Al. During a visit of the CNEA Project Technical Manager to the Argonne National Laboratory (ANL) in July 1981, and after exchanging ideas with ANL professional staff, the CNEA decided to incorporate a new line of development, that of U 3 Si-Al. Three months later

  9. Consolidated fuel reprocessing program

    Science.gov (United States)

    1985-04-01

    A survey of electrochemical methods applications in fuel reprocessing was completed. A dummy fuel assembly shroud was cut using the remotely operated laser disassembly equipment. Operations and engineering efforts have continued to correct equipment operating, software, and procedural problems experienced during the previous uranium compaigns. Fuel cycle options were examined for the liquid metal reactor fuel cycle. In high temperature gas cooled reactor spent fuel studies, preconceptual designs were completed for the concrete storage cask and open field drywell storage concept. These and other tasks operating under the consolidated fuel reprocessing program are examined.

  10. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1979-01-01

    A method and apparatus for the storage of fuel in a stainless steel egg crate structure within a storage pool are described. Fuel is initially stored in a checkerboard pattern or in each opening if the fuel is of low enrichment. Additional fuel (or fuel of higher enrichment) is later stored by adding stainless steel angled plates within each opening, thereby forming flux traps between the openings. Still higher enrichment fuel is later stored by adding poison plates either with or without the stainless steel angles. 8 claims

  11. BNFL Springfields Fuel Division

    International Nuclear Information System (INIS)

    Tarkiainen, S.; Plit, H.

    1998-01-01

    The Fuel Division of British Nuclear Fuels Ltd (BNFL) manufactures nuclear fuel elements for British Magnox and AGR power plants as well as for LWR plants. The new fuel factory - Oxide Fuel Complex (OFC), located in Springfields, is equipped with modern technology and the automation level of the factory is very high. With their quality products, BNFL aims for the new business areas. A recent example of this expansion was shown, when BNFL signed a contract to design and license new VVER-440 fuel for Finnish Loviisa and Hungarian Paks power plants. (author)

  12. Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

    International Nuclear Information System (INIS)

    Park, Y.; Yoo, J.; Huang, K.; Keiser, D.D.; Jue, J.F.; Rabin, B.; Moore, G.; Sohn, Y.H.

    2014-01-01

    Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45–345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface between the U10Mo and Zr, following the co-rolling, the UZr 2 phase was observed to develop adjacent to Zr, and the α-U phase was found between the UZr 2 and U10Mo, while the Mo 2 Zr was found as precipitates mostly within the α-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si) 3 Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the α-U, Mo 2 Zr, and UZr 2 phases

  13. Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

    Energy Technology Data Exchange (ETDEWEB)

    Park, Y.; Yoo, J.; Huang, K. [Advanced Materials Processing and Analysis Center, Department of Materials Science and Engineering, University of Central Florida, Orlando, FL 32816 (United States); Keiser, D.D.; Jue, J.F.; Rabin, B.; Moore, G. [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83401 (United States); Sohn, Y.H., E-mail: Yongho.sohn@ucf.edu [Advanced Materials Processing and Analysis Center, Department of Materials Science and Engineering, University of Central Florida, Orlando, FL 32816 (United States)

    2014-04-01

    Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45–345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface between the U10Mo and Zr, following the co-rolling, the UZr{sub 2} phase was observed to develop adjacent to Zr, and the α-U phase was found between the UZr{sub 2} and U10Mo, while the Mo{sub 2}Zr was found as precipitates mostly within the α-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si){sub 3}Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the α-U, Mo{sub 2}Zr, and UZr{sub 2} phases.

  14. Growth kinetics and microstructural evolution during hot isostatic pressing of U-10 wt.% Mo monolithic fuel plate in AA6061 cladding with Zr diffusion barrier

    Science.gov (United States)

    Park, Y.; Yoo, J.; Huang, K.; Keiser, D. D.; Jue, J. F.; Rabin, B.; Moore, G.; Sohn, Y. H.

    2014-04-01

    Phase constituents and microstructure changes in RERTR fuel plate assemblies as functions of temperature and duration of hot-isostatic pressing (HIP) during fabrication were examined. The HIP process was carried out as functions of temperature (520, 540, 560 and 580 °C for 90 min) and time (45-345 min at 560 °C) to bond 6061 Al-alloy to the Zr diffusion barrier that had been co-rolled with U-10 wt.% Mo (U10Mo) fuel monolith prior to the HIP process. Scanning and transmission electron microscopies were employed to examine the phase constituents, microstructure and layer thickness of interaction products from interdiffusion. At the interface between the U10Mo and Zr, following the co-rolling, the UZr2 phase was observed to develop adjacent to Zr, and the α-U phase was found between the UZr2 and U10Mo, while the Mo2Zr was found as precipitates mostly within the α-U phase. The phase constituents and thickness of the interaction layer at the U10Mo-Zr interface remained unchanged regardless of HIP processing variation. Observable growth due to HIP was only observed for the (Al,Si)3Zr phase found at the Zr/AA6061 interface, however, with a large activation energy of 457 ± 28 kJ/mole. Thus, HIP can be carried to improve the adhesion quality of fuel plate without concern for the excessive growth of the interaction layer, particularly at the U10Mo-Zr interface with the α-U, Mo2Zr, and UZr2 phases.

  15. Nuclear fuel activities in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Bairiot, H

    1997-12-01

    In his presentation on nuclear fuel activities in belgium the author considers the following directions of this work: fuel fabrication, NPP operation, fuel performance, research and development programmes.

  16. Advanced fuel development at AECL: What does the future hold for CANDU fuels/fuel cycles?

    Energy Technology Data Exchange (ETDEWEB)

    Kupferschmidt, W.C.H. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This paper outlines advanced fuel development at AECL. It discusses expanding the limits of fuel utilization, deploy alternate fuel cycles, increase fuel flexibility, employ recycled fuels; increase safety and reliability, decrease environmental impact and develop proliferation resistant fuel and fuel cycle.

  17. 76 FR 37703 - Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel Standards; Public Hearing

    Science.gov (United States)

    2011-06-28

    ... Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel Standards; Public Hearing AGENCY: Environmental... hearing to be held for the proposed rule ``Regulation of Fuels and Fuel Additives: 2012 Renewable Fuel... be proposing amendments to the renewable fuel standard program regulations to establish annual...

  18. DUPIC fuel compatibility assessment

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. [and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition.

  19. DUPIC fuel compatibility assessment

    International Nuclear Information System (INIS)

    Choi, Hang Bok; Rho, G. H.; Park, J. W. and others

    2000-03-01

    The purpose of this study is to assess the compatibility of DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) fuel with the current CANDU 6 reactor, which is one of the technology being developed to utilize the spent PWR fuel in CANDU reactors. The phase 1 study of this project includes the feasibility analysis on applicability of the current core design method, the feasibility analysis on operation of the DUPIC fuel core, the compatibility analysis on individual reactor system, the sensitivity analysis on the fuel composition, and the economic analysis on DUPIC fuel cycle. The results of the validation calculations have confirmed that the current core analysis system is acceptable for the feasibility study of the DUPIC fuel compatibility analysis. The results of core simulations have shown that both natural uranium and DUPIC fuel cores are almost the same from the viewpoint of the operational performance. For individual reactor system including reactively devices, the functional requirements of each system are satisfied in general. However, because of the pronounced power flattening in the DUPIC core, the radiation damage on the critical components increases, which should be investigated more in the future. The DUPIC fuel composition heterogeneity dose not to impose any serious effect on the reactor operation if the fuel composition is adjusted. The economics analysis has been performed through conceptual design studies on the DUPIC fuel fabrication, fuel handling in a plant, and spent fuel disposal, which has shown that the DUPIC fuel cycle is comparable to the once-trough fuel cycle considering uncertainties associated with unit costs of the fuel cycle components. The results of Phase 1 study have shown that it is feasible to use the DUPIC fuel in CANDU reactors without major changes in hardware. However further studies are required to confirm the safety of the reactor under accident condition

  20. The Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    2011-08-01

    This brochure describes the nuclear fuel cycle, which is an industrial process involving various activities to produce electricity from uranium in nuclear power reactors. The cycle starts with the mining of uranium and ends with the disposal of nuclear waste. The raw material for today's nuclear fuel is uranium. It must be processed through a series of steps to produce an efficient fuel for generating electricity. Used fuel also needs to be taken care of for reuse and disposal. The nuclear fuel cycle includes the 'front end', i.e. preparation of the fuel, the 'service period' in which fuel is used during reactor operation to generate electricity, and the 'back end', i.e. the safe management of spent nuclear fuel including reprocessing and reuse and disposal. If spent fuel is not reprocessed, the fuel cycle is referred to as an 'open' or 'once-through' fuel cycle; if spent fuel is reprocessed, and partly reused, it is referred to as a 'closed' nuclear fuel cycle.

  1. The plutonium fuel cycles

    International Nuclear Information System (INIS)

    Pigford, T.H.; Ang, K.P.

    1975-01-01

    The quantities of plutonium and other fuel actinides have been calculated for equilibrium fuel cycles for 1000-MW water reactors fueled with slightly enriched uranium, water reactors fueled with plutonium and natural uranium, fast-breder reactors, gas-cooled reactors fueled with thorium and highly enriched uranium, and gas-cooled reactors fueled with thorium, plutonium and recycled uranium. The radioactivity quantities of plutonium, americium and curium processed yearly in these fuel cycles are greatest for the water reactors fueled with natural uranium and recycled plutonium. The total amount of actinides processed is calculated for the predicted future growth of the U.S. nuclear power industry. For the same total installed nuclear power capacity, the introduction of the plutonium breeder has little effect upon the total amount of plutonium in this century. The estimated amount of plutonium in the low-level process wastes in the plutonium fuel cycles is comparable to the amount of plutonium in the high-level fission product wastes. The amount of plutonium processed in the nuclear fuel cycles can be considerably reduced by using gas-cooled reactors to consume plutonium produced in uranium-fueled water reactors. These, and other reactors dedicated for plutonium utilization, could be co-located with facilities for fuel reprocessing ad fuel fabrication to eliminate the off-site transport of separated plutonium. (author)

  2. Fuel Assembly Damping Summary

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kanghee; Kang, Heungseok; Oh, Dongseok; Yoon, Kyungho; Kim, Hyungkyu; Kim, Jaeyong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper summary the fuel assembly damping data in air/in still water/under flow, released from foreign fuel vendors, compared our data with the published data. Some technical issues in fuel assembly damping measurement testing are also briefly discussed. Understanding of each fuel assembly damping mechanisms according to the surrounding medium and flow velocity can support the fuel design improvement in fuel assembly dynamics and structural integrity aspect. Because the upgraded requirements of the newly-developed advanced reactor system will demands to minimize fuel design margin in integrity evaluation, reduction in conservatism of fuel assembly damping can contribute to alleviate the fuel design margin for sure. Damping is an energy dissipation mechanism in a vibrating mechanical structure and prevents a resonant structure from having infinite vibration amplitudes. The sources of fuel assembly damping are various from support friction to flow contribution, and it can be increased by the viscosity or drag of surrounding fluid medium or the average velocity of water flowing. Fuel licensing requires fuel design evaluation in transient or accidental condition. Dynamic response analysis of fuel assembly is to show fuel integrity and requires information on assembly-wise damping in dry condition and under wet or water flowing condition. However, damping measurement test for the full-scale fuel assembly prototype is not easy to carry out because of the scale (fuel prototype, test facility), unsteadiness of test data (scattering, random sampling and processing), instrumentation under water flowing (water-proof response measurement), and noise. LWR fuel technology division in KAERI is preparing the infra structure for damping measurement test of full-scale fuel assembly, to support fuel industries and related research activities. Here is a preliminary summary of fuel assembly damping, published in the literature. Some technical issues in fuel assembly damping

  3. Romanian nuclear fuel program

    International Nuclear Information System (INIS)

    Budan, O.

    1999-01-01

    The paper presents and comments the policy adopted in Romania for the production of CANDU-6 nuclear fuel before and after 1990. The CANDU-6 nuclear fuel manufacturing started in Romania in December 1983. Neither AECL nor any Canadian nuclear fuel manufacturer were involved in the Romanian industrial nuclear fuel production before 1990. After January 1990, the new created Romanian Electricity Authority (RENEL) assumed the responsibility for the Romanian Nuclear Power Program. It was RENEL's decision to stop, in June 1990, the nuclear fuel production at the Institute for Nuclear Power Reactors (IRNE) Pitesti. This decision was justified by the Canadian specialists team findings, revealed during a general, but well enough technically founded analysis performed at IRNE in the spring of 1990. All fuel manufactured before June 1990 was quarantined as it was considered of suspect quality. By that time more than 31,000 fuel bundles had already been manufactured. This fuel was stored for subsequent assessment. The paper explains the reasons which provoked this decision. The paper also presents the strategy adopted by RENEL after 1990 regarding the Romanian Nuclear Fuel Program. After a complex program done by Romanian and Canadian partners, in November 1994, AECL issued a temporary certification for the Romanian nuclear fuel plant. During the demonstration manufacturing run, as an essential milestone for the qualification of the Romanian fuel supplier for CANDU-6 reactors, 202 fuel bundles were produced. Of these fuel bundles, 66 were part of the Cernavoda NGS Unit 1 first fuel load (the balance was supplied by Zircatec Precision Industries Inc. ZPI). The industrial nuclear fuel fabrication re-started in Romania in January 1995 under AECL's periodical monitoring. In December 1995, AECL issued a permanent certificate, stating the Romanian nuclear fuel plant as a qualified and authorised CANDU-6 fuel supplier. The re-loading of the Cernavoda NGS Unit 1 started in the middle

  4. Fuel Cell Electric Bus Evaluations | Hydrogen and Fuel Cells | NREL

    Science.gov (United States)

    Bus Evaluations Fuel Cell Electric Bus Evaluations NREL's technology validation team evaluates fuel cell electric buses (FCEBs) to provide comprehensive, unbiased evaluation results of fuel cell bus early transportation applications for fuel cell technology. Buses operate in congested areas where

  5. Fuel Cell and Hydrogen Technologies Program | Hydrogen and Fuel Cells |

    Science.gov (United States)

    NREL Fuel Cell and Hydrogen Technologies Program Fuel Cell and Hydrogen Technologies Program Through its Fuel Cell and Hydrogen Technologies Program, NREL researches, develops, analyzes, and validates fuel cell and hydrogen production, delivery, and storage technologies for transportation

  6. Oxy-fuel combustion of pulverized fuels

    DEFF Research Database (Denmark)

    Yin, Chungen; Yan, Jinyue

    2016-01-01

    Oxy-fuel combustion of pulverized fuels (PF), as a promising technology for CO2 capture from power plants, has gained a lot of concerns and also advanced considerable research, development and demonstration in the last past years worldwide. The use of CO2 or the mixture of CO2 and H2O vapor as th...

  7. Logistic Fuel Processor Development

    National Research Council Canada - National Science Library

    Salavani, Reza

    2004-01-01

    ... to light gases then steam reform the light gases into hydrogen rich stream. This report documents the efforts in developing a fuel processor capable of providing hydrogen to a 3kW fuel cell stack...

  8. Future automotive fuels

    International Nuclear Information System (INIS)

    Lepik, M.

    1993-01-01

    There are several important factors which are fundamental to the choice of alternative automobile fuels: the chain of energetic efficiency of fuels; costs; environmental friendliness; suitability for usual engines or adapting easiness; existing reserves of crude oil, natural gas or the fossil energy sources; and, alternatively, agricultural potentiality. This paper covers all these factors. The fuels dealt with in this paper are alcohol, vegetable oil, gaseous fuel, hydrogen and ammonia fuels. Renewable fuels are the most valuable forms of renewable energy. In addition to that rank, they can contribute to three other problem areas: agricultural surpluses, environmental degradation, and conservation of natural resources. Due to the competitive utilization of biomass for food energy production, bio-fuels should mainly be produced in those countries where an energy shortage is combined with a food surplus. The fuels arousing the most interest are alcohol and vegetable oil, the latter for diesel engines, even in northern countries. (au)

  9. Fuel cells: Project Volta

    Energy Technology Data Exchange (ETDEWEB)

    Vellone, R.; Di Mario, F.

    1987-09-01

    This paper discusses research and development in the field of fuel cell power plants. Reference is made to the Italian research Project Volta. Problems related to research program financing and fuel cell power plant marketing are discussed.

  10. Fuel transporting device

    International Nuclear Information System (INIS)

    Shiratori, Hirozo.

    1979-01-01

    Purpose: In a liquid-metal cooled reactor, to reduce the waiting time of fuel handling apparatuses and shorten the fuel exchange time. Constitution: A fuel transporting machine is arranged between a reactor vessel and an out-pile storage tank, thereby dividing the transportation line of the pot for contracting fuel and transporting the same. By assuming such a construction, the flow of fuel transportation which has heretofore been carried out through fuel transportation pipes is not limited to one direction but the take-out of fuels from the reactor and the take-in thereof from the storage tank can be carried out constantly, and much time is not required for fuel exchange. (Kamimura, M.)

  11. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1981-01-01

    An array of rods comprising zirconium alloy sheathed nuclear fuel pellets assembled to form a fuel element for a pressurised water reactor is claimed. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  12. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Hindle, E.D.

    1984-01-01

    The fuel elements for a pressurised water reactor comprise arrays of rods of zirconium alloy sheathed nuclear fuel pellets. The helium gas pressure within each rod differs substantially from that of its closest neighbours

  13. Fuel assembly guide tube

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations

  14. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    Jones, P.M.S.

    1987-01-01

    This chapter explains the distinction between fissile and fertile materials, examines briefly the processes involved in fuel manufacture and management, describes the alternative nuclear fuel cycles and considers their advantages and disadvantages. Fuel management is usually divided into three stages; the front end stage of production and fabrication, the back end stage which deals with the fuel after it is removed from the reactor (including reprocessing and waste treatment) and the stage in between when the fuel is actually in the reactor. These stages are illustrated and explained in detail. The plutonium fuel cycle and thorium-uranium-233 fuel cycle are explained. The differences between fuels for thermal reactors and fast reactors are explained. (U.K.)

  15. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Ueda, Tomihiro.

    1970-01-01

    The present invention relates to fuel assemblies employing wire wrap spacers for retaining uniform spatial distribution between fuel elements. Clad fuel elements are helically wound in the oxial direction with a wave-formed wire strand. The strand is therefore provided with spring action which permits the fuel elements to expand freely in the axial and radial directions so as to retain proper spacing and reduce stresses due to thermal deformation. (Ownes, K.J.)

  16. Fuels and auxiliary materials

    International Nuclear Information System (INIS)

    Svab, V.

    A brief survey is given of the problems of fuels, fuel cans, absorption and moderator materials proceeding from the papers presented at the 1971 4th Geneva Conference on the Peaceful Uses of Nuclear Energy and the 1970 IAEA Conference in New York. Attention is focused on the behaviour of fuel and fuel can materials for thermal and fast reactors during irradiation, radiation stability of absorption materials and the effects of radiation on concrete and on moderator materials. (Z.M.)

  17. Fuel management and economics

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G

    1972-11-01

    From international conference on nuclear solutions to world energy problems; Washington, District of Columbia, USA (12 Nov The low cost of the fuel cycle is the most attractive feature of the fast neutron breeder reactor. In order to achieve it a good fuel management is essential, with well balanced fixed investment and renewal fuel costs. In addition the designer can optimize the power station as a whole (fuel cycle and thermal characteristics). (auth)

  18. Direct hydrocarbon fuel cells

    Science.gov (United States)

    Barnett, Scott A.; Lai, Tammy; Liu, Jiang

    2010-05-04

    The direct electrochemical oxidation of hydrocarbons in solid oxide fuel cells, to generate greater power densities at lower temperatures without carbon deposition. The performance obtained is comparable to that of fuel cells used for hydrogen, and is achieved by using novel anode composites at low operating temperatures. Such solid oxide fuel cells, regardless of fuel source or operation, can be configured advantageously using the structural geometries of this invention.

  19. Spent fuels program

    International Nuclear Information System (INIS)

    Shappert, L.B.

    1983-01-01

    The goal of this task is to support the Domestic Spent Fuel Storage Program through studies involving the transport of spent fuel. A catalog was developed to provide authoritative, timely, and accessible transportation information for persons involved in the transport of irradiated reactor fuel. The catalog, drafted and submitted to the Transportation Technology Center, Sandia National Laboratories, for their review and approval, covers such topics as federal, state, and local regulations, spent fuel characteristics, cask characteristics, transportation costs, and emergency response information

  20. Fuel vapor pressure (FVAPRS)

    International Nuclear Information System (INIS)

    Mason, R.E.

    1979-04-01

    A subcode (FVAPRS) is described which calculates fuel vapor pressure. This subcode was developed as part of the fuel rod behavior modeling task performed at EG and G Idaho, Inc. The fuel vapor pressure subcode (FVAPRS), is presented and a discussion of literature data, steady state and transient fuel vapor pressure equations and estimates of the standard error of estimate to be expected with the FVAPRS subcode are included

  1. Hydrogen Fuel Cell Vehicles

    OpenAIRE

    Anton Francesch, Judit

    1992-01-01

    Hydrogen is an especially attractive transportation fuel. It is the least polluting fuel available, and can be produced anywhere there is water and a clean source of electricity. A fuel cycle in which hydrogen is produced by solar-electrolysis of water, or by gasification of renewably grown biomass, and then used in a fuel-cell powered electric-motor vehicle (FCEV), would produce little or no local, regional, or global pollution. Hydrogen FCEVs would combine the best features of bat...

  2. Denatured fuel cycles

    International Nuclear Information System (INIS)

    Till, C.E.

    1979-01-01

    This paper traces the history of the denatured fuel concept and discusses the characteristics of fuel cycles based on the concept. The proliferation resistance of denatured fuel cycles, the reactor types they involve, and the limitations they place on energy generation potential are discussed. The paper concludes with some remarks on the outlook for such cycles

  3. Hydrogen and fuel cells

    International Nuclear Information System (INIS)

    2006-06-01

    This road-map proposes by the Group Total aims to inform the public on the hydrogen and fuel cells. It presents the hydrogen technology from the production to the distribution and storage, the issues as motor fuel and fuel cells, the challenge for vehicles applications and the Total commitments in the domain. (A.L.B.)

  4. Nuclear fuel cycle

    International Nuclear Information System (INIS)

    1993-01-01

    Status of different nuclear fuel cycle phases in 1992 is discussed including the following issues: uranium exploration, resources, supply and demand, production, market prices, conversion, enrichment; reactor fuel technology; spent fuel management, as well as trends of these phases development up to the year 2010. 10 refs, 11 figs, 15 tabs

  5. PWR fuel thermomechanics

    International Nuclear Information System (INIS)

    Traccucci, R.; Leclercq, J.

    1986-01-01

    Fuel thermo-mechanics means the studies of mechanical and thermal effects, and more generally, the studies of the behavior of the fuel assembly under stresses including thermal and mechanical loads, hydraulic effects and phenomena induced by materials irradiation. This paper describes the studies dealing with the fuel assembly behavior, first in normal operating conditions, and then in accidental conditions. 43 refs [fr

  6. Plutonium fuel program

    International Nuclear Information System (INIS)

    1979-09-01

    A review is presented of the development of the (UPu)C sphere-pac fuel project during 1978. In particular, the problems encountered in obtaining good fuel quality in the fabrication process and their solution is discussed. The development of a fabrication pilot plant is considered, and the post-irradiation examination of fuel pins is presented. (Auth.)

  7. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sasaki, Y.; Tashima, J.

    1975-01-01

    A description is given of nuclear reactor fuel assemblies arranged in the form of a lattice wherein there is attached to the interface of one of two adjacent fuel assemblies a plate spring having a concave portion curved toward said interface and to the interface of the other fuel assembly a plate spring having a convex portion curved away from said interface

  8. Gelled fuel simulant

    International Nuclear Information System (INIS)

    Christy, J.; Hiser, E.J.; Sippel, N.J.

    1980-01-01

    A relatively stable inert simulant formulation for a hazardous metallized fuel has the density, shear rate and yield stress of the duplicated fuel. This formulation provides inexpensive and safe testing of exploratory hydraulic studies, or testing of the mechanical strength of containers, plumbing, etc., in which the metallized fuels are to be used

  9. Fireplaces and Fireplace Fuels.

    Science.gov (United States)

    Metz, Ron

    This instructional unit is one of 10 developed by students on various energy-related areas that deals specifically with fireplaces and fuels. Its objective is for the student to be able to discuss the structural design, operation, and efficiency of fireplaces and characteristics of different fireplace fuels. Some topics covered are fuels, elements…

  10. Metallic fuel development

    International Nuclear Information System (INIS)

    Walters, L.C.

    1987-01-01

    Metallic fuels are capable of achieving high burnup as a result of design modifications instituted in the late 1960's. The gap between the fuel slug and the cladding is fixed such that by the time the fuel swells to the cladding the fission gas bubbles interconnect and release the fission gas to an appropriately sized plenum volume. Interconnected porosity thus provides room for the fuel to deform from further swelling rather than stress the cladding. In addition, the interconnected porosity allows the fuel pin to be tolerant to transient events because as stresses are generated during a transient event the fuel flows rather than applying significant stress to the cladding. Until 1969 a number of metallic fuel alloys were under development in the US. At that time the metallic fuel development program in the US was discontinued in favor of ceramic fuels. However, development had proceeded to the point where it was clear that the zirconium addition to uranium-plutonium fuel would yield a ternary fuel with an adequately high solidus temperature and good compatibility with austenitic stainless steel cladding. Furthermore, several U-Pu-Zr fuel pins had achieved about 6 at.% bu by the late 1960's, without failure, and thus the prospect for high burnup was promising

  11. Modeling fuel succession

    Science.gov (United States)

    Brett Davis; Jan van Wagtendonk; Jen Beck; Kent van Wagtendonk

    2009-01-01

    Surface fuels data are of critical importance for supporting fire incident management, risk assessment, and fuel management planning, but the development of surface fuels data can be expensive and time consuming. The data development process is extensive, generally beginning with acquisition of remotely sensed spatial data such as aerial photography or satellite...

  12. Fuel and fuel cycles with high burnup for WWER reactors

    International Nuclear Information System (INIS)

    Chernushev, V.; Sokolov, F.

    2002-01-01

    The paper discusses the status and trends in development of nuclear fuel and fuel cycles for WWER reactors. Parameters and main stages of implementation of new fuel cycles will be presented. At present, these new fuel cycles are offered to NPPs. Development of new fuel and fuel cycles based on the following principles: profiling fuel enrichment in a cross section of fuel assemblies; increase of average fuel enrichment in fuel assemblies; use of refuelling schemes with lower neutron leakage ('in-in-out'); use of integrated fuel gadolinium-based burnable absorber (for a five-year fuel cycle); increase of fuel burnup in fuel assemblies; improving the neutron balance by using structural materials with low neutron absorption; use of zirconium alloy claddings which are highly resistant to irradiation and corrosion. The paper also presents the results of fuel operation. (author)

  13. Integral nuclear fuel element assembly

    International Nuclear Information System (INIS)

    Schluderberg, D. C.

    1985-01-01

    An integral nuclear fuel element assembly utilizes longitudinally finned fuel pins. The continuous or interrupted fins of the fuel pins are brazed to fins of juxtaposed fuel pins or directly to the juxtaposed fuel pins or both. The integrally brazed fuel assembly is designed to satisfy the thermal and hydraulic requirements of a fuel assembly lattice having moderator to fuel atom ratios required to achieve high conversion and breeding ratios

  14. Reactor fuel element and fuel assembly

    International Nuclear Information System (INIS)

    Okada, Seiji; Ishida, Tsuyoshi; Ikeda, Atsuko.

    1997-01-01

    A mixture of fission products and burnable poisons is disposed at least to a portion between MOX pellets to form a burnable poison-incorporated fuel element without mixing burnable poisons to the MOX pellets. Alternatively, a mixture of materials other than the fission products and burnable poisons is formed into disks, a fuel lamination portion is divided into at least to two regions, and the ratio of number of the disks of the mixture relative to the volume of the region is increased toward the lower portion of the fuel lamination portion. With such a constitution, the axial power distribution of fuels can be made flat easily. Alternatively, the thickness of the disk of the mixture is increased toward the lower region of the fuel lamination portion to flatten the axial power distribution of the fuels in the same manner easily. The time and the cost required for the manufacture are reduced, and MOX fuels filled with burnable poisons with easy maintenance and control can be realized. (N.H.)

  15. 77 FR 61313 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel...

    Science.gov (United States)

    2012-10-09

    ... transportation fuels, including gasoline and diesel fuel, or renewable fuels such as ethanol and biodiesel, as... that which arose under RFS1 for certain renewable fuels (in particular biodiesel) that were produced...

  16. Fuel transfer machine

    International Nuclear Information System (INIS)

    Bernstein, I.

    1978-01-01

    A nuclear fuel transfer machine for transferring fuel assemblies through the fuel transfer tube of a nuclear power generating plant containment structure is described. A conventional reversible drive cable is attached to the fuel transfer carriage to drive it horizontally through the tube. A shuttle carrying a sheave at each end is arranged in parallel with the carriage to also travel into the tube. The cable cooperating with the sheaves permit driving a relatively short fuel transfer carriage a large distance without manually installing sheaves or drive apparatus in the tunnel. 8 claims, 3 figures

  17. Transportation of nuclear fuel

    International Nuclear Information System (INIS)

    Prowse, D.R.

    1979-01-01

    Shipment of used fuel from nuclear reactors to a central fuel management facility is discussed with particular emphasis on the assessment of the risk to the public due to these shipments. The methods of transporting used fuel in large shipping containers is reviewed. In terms of an accident scenario, it is demonstrated that the primary risk of transport of used fuel is due to injury and death in common road accidents. The radiological nature of the used fuel cargo is, for all practical purposes, an insignificant factor in the total risk to the public. (author)

  18. Nuclear fuel lease accounting

    International Nuclear Information System (INIS)

    Danielson, A.H.

    1986-01-01

    The subject of nuclear fuel lease accounting is a controversial one that has received much attention over the years. This has occurred during a period when increasing numbers of utilities, seeking alternatives to traditional financing methods, have turned to leasing their nuclear fuel inventories. The purpose of this paper is to examine the current accounting treatment of nuclear fuel leases as prescribed by the Financial Accounting Standards Board (FASB) and the Federal Energy Regulatory Commission's (FERC's) Uniform System of Accounts. Cost accounting for leased nuclear fuel during the fuel cycle is also discussed

  19. Nuclear fuel string assembly

    International Nuclear Information System (INIS)

    Ip, A.K.; Koyanagi, K.; Tarasuk, W.R.

    1976-01-01

    A method of fabricating rodded fuels suitable for use in pressure tube type reactors and in pressure vessel type reactors is described. Fuel rods are secured as an inner and an outer sub-assembly, each rod attached between mounting rings secured to the rod ends. The two sub-assemblies are telescoped together and positioned by spaced thimbles located between them to provide precise positioning while permittng differential axial movement between the sub-assemblies. Such sub-assemblies are particularly suited for mounting as bundle strings. The method provides particular advantages in the assembly of annular-section fuel pins, which includes booster fuel containing enriched fuel material. (LL)

  20. Mox fuels recycling

    International Nuclear Information System (INIS)

    Gay, A.

    1998-01-01

    This paper will firstly emphasis that the first recycling of plutonium is already an industrial reality in France thanks to the high degree of performance of La Hague and MELOX COGEMA's plants. Secondly, recycling of spent Mixed OXide fuel, as a complete MOX fuel cycle, will be demonstrated through the ability of the existing plants and services which have been designed to proceed with such fuels. Each step of the MOX fuel cycle concept will be presented: transportation, reception and storage at La Hague and steps of spent MOX fuel reprocessing. (author)

  1. Fuel cell opportunities

    Energy Technology Data Exchange (ETDEWEB)

    Harris, K. [Hydrogenics Corporation, Mississauga, ON (Canada)

    2002-07-01

    The opportunities for fuel cell development are discussed. Fuel cells are highly efficient, reliable and require little maintenance. They also produce virtually zero emissions. The author stated that there are some complicated issues to resolve before fuel cells can be widely used. These include hydrogen availability and infrastructure. While the cost of fuel cells is currently very high, these costs are constantly coming down. The industry is still in the early stages of development. The driving forces for the development of fuel cells are: deregulation of energy markets, growing expectations for distributed power generation, discontinuity between energy supply and demand, and environmental concerns. 12 figs.

  2. Fuel loads and fuel type mapping

    Science.gov (United States)

    Chuvieco, Emilio; Riaño, David; Van Wagtendonk, Jan W.; Morsdof, Felix; Chuvieco, Emilio

    2003-01-01

    Correct description of fuel properties is critical to improve fire danger assessment and fire behaviour modeling, since they guide both fire ignition and fire propagation. This chapter deals with properties of fuel that can be considered static in short periods of time: biomass loads, plant geometry, compactness, etc. Mapping these properties require a detail knowledge of vegetation vertical and horizontal structure. Several systems to classify the great diversity of vegetation characteristics in few fuel types are described, as well as methods for mapping them with special emphasis on those based on remote sensing images.

  3. Fuel cells : a viable fossil fuel alternative

    Energy Technology Data Exchange (ETDEWEB)

    Paduada, M.

    2007-02-15

    This article presented a program initiated by Natural Resources Canada (NRCan) to develop proof-of-concept of underground mining vehicles powered by fuel cells in order to eliminate emissions. Recent studies on American and Canadian underground mines provided the basis for estimating the operational cost savings of switching from diesel to fuel cells. For the Canadian mines evaluated, the estimated ventilation system operating cost reductions ranged from 29 per cent to 75 per cent. In order to demonstrate the viability of a fuel cell-powered vehicle, NRCan has designed a modified Caterpillar R1300 loader with a 160 kW hybrid power plant in which 3 stacks of fuel cells deliver up to 90 kW continuously, and a nickel-metal hydride battery provides up to 70 kW. The battery subsystem transiently boosts output to meet peak power requirements and also accommodates regenerative braking. Traction for the loader is provided by a brushless permanent magnet traction motor. The hydraulic pump motor is capable of a 55 kW load continuously. The loader's hydraulic and traction systems are operated independently. Future fuel cell-powered vehicles designed by the program may include a locomotive and a utility vehicle. Future mines running their operations with hydrogen-fueled equipment may also gain advantages by employing fuel cells in the operation of handheld equipment such as radios, flashlights, and headlamps. However, the proton exchange membrane (PEM) fuel cells used in the project are prohibitively expensive. The catalytic content of a fuel cell can add hundreds of dollars per kW of electric output. Production of catalytic precious metals will be strongly connected to the scale of use and acceptance of fuel cells in vehicles. In addition, the efficiency of hydrogen production and delivery is significantly lower than the well-to-tank efficiency of many conventional fuels. It was concluded that an adequate hydrogen infrastructure will be required for the mining industry

  4. Fuel characteristics pertinent to the design of aircraft fuel systems

    Science.gov (United States)

    Barnett, Henry C; Hibbard, R R

    1953-01-01

    Because of the importance of fuel properties in design of aircraft fuel systems the present report has been prepared to provide information on the characteristics of current jet fuels. In addition to information on fuel properties, discussions are presented on fuel specifications, the variations among fuels supplied under a given specification, fuel composition, and the pertinence of fuel composition and physical properties to fuel system design. In some instances the influence of variables such as pressure and temperature on physical properties is indicated. References are cited to provide fuel system designers with sources of information containing more detail than is practicable in the present report.

  5. Fuel charging machine

    International Nuclear Information System (INIS)

    Uchikawa, Sadao.

    1978-01-01

    Purpose: To enable continuous fuel discharging and charging steps in a bwr type reactor by effecting positioning only for once by providing a plurality of fuel assembly grippers and their drives co-axially on a rotatable surface. Constitution: A plurality of fuel assembly grippers and their drives are provided co-axially on a rotatable surface. For example, a gripper A, a drive B, a gripper C and a drive D are arranged co-axially in symmetric positions on a disk rotated on rails by wheels and rotational drives. A new fuel in a fuel pool is gripped by the gripper A and transported above the reactor core. Then, the disk is positioned so that the gripper C can grip the spent fuel in the core, and the fuel to be discharged is gripped and raised by the gripper C. Then the disk is rotated by 180 0 and the new fuel in the gripper A is charged into the position from which the old fuel has been discharged and, finally, the discharged fuel is sent to the fuel pool for storage. (Seki, T.)

  6. Fuel nozzle assembly

    Science.gov (United States)

    Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  7. Ducted fuel injection

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Charles J.

    2018-03-06

    Various technologies presented herein relate to enhancing mixing inside a combustion chamber to form one or more locally premixed mixtures comprising fuel and charge-gas with low peak fuel to charge-gas ratios to enable minimal, or no, generation of soot and other undesired emissions during ignition and subsequent combustion of the locally premixed mixtures. To enable sufficient mixing of the fuel and charge-gas, a jet of fuel can be directed to pass through a bore of a duct causing charge-gas to be drawn into the bore creating turbulence to mix the fuel and the drawn charge-gas. The duct can be located proximate to an opening in a tip of a fuel injector. The duct can comprise of one or more holes along its length to enable charge-gas to be drawn into the bore, and further, the duct can cool the fuel and/or charge-gas prior to combustion.

  8. Reactor fuel charging equipment

    International Nuclear Information System (INIS)

    Wade, Elman.

    1977-01-01

    In many types of reactor fuel charging equipment, tongs or a grab, attached to a trolley, housed in a guide duct, can be used for withdrawing from the core a selected spent fuel assembly or to place a new fuel assembly in the core. In these facilities, the trolley may have wheels that roll on rails in the guide duct. This ensures the correct alignment of the grab, the trolley and fuel assembly when this fuel assembly is being moved. By raising or lowering such a fuel assembly, the trolley can be immerged in the coolant bath of the reactor, whereas at other times it can be at a certain level above the upper surface of the coolant bath. The main object of the invention is to create a fuel handling apparatus for a sodium cooled reactor with bearings lubricated by the sodium coolant and in which the contamination of these bearings is prevented [fr

  9. Fuel element services

    International Nuclear Information System (INIS)

    Marta, H.; Alvarez, P.; Jimenez, J.

    2006-01-01

    Refuelling outages comprise a number of maintenance tasks scheduled long in advance to assure a reliable operation throughout the next cycle and, in the long run, a safer and more efficient plant. Most of these tasks are routine service of mechanical and electrical system and likewise fuel an be considered a critical component as to handling, inspection, cleaning and repair. ENUSA-ENWESA AIE has been working in this area since 1995 growing from fuel repair to a more integrated service that includes new and spent fuel handling, inserts, failed fuel rod detection systems, ultrasonic fuel cleaning, fuel repair and a comprehensive array of inspection and tests related to the reliability of the mechanical components in the fuel assembly, all this, performed in compliance with quality, safety, health physics and any other nuclear standard. (Author)

  10. Fuel cells 101

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, B.

    2003-06-01

    A capsule history of fuel cells is given, beginning with the first discovery in 1839 by William Grove, a Welsh judge who, when experimenting with electrolysis discovered that by re-combining the two components of electrolysis (water and oxygen) an electric charge was produced. A century later, in 1958, Francis Thomas Bacon, a British scientist demonstrated the first working fuel cell stack, a technology which was licensed and used in the Apollo spacecraft. In Canada, early research on the development of fuel cells was carried out at the University of Toronto, the Defence Research Establishment and the National Research Council. Most of the early work concentrated on alkaline and phosphoric acid fuel cells. In 1983, Ballard Research began the development of the electrolyte membrane fuel cell, which marked the beginning of Canada becoming a world leader in fuel cell technology development. The paper provides a brief account of how fuel cells work, describes the distinguishing characteristics of the various types of fuel cells (alkaline, phosphoric acid, molten-carbonate, solid oxide, and proton exchange membrane types) and their principal benefits. The emphasis is on proton exchange membrane fuel cells because they are the only fuel cell technology that is appropriate for providing primary propulsion power onboard a vehicle. Since vehicles are by far the greatest consumers of fossil fuels, it follows that proton exchange membrane fuel cells will have the greatest potential impact on both environmental matters and on our reliance on oil as our primary fuel. Various on-going and planned fuel cell demonstration projects are also described. 1 fig.

  11. 77 FR 72746 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel...

    Science.gov (United States)

    2012-12-06

    ... Fuels and Fuel Additives: Modifications to Renewable Fuel Standard and Diesel Sulfur Programs AGENCY... Fuel Standard (``RFS'') program under section 211(o) of the Clean Air Act. The direct final rule also... marine diesel fuel produced by transmix processors, and the fuel marker requirements for 500 ppm sulfur...

  12. 78 FR 12005 - Regulation of Fuels and Fuel Additives: 2013 Renewable Fuel Standards; Public Hearing

    Science.gov (United States)

    2013-02-21

    ... Regulation of Fuels and Fuel Additives: 2013 Renewable Fuel Standards; Public Hearing AGENCY: Environmental... EPA is announcing a public hearing to be held for the proposed rule ``Regulation of Fuels and Fuel Additives: 2013 Renewable Fuel Standards,'' which was published separately in the Federal Register on...

  13. Oxy-fuel combustion of solid fuels

    DEFF Research Database (Denmark)

    Toftegaard, Maja Bøg; Brix, Jacob; Jensen, Peter Arendt

    2010-01-01

    Oxy-fuel combustion is suggested as one of the possible, promising technologies for capturing CO2 from power plants. The concept of oxy-fuel combustion is removal of nitrogen from the oxidizer to carry out the combustion process in oxygen and, in most concepts, recycled flue gas to lower the flame...... provide additional options for improvement of process economics are however likewise investigated. Of particular interest is the change of the combustion process induced by the exchange of carbon dioxide and water vapor for nitrogen as diluent. This paper reviews the published knowledge on the oxy......-fuel process and focuses particularly on the combustion fundamentals, i.e. flame temperatures and heat transfer, ignition and burnout, emissions, and fly ash characteristics. Knowledge is currently available regarding both an entire oxy-fuel power plant and the combustion fundamentals. However, several...

  14. Nuclear fuel storage

    International Nuclear Information System (INIS)

    Bevilacqua, F.

    1981-01-01

    A nuclear fuel storage apparatus for use in a water-filled pool is fabricated of a material such as stainless steel in the form of an egg crate structure having vertically extending openings. Fuel may be stored in this basic structure in a checkerboard pattern with high enrichment fuel, or in all openings when the fuel is of low effective enrichment. Inserts of a material such as stainless steel are adapted to fit within these openings so that a water gap and, therefore, a flux trap is formed between adjacent fuel storage locations. These inserts may be added at a later time and fuel of a higher enrichment may be stored in each opening. When it is desired to store fuel of still greater enrichment, poison plates may be added to the water gap formed by the installed insert plates, or substituted for the insert plates. Alternately, or in addition, fuel may be installed in high neutron absorption poison boxes which surround the fuel assembly. The stainless steel inserts and the poison plates are each not required until the capacity of the basic egg crate structure is approached. Purchase of these items can, therefore, be deferred for many years. Should the fuel to be stored be of higher enrichment than initially forecast, the deferred decision on the poison plates makes it possible to obtain increased poison in the plates to satisfy the newly discovered requirement

  15. Diesel fuel filtration system

    International Nuclear Information System (INIS)

    Schneider, D.

    1996-01-01

    The American nuclear utility industry is subject to tight regulations on the quality of diesel fuel that is stored at nuclear generating stations. This fuel is required to supply safety-related emergency diesel generators--the backup power systems associated with the safe shutdown of reactors. One important parameter being regulated is the level of particulate contamination in the diesel fuel. Carbon particulate is a natural byproduct of aging diesel fuel. Carbon particulate precipitates from the fuel's hydrocarbons, then remains suspended or settles to the bottom of fuel oil storage tanks. If the carbon particulate is not removed, unacceptable levels of particulate contamination will eventually occur. The oil must be discarded or filtered. Having an outside contractor come to the plant to filter the diesel fuel can be costly and time consuming. Time is an even more critical factor if a nuclear plant is in a Limiting Condition of Operation (LCO) situation. A most effective way to reduce both cost and risk is for a utility to build and install its own diesel fuel filtration system. The cost savings associated with designing, fabricating and operating the system inhouse can be significant, and the value of reducing the risk of reactor shutdown because of uncertified diesel fuel may be even higher. This article describes such a fuel filtering system

  16. Fuel safety research 1999

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-07-01

    In April 1999, the Fuel Safety Research Laboratory was newly established as a result of reorganization of the Nuclear Safety Research Center, JAERI. The laboratory was organized by combining three laboratories, the Reactivity Accident Laboratory, the Fuel Reliability Laboratory, and a part of the Sever Accident Research Laboratory. Consequently, the Fuel Safety Research Laboratory is now in charge of all the fuel safety research in JAERI. Various types of experimental and analytical researches are conducted in the laboratory by using the unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and hot cells in JAERI. The laboratory consists of five research groups corresponding to each research fields. They are; (a) Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). (b) Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). (c) Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). (d) Research group of fuel behavior analysis (FEMAXI group). (e) Research group of FP release/transport behavior from irradiated fuel (VEGA group). This report summarizes the outline of research activities and major outcomes of the research executed in 1999 in the Fuel Safety Research Laboratory. (author)

  17. Study on usage of low enriched uranium Russian type fuel elements for design of an experimental ADS research reactor

    International Nuclear Information System (INIS)

    Pesic, M.P.

    2005-01-01

    Conceptual design of an accelerator driven sub-critical experimental research reactor (ADSRR) was initiated in 1999 at the Vinca Institute of Nuclear Sciences, Serbia and Montenegro. Initial results of neutronic analyses of the proposed ADSRR-H were carried out by Monte Carlo based codes and available high-enriched uranium dioxide (HEU) dispersed Russian type TVR-S fuel elements (FE) placed in a lead matrix. Beam of charged particles (proton or deuteron) would be extracted from the high-energy channel H5B of the VINCY cyclotron of the TESLA Accelerator Installation. In 2002, the Vinca Institute has, in compliance with the Reduced Enrichment for Research and Test Reactors (RERTR) Program, returned fresh HEU TVR-S type FEs back to the Russian Federation. Since usage of HEU FEs in research reactors is not further recommended, a new study of an ADSRR-L conceptual design has initiated in Vinca Institute in last two years, based on assumed availability of low-enriched uranium (LEU) dispersed type TVR-S FEs. Initial results of numerical simulations of this new ADSRR-L, published for the first time in this paper, shows that such a small low neutron flux system can be used as an experimental - 'demonstration' - ADS with neutron characteristics similar to proposed well-known lead moderated and cooled power sub-critical ADS with intermediate neutron spectrum. Neutron spectrum characteristics of the ADSRR-L are compared to ones of the ADSRR-H with the same mass (7.7 g) of 235 U nuclide per TVR-S FE. (author)

  18. Fuel related risks; Braenslerisker

    Energy Technology Data Exchange (ETDEWEB)

    Englund, Jessica; Sernhed, Kerstin; Nystroem, Olle; Graveus, Frank (Grontmij AB, (Sweden))

    2012-02-15

    The project, within which this work report was prepared, aimed to complement the Vaermeforsk publication 'Handbook of fuels' on fuel related risks and measures to reduce the risks. The fuels examined in this project where the fuels included in the first version of the handbook from 2005 plus four additional fuels that will be included in the second and next edition of the handbook. Following fuels were included: woodfuels (sawdust, wood chips, powder, briquettes), slash, recycled wood, salix, bark, hardwood, stumps, straw, reed canary grass, hemp, cereal, cereal waste, olive waste, cocoa beans, citrus waste, shea, sludge, forest industrial sludge, manure, Paper Wood Plastic, tyre, leather waste, cardboard rejects, meat and bone meal, liquid animal and vegetable wastes, tall oil pitch, peat, residues from food industry, biomal (including slaughterhouse waste) and lignin. The report includes two main chapters; a general risk chapter and a chapter of fuel specific risks. The first one deals with the general concept of risk, it highlights laws and rules relevant for risk management and it discuss general risks that are related to the different steps of fuel handling, i.e. unloading, storing, processing the fuel, transportation within the facility, combustion and handling of ashes. The information that was used to produce this chapter was gathered through a literature review, site visits, and the project group's experience from risk management. The other main chapter deals with fuel-specific risks and the measures to reduce the risks for the steps of unloading, storing, processing the fuel, internal transportation, combustion and handling of the ashes. Risks and measures were considered for all the biofuels included in the second version in the handbook of fuels. Information about the risks and risk management was gathered through interviews with people working with different kinds of fuels in electricity and heat plants in Sweden. The information from

  19. Method of decladding spent fuel

    International Nuclear Information System (INIS)

    Fukutome, Kazuyuki; Kitagawa, Kazuo.

    1988-01-01

    Purpose: To enable to safety and easy decladding of nuclear fuels thereby reduce the processing cost. Constitution: Upon dismantling of a spent fuel rod, the fuel rod is heated at least to such a temperature that the ductility of a fuel can is recovered, then transported by using seizing rollers, by which the fuel rod is pressurized from the outer circumference to break the nuclear fuels at the inside thereof. Then, the destructed fuels are recovered from both ends of the fuel can. With such a constitution, since the ductility of the fuel can is recovered by heating, when the fuel rod is passed through the rollers in this state, the fuel can is deformed to destroy the nuclear fuels at the inside thereof. Since the nuclear fuels are destroyed into small pieces, they can be taken out easily from both ends of the fuel can. (Kawakami, Y.)

  20. Two-Dimensional Mapping of the Calculated Fission Power for the Full-Size Fuel Plate Experiment Irradiated in the Advanced Test Reactor

    Science.gov (United States)

    Chang, G. S.; Lillo, M. A.

    2009-08-01

    The National Nuclear Security Administrations (NNSA) Reduced Enrichment for Research and Test Reactors (RERTR) program assigned to the Idaho National Laboratory (INL) the responsibility of developing and demonstrating high uranium density research reactor fuel forms to enable the use of low enriched uranium (LEU) in research and test reactors around the world. A series of full-size fuel plate experiments have been proposed for irradiation testing in the center flux trap (CFT) position of the Advanced Test Reactor (ATR). These full-size fuel plate tests are designated as the AFIP tests. The AFIP nominal fuel zone is rectangular in shape having a designed length of 21.5-in (54.61-cm), width of 1.6-in (4.064-cm), and uniform thickness of 0.014-in (0.03556-cm). This gives a nominal fuel zone volume of 0.482 in3 (7.89 cm3) per fuel plate. The AFIP test assembly has two test positions. Each test position is designed to hold 2 full-size plates, for a total of 4 full-size plates per test assembly. The AFIP test plates will be irradiated at a peak surface heat flux of about 350 W/cm2 and discharged at a peak U-235 burn-up of about 70 at.%. Based on limited irradiation testing of the monolithic (U-10Mo) fuel form, it is desirable to keep the peak fuel temperature below 250°C to achieve this, it will be necessary to keep plate heat fluxes below 500 W/cm2. Due to the heavy U-235 loading and a plate width of 1.6-in (4.064-cm), the neutron self-shielding will increase the local-to-average-ratio (L2AR) fission power near the sides of the fuel plates. To demonstrate that the AFIP experiment will meet the ATR safety requirements, a very detailed 2-dimensional (2D) Y-Z fission power profile was evaluated in order to best predict the fuel plate temperature distribution. The ability to accurately predict fuel plate power and burnup are essential to both the design of the AFIP tests as well as evaluation of the irradiated fuel performance. To support this need, a detailed MCNP Y

  1. Segmented fuel and moderator rod

    International Nuclear Information System (INIS)

    Doshi, P.K.

    1987-01-01

    This patent describes a continuous segmented fuel and moderator rod for use with a water cooled and moderated nuclear fuel assembly. The rod comprises: a lower fuel region containing a column of nuclear fuel; a moderator region, disposed axially above the fuel region. The moderator region has means for admitting and passing the water moderator therethrough for moderating an upper portion of the nuclear fuel assembly. The moderator region is separated from the fuel region by a water tight separator

  2. Nuclear fuel element

    International Nuclear Information System (INIS)

    1974-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding, and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  3. HTPEM Fuel Cell Impedance

    DEFF Research Database (Denmark)

    Vang, Jakob Rabjerg

    As part of the process to create a fossil free Denmark by 2050, there is a need for the development of new energy technologies with higher efficiencies than the current technologies. Fuel cells, that can generate electricity at higher efficiencies than conventional combustion engines, can...... potentially play an important role in the energy system of the future. One of the fuel cell technologies, that receives much attention from the Danish scientific community is high temperature proton exchange membrane (HTPEM) fuel cells based on polybenzimidazole (PBI) with phosphoric acid as proton conductor....... This type of fuel cell operates at higher temperature than comparable fuel cell types and they distinguish themselves by high CO tolerance. Platinum based catalysts have their efficiency reduced by CO and the effect is more pronounced at low temperature. This Ph.D. Thesis investigates this type of fuel...

  4. Boosting nuclear fuels

    International Nuclear Information System (INIS)

    Demarthon, F.; Donnars, O.; Dupuy-Maury, F.

    2002-01-01

    This dossier gives a broad overview of the present day status of the nuclear fuel cycle in France: 1 - the revival of nuclear power as a solution to the global warming and to the increase of worldwide energy needs; 2 - the security of uranium supplies thanks to the reuse of weapon grade highly enriched uranium; 3 - the fabrication of nuclear fuels from the mining extraction to the enrichment processes, the fabrication of fuel pellets and the assembly of fuel rods; 4 - the new composition of present day fuels (UO x and chromium-doped pellets); 5 - the consumption of plutonium stocks and the Corail and Apa fuel assemblies for the reduction of plutonium stocks and the preservation of uranium resources. (J.S.)

  5. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Butterfield, R.S.; Garner, D.L.M.

    1977-01-01

    Reference is made to nuclear fuel assemblies designed for cooling on the 'tube-in-shell' principle in which the fuel is contained by a shell and is cooled by coolant passed through tubes extending through the shell. It has been proposed to employ coated particle fuel as a porous bed on the tube side and the bleed coolant from the tubes into direct contact with the fuel particles. In this way heat is extracted both by direct contact with the fuel and by heat transfer through the coolant tube walls. The system described aims to provide an improved structure of tube and shell for a fuel assembly of this kind and is particularly suitable for use in a gas cooled fast reactor, being able to withstand the neutron flux and high temperature conditions in these reactors. Constructional details are given. (U.K.)

  6. Nuclear fuel element

    International Nuclear Information System (INIS)

    Thompson, J.R.; Rowland, T.C.

    1976-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. A heat conducting, fission product retaining metal liner of a refractory metal is incorporated in the fuel element between the cladding and the nuclear fuel to inhibit mechanical interaction between the nuclear fuel and the cladding, to isolate fission products and nuclear fuel impurities from contacting the cladding and to improve the axial thermal peaking gradient along the length of the fuel rod. The metal liner can be in the form of a tube or hollow cylindrical column, a foil of single or multiple layers in the shape of a hollow cylindrical column, or a coating on the internal surface of the cladding. Preferred refractory metal materials are molybdenum, tungsten, rhenium, niobium and alloys of the foregoing metals

  7. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Hayashi, Hiroshi; Watari, Yoshio; Hizahara, Hiroshi; Masuoka, Ryuzo.

    1970-01-01

    When exchanging nuclear fuel assemblies during the operation of a nuclear reactor, melting of fuel bodies, and severence of tubular claddings is halted at the time of insertion by furnishing a neutron absorbing material such as B 10 , Cd, Gd or the like at the forward end of the fuel assembly to thereby lower the power peak at the forward ends of the fuel elements to within tolerable levels and thus prevent both fuel liquification and excessive expansion. The neutron absorbing material may be attached in the form of a plate to the fuel assembly forward tie plate, or may be inserted as a pellet into the front end of the tubular cladding. (Owens, K.J.)

  8. Fuel assembly spacer

    International Nuclear Information System (INIS)

    Shirakawa, Ken-etsu.

    1988-01-01

    Purpose: To reduce the pressure loss of coolants by fuel assembly spacers. Constitution: Spacers for supporting a fuel assembly are attached by means of a plurality of wires to an outer frame. The outer frame is made of shape memory alloy such that the wires are caused to slacken at normal temperature and the slacking of the wires is eliminated in excess of the transition temperature. Since the wires slacken at the normal temperature, fuel rods can be inserted easily. After the insertion of the fuel rods, when the entire portion or the outer frame is heated by water or gas at a predetermined temperature, the outer frame resumes its previously memorized shape to tighten the wires and, accordingly, the fuel rods can be supported firmly. In this way, since the fuel rods are inserted in the slacken state of the wires and, after the assembling, the outer frame resumes its memorized shape, the assembling work can be conducted efficiently. (Kamimura, M.)

  9. Nuclear fuel element

    International Nuclear Information System (INIS)

    Hirayama, Satoshi; Kawada, Toshiyuki; Matsuzaki, Masayoshi.

    1980-01-01

    Purpose: To provide a fuel element for reducing the mechanical interactions between a fuel-cladding tube and the fuel element and for alleviating the limits of the operating conditions of a reactor. Constitution: A fuel element having mainly uranium dioxide consists of a cylindrical outer pellet and cylindrical inner pellet inserted into the outer pellet. The outer pellet contains two or more additives selected from aluminium oxide, beryllium oxide, magnesium oxide, silicon oxide, sodium oxide, phosphorus oxide, calcium oxide and iron oxide, and the inner pellet contains nuclear fuel substance solely or one additive selected from calcium oxide, silicon oxide, aluminium oxide, magnesium oxide, zirconium oxide and iron oxide. The outer pellet of the fuel thus constituted is reduced in mechanical strength and also in the mechanical interactions with the cladding tube, and the plastic fluidity of the entire pellet is prevented by the inner pellet increased in the mechanical strength. (Kamimura, M.)

  10. Nuclear fuel assembly

    International Nuclear Information System (INIS)

    Betten, P.R.

    1976-01-01

    Under the invention the fuel assembly is particularly suitable for liquid metal cooled fast neutron breeder reactors. Hence, according to the invention a fuel assembly cladding includes inward corrugations with respect to the remainder of the cladding according to a recurring pattern determined by the pitch of the metal wire helically wound round the fuel rods of the assembly. The parts of the cladding pressed inwards correspond to the areas in which the wire encircling the peripheral fuel rods is generally located apart from the cladding, thereby reducing the play between the cladding and the peripheral fuel rods situated in these areas. The reduction in the play in turn improves the coolant flow in the internal secondary channels of the fuel assembly to the detriment of the flow in the peripheral secondary channels and thereby establishes a better coolant fluid temperature profile [fr

  11. Nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Natori, Hisahide; Kurihara, Kunitoshi.

    1982-01-01

    Purpose: To increase the fuel safety by decreasing the gap conductance between fuels and cladding tubes, as well as improve the reactor core controllability by rendering the void coefficient negative. Constitution: Fuel assemblies in a pressure tube comprise a tie-rod, fuel rods in a central region, and fuel rods with burnable poison in the outer circumference region. Here, B 4 C is used as the burnable poison by 1.17 % by weight ratio. The degrees of enrichment for the fissile plutonium as PuO 2 -UO 2 fuel used in the assemblies are 2.7 %, 2.7 % and 1.5 % respectively in the innermost layer, the intermediate layer and the outermost layer. This increases the burn-up degree to improve the plant utilizability, whereby the void coefficient is rendered negative to improve the reactor core controllability. (Horiuchi, T.)

  12. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Sakurai, Shungo; Ogiya, Shunsuke.

    1990-01-01

    In a fuel assembly, if the entire fuels comprise mixed oxide fuels, reactivity change in cold temperature-power operation is increased to worsen the reactor shutdown margin. The reactor shutdown margin has been improved by increasing the burnable poison concentration thereby reducing the reactivity of the fuel assembly. However, since unburnt poisons are present at the completion of the reactor operation, the reactivity can not be utilized effectively to bring about economical disadvantage. In view of the above, the reactivity change between lower temperature-power operations is reduced by providing a non-boiling range with more than 9.1% of cross sectional area at the inside of a channel at the central portion of the fuel assembly. As a result, the amount of the unburnt burnable poisons is decreased, the economy of fuel assembly is improved and the reactor shutdown margin can be increase. (N.H.)

  13. Fuel Cell and Hydrogen Technology Validation | Hydrogen and Fuel Cells |

    Science.gov (United States)

    NREL Fuel Cell and Hydrogen Technology Validation Fuel Cell and Hydrogen Technology Validation The NREL technology validation team works on validating hydrogen fuel cell electric vehicles; hydrogen fueling infrastructure; hydrogen system components; and fuel cell use in early market applications such as

  14. Alternative Fuels Data Center: Krug Energy Opens Natural Gas Fueling

    Science.gov (United States)

    Station in Arkansas Krug Energy Opens Natural Gas Fueling Station in Arkansas to someone by E -mail Share Alternative Fuels Data Center: Krug Energy Opens Natural Gas Fueling Station in Arkansas on Facebook Tweet about Alternative Fuels Data Center: Krug Energy Opens Natural Gas Fueling Station in

  15. Fuel cell generator with fuel electrodes that control on-cell fuel reformation

    Science.gov (United States)

    Ruka, Roswell J [Pittsburgh, PA; Basel, Richard A [Pittsburgh, PA; Zhang, Gong [Murrysville, PA

    2011-10-25

    A fuel cell for a fuel cell generator including a housing including a gas flow path for receiving a fuel from a fuel source and directing the fuel across the fuel cell. The fuel cell includes an elongate member including opposing first and second ends and defining an interior cathode portion and an exterior anode portion. The interior cathode portion includes an electrode in contact with an oxidant flow path. The exterior anode portion includes an electrode in contact with the fuel in the gas flow path. The anode portion includes a catalyst material for effecting fuel reformation along the fuel cell between the opposing ends. A fuel reformation control layer is applied over the catalyst material for reducing a rate of fuel reformation on the fuel cell. The control layer effects a variable reformation rate along the length of the fuel cell.

  16. Fuel safety research 2001

    Energy Technology Data Exchange (ETDEWEB)

    Uetsuka, Hiroshi (ed.) [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-11-01

    The Fuel Safety Research Laboratory is in charge of research activity which covers almost research items related to fuel safety of water reactor in JAERI. Various types of experimental and analytical researches are being conducted by using some unique facilities such as the Nuclear Safety Research Reactor (NSRR), the Japan Material Testing Reactor (JMTR), the Japan Research Reactor 3 (JRR-3) and the Reactor Fuel Examination Facility (RFEF) of JAERI. The research to confirm the safety of high burn-up fuel and MOX fuel under accident conditions is the most important item among them. The laboratory consists of following five research groups corresponding to each research fields; Research group of fuel behavior under the reactivity initiated accident conditions (RIA group). Research group of fuel behavior under the loss-of-coolant accident conditions (LOCA group). Research group of fuel behavior under the normal operation conditions (JMTR/BOCA group). Research group of fuel behavior analysis (FEMAXI group). Research group of radionuclides release and transport behavior from irradiated fuel under severe accident conditions (VEGA group). The research conducted in the year 2001 produced many important data and information. They are, for example, the fuel behavior data under BWR power oscillation conditions in the NSRR, the data on failure-bearing capability of hydrided cladding under LOCA conditions and the FP release data at very high temperature in steam which simulate the reactor core condition during severe accidents. This report summarizes the outline of research activities and major outcomes of the research executed in 2001 in the Fuel Safety Research Laboratory. (author)

  17. Transport of MOX fuel

    International Nuclear Information System (INIS)

    Porter, I.R.; Carr, M.

    1997-01-01

    The regulatory framework which governs the transport of MOX fuel is set out, including packages, transport modes and security requirements. Technical requirements for the packages are reviewed and BNFL's experience in plutonium and MOX fuel transport is described. The safety of such operations and the public perception of safety are described and the question of gaining public acceptance for MOX fuel transport is addressed. The paper concludes by emphasising the need for proactive programmes to improve the public acceptance of these operations. (Author)

  18. Fuel assembly storage pool

    International Nuclear Information System (INIS)

    Hiranuma, Hiroshi.

    1976-01-01

    Object: To remove limitation of the number of storage of fuel assemblies to increase the number of storage thereof so as to relatively reduce the water depth required for shielding radioactive rays. Structure: Fuel assembly storage rack containers for receiving a plurality of spent fuel assembly racks are stacked in multi-layer fashion within a storage pool filled with water for shielding radioactive rays and removing heat. (Furukawa, Y.)

  19. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    Butterfield, C.E.; Waite, E.

    1982-01-01

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  20. Spent fuel storage

    International Nuclear Information System (INIS)

    Huppert

    1976-01-01

    To begin with, the author explains the reasons for intermediate storage of fuel elements in nuclear power stations and in a reprocessing plant and gives the temperature and radioactivity curves of LWR fuel elements after removal from the reactor. This is followed by a description of the facilities for fuel element storage in a reprocessing plant and of their functions. Futher topics are criticality and activity control, the problem of cooling time and safety systems. (HR) [de

  1. Liquid fuel cells

    Directory of Open Access Journals (Sweden)

    Grigorii L. Soloveichik

    2014-08-01

    Full Text Available The advantages of liquid fuel cells (LFCs over conventional hydrogen–oxygen fuel cells include a higher theoretical energy density and efficiency, a more convenient handling of the streams, and enhanced safety. This review focuses on the use of different types of organic fuels as an anode material for LFCs. An overview of the current state of the art and recent trends in the development of LFC and the challenges of their practical implementation are presented.

  2. Nuclear fuel accounting

    International Nuclear Information System (INIS)

    Aisch, D.E.

    1977-01-01

    After a nuclear power plant has started commercial operation the actual nuclear fuel costs have to be demonstrated in the rate making procedure. For this purpose an accounting system has to be developed which comprises the following features: 1) All costs associated with nuclear fuel shall be correctly recorded; 2) it shall be sufficiently flexible to cover also deviations from proposed core loading patterns; 3) it shall be applicable to different fuel cycle schemes. (orig./RW) [de

  3. Nuclear fuel financing

    International Nuclear Information System (INIS)

    Lurf, G.

    1975-01-01

    Fuel financing is only at its beginning. A logical way of developing financing model is a step by step method starting with the financing of pre-payments. The second step will be financing of natural uranium and enrichment services to the point where the finished fuel elements are delivered to the reactor operator. The third step should be the financing of fuel elements during the time the elements are inserted in the reactor. (orig.) [de

  4. Alternative Fuels (Briefing Charts)

    Science.gov (United States)

    2009-06-19

    feedstock for HRJ, plant cost for F-T) Courtesy AFRL, Dr. Tim Edwards Unclassified • Agricultural crop oils (canola, jatropha, soy, palm , etc...Fuels Focus  Various conversion processes  Upgraded to meet fuel specs Diverse energy sources Petroleum Crude Oil Petroleum based Single Fuel in the...data and resources – Conduct gap analysis – synfuel efforts, expand to biofuels, ID potential joint efforts – Increase visibility outside SCP world

  5. Fuel Thermal Expansion (FTHEXP)

    International Nuclear Information System (INIS)

    Reymann, G.A.

    1978-07-01

    A model is presented which deals with dimensional changes in LWR fuel pellets caused by changes in temperature. It is capable of dealing with any combination of UO 2 and PuO 2 in solid, liquid or mixed phase states, and includes expansion due to the solid-liquid phase change. The function FTHEXP models fuel thermal expansion as a function of temperature, fraction of PuO 2 , and the fraction of fuel which is molten

  6. Fusion fuel and renewables

    International Nuclear Information System (INIS)

    Entler, Slavomir

    2015-01-01

    It is shown that fusion fuel meets all aspects applied when defining renewables. A table of definitions of renewables is presented. The sections of the paper are as follows: An industrial renewable source; Nuclear fusion; Current situation in research; Definitions of renewable sources; Energy concept of nuclear fusion; Fusion fuel; Natural energy flow; Environmental impacts; Fusion fuel assessment; Sustainable power; and Energy mix from renewables. (P.A.)

  7. The nuclear fuel cycle

    International Nuclear Information System (INIS)

    1998-05-01

    After a short introduction about nuclear power in the world, fission physics and the French nuclear power plants, this brochure describes in a digest way the different steps of the nuclear fuel cycle: uranium prospecting, mining activity, processing of uranium ores and production of uranium concentrates (yellow cake), uranium chemistry (conversion of the yellow cake into uranium hexafluoride), fabrication of nuclear fuels, use of fuels, reprocessing of spent fuels (uranium, plutonium and fission products), recycling of energetic materials, and storage of radioactive wastes. (J.S.)

  8. Protocol Fuel Mix reporting

    International Nuclear Information System (INIS)

    2002-07-01

    The protocol in this document describes a method for an Electricity Distribution Company (EDC) to account for the fuel mix of electricity that it delivers to its customers, based on the best available information. Own production, purchase and sale of electricity, and certificates trading are taken into account. In chapter 2 the actual protocol is outlined. In the appendixes additional (supporting) information is given: (A) Dutch Standard Fuel Mix, 2000; (B) Calculation of the Dutch Standard fuel mix; (C) Procedures to estimate and benchmark the fuel mix; (D) Quality management; (E) External verification; (F) Recommendation for further development of the protocol; (G) Reporting examples

  9. FAILED FUEL DISPOSITION STUDY

    Energy Technology Data Exchange (ETDEWEB)

    THIELGES, J.R.

    2004-12-20

    In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

  10. Reference thorium fuel cycle

    International Nuclear Information System (INIS)

    Driggers, F.E.

    1978-08-01

    In the reference fuel cycle for the TFCT program, fissile U will be denatured by mixing with 238 U; the plants will be located in secure areas, with Pu being recycled within these secure areas; Th will be recycled with recovered U and Pu; the head end will handle a variety of core and blanket fuel assembly designs for LWRs and HWRs; the fuel may be a homogeneous mixture either of U and Th oxide pellets or sol-gel microspheres; the cladding will be Zircaloy; and MgO may be added to the fuel to improve Th dissolution. Th is being considered as the fertile component of fuel in order to increase proliferation resistance. Spent U recovered from Th-based fuels must be re-enriched before recycle to prevent very rapid buildup of 238 U. Stainless steel will be considered as a backup to Zircaloy cladding in case Zr is incompatible with commercial aqueous dissolution. Storage of recovered irradiated Th will be considered as a backup to its use in the recycle of recovered Pu and U. Estimates are made of the time for introducing the Th fuel cycle into the LWR power industry. Since U fuel exposures in LWRs are likely to increase from 30,000 to 50,000 MWD/MT, the Th reprocessing plant should also be designed for Th fuel with 50,000 MWD/MT exposure

  11. A perfect fuel supplier

    International Nuclear Information System (INIS)

    Terasvirta, R.

    2008-01-01

    WWER fuel market is dominated by the Russian fuel vendor JSC TVEL. There have been attempts to open up the market also for other suppliers, such as BNFL/Westinghouse for Finland, Czech Republic, and Ukraine. However, at the moment it seems that JSC TVEL is the only real alternative to supply fuel to WWER reactors. All existing fuel suppliers have certified quality management systems which put a special emphasis on the customer satisfaction. This paper attempts to define from the customer's point of view, what are the important issues concerning the customer satisfaction. (author)

  12. Fuel cells - a perspective

    International Nuclear Information System (INIS)

    Biegler, T.

    2005-01-01

    Unfortunately, fuel cell publicity conveys expectations and hopes that are often based on uncritical interpretations of the underlying science. The aim here is to use that science to analyse how the technology has developed and what can realistically be delivered by fuel cells. There have been great achievements in fuel cell technology over the past decade, with most types reaching an advanced stage of engineering development. But there has been some muddled thinking about one critical aspect, fuel cell energy efficiency. The 'Carnot cycle' argument, that fuel cells must be much more efficient than heat engines, is a red herring, of no help in predicting real efficiencies. In practice, fuel cells are not always particularly efficient and there are good scientific reasons for this. Cost reduction is a big issue for fuel cells. They are not in principle especially simple devices. Better engineering and mass production will presumably bring costs down, but because of their inherent complexity there is no reason to expect them to be cheap. It is fair to conclude that predictions of fuel cells as commonplace components of energy systems (including a hydrogen economy) need to be treated with caution, at least until major improvements eventuate. However, one type, the direct methanol fuel cell, is aimed at a clear existing market in consumer electronics

  13. FAILED FUEL DISPOSITION STUDY

    International Nuclear Information System (INIS)

    THIELGES, J.R.

    2004-01-01

    In May 2004 alpha contamination was found on the lid of the pre-filter housing in the Sodium Removal Ion Exchange System during routine filter change. Subsequent investigation determined that the alpha contamination likely came from a fuel pin(s) contained in an Ident-69 (ID-69) type pin storage container serial number 9 (ID-69-9) that was washed in the Sodium Removal System (SRS) in January 2004. Because all evidence indicated that the wash water interacted with the fuel, this ID49 is designated as containing a failed fuel pin with gross cladding defect and was set aside in the Interim Examination and Maintenance (IEM) Cell until it could be determined how to proceed for long term dry storage of the fuel pin container. This ID49 contained fuel pins from the driver fuel assembly (DFA) 16392, which was identified as a Delayed Neutron Monitor (DNM) leaker assembly. However, this DFA was disassembled and the fuel pin that was thought to be the failed pin was encapsulated and was not located in this ID49 container. This failed fuel disposition study discusses two alternatives that could be used to address long term storage for the contents of ID-69-9. The first alternative evaluated utilizes the current method of identifying and storing DNM leaker fuel pin(s) in tubes and thus, verifying that the alpha contamination found in the SRS came from a failed pin in this pin container. This approach will require unloading selected fuel pins from the ID-69, visually examining and possibly weighing suspect fuel pins to identify the failed pin(s), inserting the failed pin(s) in storage tubes, and reloading the fuel pins into ID49 containers. Safety analysis must be performed to revise the 200 Area Interim Storage Area (ISA) Final Safety Analysis Report (FSAR) (Reference 1) for this fuel configuration. The second alternative considered is to store the failed fuel as-is in the ID-69. This was evaluated to determine if this approach would comply with storage requirements. This

  14. Fuel assembly reconstitution

    International Nuclear Information System (INIS)

    Morgado, Mario M.; Oliveira, Monica G.N.; Ferreira Junior, Decio B.M.; Santos, Barbara O. dos; Santos, Jorge E. dos

    2009-01-01

    Fuel failures have been happened in Nuclear Power Plants worldwide, without lost of integrity and safety, mainly for the public, environment and power plants workers. The most common causes of these events are corrosion (CRUD), fretting and pellet cladding interaction. These failures are identified by increasing the activity of fission products, verified by chemical analyses of reactor coolant. Through these analyses, during the fourth operation cycle of Angra 2 Nuclear Power Plant, was possible to observe fuel failure indication. This indication was confirmed in the end of the cycle during the unloading of reactor core through leakage tests of fuel assembly, using the equipment called 'In Mast Sipping' and 'Box Sipping'. After confirmed, the fuel assembly reconstitution was scheduled, and happened in April, 2007, where was identified the cause and the fuel rod failure, which was substitute by dummy rods (zircaloy). The cause was fretting by 'debris'. The actions to avoid and prevent fuel assemblies failures are important. The goals of this work are to describe the methodology of fuel assembly reconstitution using the FARE (Fuel Assembly Reconstitution Equipment) system, to describe the results of this task in economic and security factors of the company and show how the fuel assembly failures are identified during operation and during the outage. (author)

  15. Fuel cell catalyst degradation

    DEFF Research Database (Denmark)

    Arenz, Matthias; Zana, Alessandro

    2016-01-01

    Fuel cells are an important piece in our quest for a sustainable energy supply. Although there are several different types of fuel cells, the by far most popular is the proton exchange membrane fuel cell (PEMFC). Among its many favorable properties are a short start up time and a high power density...... increasing focus. Activity of the catalyst is important, but stability is essential. In the presented perspective paper, we review recent efforts to investigate fuel cell catalysts ex-situ in electrochemical half-cell measurements. Due to the amount of different studies, this review has no intention to give...

  16. ABB high burnup fuel

    International Nuclear Information System (INIS)

    Andersson, S.; Helmersson, S.; Nilsson, S.; Jourdain, P.; Karlsson, L.; Limback, M.; Garde, A.M.

    1999-01-01

    Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both PWR and BWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter proven to meet the 6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10 x 10 fuel, where ABB is the only vendor to date with batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of PWR and BWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its utility customers. This paper provides an overview of recent fuel performance and reliability experience at ABB. Selected development and validation activities for PWR and BWR fuel are presented, for which the ABB test facilities in Windsor (TF-2 loop, mechanical test laboratory) and Vaesteras (FRIGG, BURE) are essential. (authors)

  17. Fuel rod leak detector

    International Nuclear Information System (INIS)

    Womack, R.E.

    1978-01-01

    A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

  18. Fuels Processing Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — NETL’s Fuels Processing Laboratory in Morgantown, WV, provides researchers with the equipment they need to thoroughly explore the catalytic issues associated with...

  19. Fuel Cell Demonstration Program

    Energy Technology Data Exchange (ETDEWEB)

    Gerald Brun

    2006-09-15

    In an effort to promote clean energy projects and aid in the commercialization of new fuel cell technologies the Long Island Power Authority (LIPA) initiated a Fuel Cell Demonstration Program in 1999 with six month deployments of Proton Exchange Membrane (PEM) non-commercial Beta model systems at partnering sites throughout Long Island. These projects facilitated significant developments in the technology, providing operating experience that allowed the manufacturer to produce fuel cells that were half the size of the Beta units and suitable for outdoor installations. In 2001, LIPA embarked on a large-scale effort to identify and develop measures that could improve the reliability and performance of future fuel cell technologies for electric utility applications and the concept to establish a fuel cell farm (Farm) of 75 units was developed. By the end of October of 2001, 75 Lorax 2.0 fuel cells had been installed at the West Babylon substation on Long Island, making it the first fuel cell demonstration of its kind and size anywhere in the world at the time. Designed to help LIPA study the feasibility of using fuel cells to operate in parallel with LIPA's electric grid system, the Farm operated 120 fuel cells over its lifetime of over 3 years including 3 generations of Plug Power fuel cells (Lorax 2.0, Lorax 3.0, Lorax 4.5). Of these 120 fuel cells, 20 Lorax 3.0 units operated under this Award from June 2002 to September 2004. In parallel with the operation of the Farm, LIPA recruited government and commercial/industrial customers to demonstrate fuel cells as on-site distributed generation. From December 2002 to February 2005, 17 fuel cells were tested and monitored at various customer sites throughout Long Island. The 37 fuel cells operated under this Award produced a total of 712,635 kWh. As fuel cell technology became more mature, performance improvements included a 1% increase in system efficiency. Including equipment, design, fuel, maintenance

  20. Data feature: Fuel procurement

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    This document is a review of the effect of fuel costs on the procurement strategies of a utility and a conjecture that the same strategies may have an effect on the price of fuel. Factors affecting fuel costs are reviewed, and a number of procurement strategies taken to trim fuel costs are reviewed. The major trend is away from long-term enrichment contracts and into such strategies as: (1) Spot market purchases, (2) Inventory reduction, (3) Purchase of CIS material, and (4) Market-related contracts instead of base-escalated contracts