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

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

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)

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

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)

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.

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

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

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

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

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

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

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

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)

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)

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

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)

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)

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

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

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

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

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

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

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)

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

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)

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)

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)

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)

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)

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

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

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.

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

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

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)

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.

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

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

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)

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.

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

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)

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

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

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

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

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

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

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

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.)

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.

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.

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.

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

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

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

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.

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)

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)

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)

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

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.

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

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)

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

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

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

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.

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

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

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)

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

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.

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)

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

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)

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)

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.

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.

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

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)

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

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.)

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)

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)

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)

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.

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)

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)

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)

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

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

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

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)

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)

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)

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)

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)

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)

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)

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...

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

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)

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)

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)

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)

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

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)

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

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

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.

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

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

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

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.

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.

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.

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

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

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

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)

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

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

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

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)

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

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)

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.

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)

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)

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.

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)

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.

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)

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.

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.

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.

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.

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

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)

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)

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)

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

DOE/NNSA introduction

International Nuclear Information System (INIS)

Black, S.

2005-01-01

The evolving threat of international terrorism and nuclear proliferation has manifested itself in a very compelling way in recent years. That' is why it is so important that experts continue to work together to minimize and eventually eliminate the use of HEU in civil nuclear applications, throughout the world, and as soon as possible. The most common civil use of HEU is in research and test reactors, either as fuel or targets. About 150 such reactors currently use HEU fuels. Research reactors and the use of uranium as targets in reactors play indispensable roles in the production of medical radioisotopes, radiation therapy, nuclear training, material testing, and scientific research. However, in nearly all cases, these missions can be accomplished without significant penalties using low enriched uranium fuels and targets, which would avoid the proliferation risk associated with HEU. To date, the RERTR program has made possible the partial or full conversion to LEU fuel of 39 reactors in 22 countries - reducing civil commerce in HEU by about 4 metric tons, sufficient material to manufacture over 100 nuclear weapons. In addition, since the start of the RERTR program in 1978, 21 research reactors in 16 countries have been built using an LEU fuel. This trend indicates that we are having an effect, that we are changing the way people think about research reactors. The focus for new construction is almost exclusively on LEU designs using fuels developed by the RERTR program. This has been a very interesting year for the RERTR program in terms of milestones and, regrettably, setbacks. Some serious fuel performance problems were encountered with the uranium-molybdenum dispersion fuel this year. On a more positive note, however, we had a major nonproliferation breakthrough with Libya and are now cooperating with Libyan experts to convert their research reactor and critical assembly to LEU fuel. As the global proliferation threat continues to evolve, we need a more

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

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)

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)

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

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.

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)

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)

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

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

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.

Recent observations at the post-irradiation examination of low-enriched U-Mo miniplates irradiated to high burn-up

International Nuclear Information System (INIS)

Hofman, G.L.; Kim, Y.S.; Finlay, M.R.; Snelgrove, J.L.; Hayes, S.L.; Meyer, M.K.; Clark, C.R.

2003-01-01

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 Cell Facility. 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 6.5 wt% to 10 wt% molybdenum. In addition, two miniplates contained solid U-10wt%Mo foils. Recent results of the postirradiation examination and analysis of RERTR-4 in conjunction with data from a companion test performed to 50% burnup, RERTR-5, are presented. (author)

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)

Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

Energy Technology Data Exchange (ETDEWEB)

Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J B [eds.

1995-07-01

Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with 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. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics.

Reduced Enrichment for Research and Test Reactors. Proceedings of the XIV international meeting

International Nuclear Information System (INIS)

Suripto, Asmedi; Hastowo, Hudi; Hersubeno, J.B.

1995-01-01

Apart from the progress of the Reduced Enrichment Research and Test Reactor (RERTR) Program the national programs of Indonesia, Japan and China were presented. The major events, findings, and activities of 1991 are reviewed with 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. A number of presentations were devoted to development of new fuel uranium silicide fuel elements, fuel irradiation testing and reactor core conversions from highly enriched (HEU) to slightly enriched uranium (LEU). Calculations and measurements of converted reactor core parameters were shown related to safety test and analysis. Fuel cycle issue were discussed as well. One should note that a significant number of papers were devoted to Indonesian GA SIWABESSY reactor core conversion and related topics

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

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

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)

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)

Uncertainties for industrial suppliers in view to conversions and designs of research reactors

International Nuclear Information System (INIS)

Roegler, H.J.; Stroemich, A.

1991-01-01

The different national RERTR-programs have had substantial results and offer clear potential for LEU-fuel. Nevertheless there are still attempts, papers and political actions based on future developments of LEU-fuel. They may create the same uncertainty as 10 years ago again, depending on what plant the industry has to supply. Only one conclusion can be drawn: There has to be an agreement on the results of the RERTR-programs binding for all parties and the industry has a right to demand such agreement after 10 years of heavily subsidised fuel development. (orig.)

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%.

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%

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.

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

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)

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

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)

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)

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.

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

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)

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

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.

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

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)

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)

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)

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.

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

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)

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

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)

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

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)

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)

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.

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.

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)

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)

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.

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

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.

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.

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.

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)

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)

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)

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

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

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.

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

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)

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)

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.

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)

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.

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.

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

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.

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.

Why does the need of HEU for high flux research reactors remain?

International Nuclear Information System (INIS)

Glaeser, W.

1991-01-01

It has shown that high performance high flux reactors need an ongoing supply of highly enriched uranium. The new fuel materials in their highly enriched version offer prospective for advanced and better neutron sources vital for the future of neutron research. This is another very attractive result of the RERTR programme. One-sided restriction would only provide marginal or no values for research. If we adopt the sometimes expressed views that high enriched RERTR developed fuel should only be made available when unique benefits to mankind could be obtained, then certainly basic research at the forefront belongs to this category. HEU would only pose theoretical difficulties, if it would remain under proper safeguards and obviously this is the way to be pursued. (orig.)

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.

Reduced enrichment for research and test reactors: Proceedings

International Nuclear Information System (INIS)

1993-07-01

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

Reduced enrichment for research and test reactors: Proceedings

Energy Technology Data Exchange (ETDEWEB)

1993-07-01

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

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.

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.

Conceptual study of multiphase structure of high uranium density alloys to reduce chemical interaction

International Nuclear Information System (INIS)

Savchenko, A.

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

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)

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.

Reduced enrichment for research and test reactors: Proceedings

Energy Technology Data Exchange (ETDEWEB)

1993-08-01

November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately.

Reduced enrichment for research and test reactors: Proceedings

International Nuclear Information System (INIS)

1993-08-01

November 9--10, 1978, marked the first of what has become an annual event--the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and it established the basis for all later meetings. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort. This report provides presentations and discussions of this original meeting. Individual papers have been cataloged separately

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)

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

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.

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

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

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

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

Welcome address to the 26th international meeting on Reduced Enrichment for Research and Test Reactors

International Nuclear Information System (INIS)

Sokolov, Y.

2005-01-01

While the IAEA has been a vigorous supporter of the RERTR programme since its inception. RERTR and the related fresh and spent fuel return efforts have gained new momentum with the launching of the Global Threat Reduction Initiative (GTRI) by U.S. Energy Secretary Abraham here in Vienna on May 25, 2004. All of the activities to be be discussed are included within the framework of the GTRI. The international programmes to qualify high density, LEU, dispersion fuels based on U-Mo alloys have run into unexpected technical difficulties that will delay qualification. A number of the presentations address the problems that have been encountered. At the same time, it is encouraging that the international resolve to reduce and eventually eliminate HEU in international commerce appears to have strengthened. In the past year, fresh HEU at research reactors in different countries have been returned to the country of origin. In all these examples, the return of the fresh fuel was accompanied by plans for conversion of existing reactors or design of new reactors to use LEU, as well as for the repatriation of spent research reactor fuel. The IAEA, particularly the Department of Technical Cooperation and my Department of Nuclear Energy has played an important role in implementing these fresh fuel return activities. In addition, several of the reactor conversion projects will be carried out under the auspices of IAEA technical cooperation projects and with important involvement of the Department of Nuclear Energy. The IAEA has also supported the repatriation of spent fuel to the country of original enrichment. The U.S. spent fuel acceptance programme has been operating for more than eight years, and was originally scheduled to terminate in 2006. Important announcements concerning the extension of the U.S. programme are expected. At the same time, the IAEA has been working hard with the U.S. and Russia to initiate the Russian research reactor spent fuel return programme. We are

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

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

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

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

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

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.

Reactivity effect of poisoned beryllium block shuffling in the MARIA reactor

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.

2000-01-01

The paper is a continuation of the analysis of beryllium blocks poisoning by Li-6 and He-3 in the MARIA reactor, presented at the 22 RERTR Meeting in Budapest. A new computational tool, the REBUS-3 code, has been used for predicting the amount of poison. The code has been put into operation on a HP computer and the beryllium transmutation chains have been activated with assistance of the ANL RERTR staff. The horizontal and vertical poison distribution within beryllium blocks has been studied. A simple shuffling of beryllium blocks has been simulated to check the effect of exchanging a block with high poison concentration, adjacent to fuel elements, with a peripheral one with a low poison concentration

A technical solution to a difficult problem - An update

International Nuclear Information System (INIS)

Asselstine, J.K.

1985-01-01

A goal of the Reduced Enrichment Research and Test Reactor Program (RERTR) established by the U.S. Department of Energy is to develop the technical means needed to utilize LEU instead of HEU fuels in research and test reactors and to do so without operational penalties and without affecting safety characteristics. This goal has been pursued now for seven years in cooperation with many of you who are participants in this conference. The program has made significant progress in the development, demonstration and application of new LEU research reactor fuels. According to recent projections, the percentage of HEU fuel in international commerce which is technically replaceable with LEU fuels has been growing steadily as research continues. There appears to be no technical barrier to eventual replacement of HEU fuel with LEU fuel in research and test reactors. While there are many politically complex and technically difficult issues involved in nonproliferation, I conclude that the use of highly enriched weapons grade uranium as a fuel in civilian research reactors stands out as one problem which has a straight forward technical solution. The technical solution is the effort you in the RERTR program are embarked on. Once that solution is in hand, the operators of research reactors are urged to take political and legal initiatives to make the conversion happen. By effecting this conversion they will help minimize the trade in highly enriched uranium and make the world a little safer

A technical solution to a difficult problem - An update

Energy Technology Data Exchange (ETDEWEB)

Asselstine, J K

1985-07-01

A goal of the Reduced Enrichment Research and Test Reactor Program (RERTR) established by the U.S. Department of Energy is to develop the technical means needed to utilize LEU instead of HEU fuels in research and test reactors and to do so without operational penalties and without affecting safety characteristics. This goal has been pursued now for seven years in cooperation with many of you who are participants in this conference. The program has made significant progress in the development, demonstration and application of new LEU research reactor fuels. According to recent projections, the percentage of HEU fuel in international commerce which is technically replaceable with LEU fuels has been growing steadily as research continues. There appears to be no technical barrier to eventual replacement of HEU fuel with LEU fuel in research and test reactors. While there are many politically complex and technically difficult issues involved in nonproliferation, I conclude that the use of highly enriched weapons grade uranium as a fuel in civilian research reactors stands out as one problem which has a straight forward technical solution. The technical solution is the effort you in the RERTR program are embarked on. Once that solution is in hand, the operators of research reactors are urged to take political and legal initiatives to make the conversion happen. By effecting this conversion they will help minimize the trade in highly enriched uranium and make the world a little safer.

On the importance of ending the use of HEU in the nuclear fuel cycle: An updated assessment

International Nuclear Information System (INIS)

Glaser, Alexander; Hippel, Frank von

2002-01-01

The events of September 2001 have created a renewed urgency with regard to the disposition and future use and management of nuclear-weapons-usable materials. Highly enriched uranium (HEU) has received particular attention because it is relatively easy to use in a nuclear weapon and therefore an obvious candidate for diversion or theft by state or nonstate actors. The role of the RERTR program in this context and its contribution to global security can hardly be overemphasized. This article reviews existing or proposed activities to reduce the threat posed by HEU, how these activities are linked to the RERTR program, and outlines the most urgent steps to be taken to approach the ultimate objective of eliminating non-weapons HEU inventories in the world. (author)

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)

OSIRIS: an example of gradual conversion

International Nuclear Information System (INIS)

Guidez, J.; Beylot, J.; Joly, C.

1997-01-01

The conversion of the OSIRIS core from 'Caramel' to silicide forms part of the move towards standardising a well-known fuel, developed in the context of the RERTR programme, and used by the majority of reactors. This standardisation makes it possible to anticipate reduced manufacturing and reprocessing costs in the long term. (author)

The reduced enrichment program for JRR-4

International Nuclear Information System (INIS)

Takayanagi, M.

1992-01-01

Japan Research Reactor No. 4(JRR-4) with the rated power of 3.5 MW, swimming pool type research reactor, 93 % enriched uranium ETR-type fuel used, light water moderated and cooled. The first criticality reached on 28th January, 1965. The reactor has operated for about 26 years. However, it was planed to the reduced enrichment of the fuels to low enrichment according to the International Reduced Enrichment for Research and Test Reactors (RERTR) program. This paper describes the program for conversion of the enrichment of fuel from 93 % to less than 20 %. (author)

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

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

Conversion (Utilizing LEU instead HEU) of research reactors in Czech Republic

International Nuclear Information System (INIS)

Matejka, K.; Sklenka, L.; Listik, E.; Ernest, J.

1998-01-01

This paper shortly describes some advantages on the RERTR-programme in the Czech Republic. Further calculations and experimental measurements finished on the VR-1 training reactor Sparrow. Paper brings results and its evaluation as well as one-year operation experiences with the Russian fuel assemblies IRT-3M and also operation experiments with mixed Core configuration (the Russian fuel assemblies IRT-2M with enrichment 80% 235 U and 36% 235 U) on the LVR-15 research reactor. (author)

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

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.

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.

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)

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

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)

An alternative LEU design for the FRM-II

International Nuclear Information System (INIS)

Hanan, N.A.; Mo, S.C.; Smith, R.S.; Matos, J.E.

1996-01-01

The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm 3 and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance. LEU silicide fuel with 4.5 g/cm 3 has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. The following issues raised by TUM were addressed in Ref. 1: qualification of HEU and LEU silicide fuels, gamma heating in the heavy water reflector, radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. The conclusions of these analyses are summarized below. This paper addresses three additional safety issues that were raised by TUM in Ref. 2: stability of the involute fuel plates, a hypothetical accident involving the configuration of the reflector, and a loss of primary coolant flow transient due to an interrupted power supply. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility

The Y-12 National Security Complex Foreign Research Reactor Uranium Supply Production

Energy Technology Data Exchange (ETDEWEB)

Nelson, T. [Nuclear Technology and Nonproliferation Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States); Keller, A.P. [Disposition and Supply Programs, B and W Y-12, L.L.C., Y-12 National Security Complex, Oak Ridge, Tennessee (United States)

2011-07-01

The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the National Nuclear Security Administration (NNSA) HEU Disposition, the Reduced Enrichment Research and Test Reactors (RERTR), and the United States (U.S.) FRR Spent Nuclear Fuel (SNF) Acceptance Programs. The FRR Supply Program supports the important U.S. government nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to Low-Enriched Uranium (LEU) fuel under the RERTR Program. The NNSA Y-12 Site Office maintains the prime contracts with foreign government agencies for the supply of LEU for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. In addition to uranium metal feedstock for fuel fabrication, Y-12 can produce LEU in different forms to support new fuel development or target fabrication for medical isotope production. With production improvements and efficient delivery preparations, Y-12 continues to successfully support the global research reactor community. (author)

An alternative LEU design for the FRM-II

International Nuclear Information System (INIS)

Hanan, N.A.; Mo, S.C.; Smith, R.S.; Matos, J.E.

1997-02-01

The Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm[sup 3] and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance (8 x 10[sup 14] n/cm[sup 2]/s in the reflector). LEU silicide fuel with 4.5 g/cm[sup 3] has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Computer models for the HEU and LEU designs have been exchanged between TUM and ANL and discrepancies have been resolved. The following issues are addressed: qualification of HEU and LEU silicide fuels, stability of the fuel plates, gamma heating in the heavy water reflector, a hypothetical accident involving the configuration of the reflector, a loss of primary coolant flow transient due to an interrupted power supply, the radiological consequences of larger fission product and plutonium inventories in the LEU core, and cost and schedule. Calculations were also done to address the possibility that new high density LEU fuels could be developed that would allow conversion of the TUM HEU design to LEU fuel. Based on the excellent results for the Alternative LEU Design that were obtained in these analyses, the RERTR Program concludes that all of the major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II have been successfully resolved and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility

Supply of enriched uranium for research reactors

International Nuclear Information System (INIS)

Mueller, Hans; Laucht, Juergen

1996-01-01

Since the RERTR meeting in 1990 at Newport/USA, NUKEM recommended that the research reactor community agree upon a worldwide unified technical specification for low enriched uranium (LEU) and high enriched uranium (HEU) since there existed numerous specifications both from suppliers/fabricators and research reactors. The target recommended by NUKEM is to arrive at a worldwide unified standard specification in order to facilitate supplies of LEU and HEU to fabricators for fabrication of research reactor fuel elements. In our paper presented at the RERTR meeting at Paris in September 1995, we pointed out that LEU and HEU supplied by the U.S. Department of Energy (DOE) in the past was never 'virgin' material, i.e., it was mixed with reprocessed uranium. Our recommendation was to include this fact in the proposed unified specification. Since the RERTR meeting in 1995 progress on a unified standard specification has been made and we would like to provide more specific information about that in this paper. Furthermore, we will deal with the question whether there is a secure supply of LEU for converted research reactors. We list current and potential suppliers of LEU, noting however, that the DOE has for a number of years been unable to supply any LEU due to production problems. The future availability of LEU of U.S. origin is, of course, essential for those research reactor operators which have converted their reactors from HEU to LEU and which are intending to return spent fuel of U.S. origin to the U.S.A. (author)

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)

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

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)

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

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)

Proceedings of the international meeting on reduced enrichment for research and test reactors

Energy Technology Data Exchange (ETDEWEB)

Tsuchihashi, Keichiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1984-05-01

The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of various programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. The papers presented during the Meeting were divided into 9 sessions and one round able discussion which concluded the Meeting. The Sessions were: Program, Fuel Development, Fuel Fabrication, Irradiation testing, Safety Analysis, Special Reactor Conversion, Reactor Design, Critical Experiments, and Reprocessing and Spent Fuel Storage. Thus, topics of this Meeting were of a very wide range that was expected to result in information exchange valuable for all the participants in the RERTR program.

Proceedings of the international meeting on reduced enrichment for research and test reactors

International Nuclear Information System (INIS)

Tsuchihashi, Keichiro

1984-05-01

The purpose of the Meeting was to exchange and discuss the most up-to-date information on the progress of various programs related to research and test reactor core conversion from high enriched uranium to lower enriched uranium. The papers presented during the Meeting were divided into 9 sessions and one round able discussion which concluded the Meeting. The Sessions were: Program, Fuel Development, Fuel Fabrication, Irradiation testing, Safety Analysis, Special Reactor Conversion, Reactor Design, Critical Experiments, and Reprocessing and Spent Fuel Storage. Thus, topics of this Meeting were of a very wide range that was expected to result in information exchange valuable for all the participants in the RERTR program

Phases in U-Si alloys

International Nuclear Information System (INIS)

Domagala, R.F.

1986-09-01

The binary (two component) U-Si system contains a total of seven ''compounds.'' The most U-rich compounds are of interest to the RERTR community because they are now being employed as fuels in research and test reactors. The nomenclature used in describing these fuels and the metallurgical significance of the notations recorded may have different meanings to people from different technical backgrounds. This paper is a succinct exploration of the principles of phase equilibria and the realities of commerical fabrication as applied to U-Si alloys. It is an attempt to record in referenceable and retrievable form information of value to the continued development, application and understanding of silicide fuels

Core conversion anaylses for the Portuguese Research Reactor

International Nuclear Information System (INIS)

Matos, J.E.; Stevens, J.G.; Feldman, E.E.; Stillman, J.A.; Dunn, F.E.; Kalimullah, K.; Marques, J.G.; Barradas, N.P.; Ramos, A.R.; Kling, A.

2006-01-01

Design and safety analyses are presented for conversion of the Portuguese Research Reactor (RPI) from the use of HEU fuel to the use of LEU fuel. The analyses were performed jointly by the RERTR Program at the Argonne National Laboratory (ANL) and the Instituto Tecnologico e Nuclear (ITN). The LEU fuel assembly design uses U 3 Si 2 -Al dispersion fuel with 4.8 g U/cm 3 and is very similar to the HEU fuel design. The results of neutronic studies, steady-state thermal-hydraulic analyses, accident analyses, and revisions to the Operating Limits and Conditions demonstrate that the RPI reactor can be operated safely with the new LEU fuel assemblies. Delivery of the LEU fuel is expected around the end of 2006, with conversion in early 2007. The HEU fuel is planned to be returned to the US in 2008.

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

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

Foreign research reactor uranium supply program: The Y-12 national security complex process

International Nuclear Information System (INIS)

Nelson, T.; Eddy, B.G.

2010-01-01

The Foreign Research Reactor (FRR) Uranium Supply Program at the Y-12 National Security Complex supports the nonproliferation objectives of the HEU Disposition Program, the Reduced Enrichment Research and Test Reactors (RERTR) Program, and the United States FRR Spent Nuclear Fuel (SNF) Acceptance Program. The Y-12 National Nuclear Security Administration (NNSA) Y-12 Site Office maintains the prime contracts with foreign governments for the supply of Low-Enriched Uranium (LEU) for their research reactors. The LEU is produced by down blending Highly Enriched Uranium (HEU) that has been declared surplus to the U.S. national defense needs. The down blending and sale of the LEU supports the Surplus HEU Disposition Program Record of Decision to make the HEU non-weapons usable and to recover the economic value of the uranium to the extent feasible. This program supports the important U.S. government and nuclear nonproliferation commitment to serve as a reliable and cost-effective uranium supplier for those foreign research reactors that are converting or have converted to LEU fuel under the guidance of the NNSA RERTR Program. In conjunction with the FRR SNF Acceptance Program which supports the global nonproliferation efforts to disposition U.S.-origin HEU, the Y-12 FRR Uranium Supply Program can provide the LEU for the replacement fuel fabrication. In addition to feedstock for fuel fabrication, Y-12 supplies LEU for target fabrication for medical isotope production. The Y-12 process uses supply forecasting tools, production improvements and efficient delivery preparations to successfully support the global research reactor community

Proceedings of the 1978 international meeting on reduced enrichment for research and test reactors

Energy Technology Data Exchange (ETDEWEB)

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

1993-08-01

November 9-10, 1978, marked the first of what has become an annual event - the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The RERTR Program had been started only three months earlier, and the meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and while it established the basis for all later meetings, it was unique in several respects. It was a time of feeling each other out, and of sharing new ideas, concerns, and hopes. In the absence of an established precedent, a number of participants came with written papers while others made only verbal presentations. Informality added spice and special importance to the discussions at the end of each presentation and, especially, to the panel discussion at the end of the meeting. An important achievement was a consensus on near-, medium-, and long-term density goals for the various fuels. This consensus resulted in a list written on the blackboard at the end of the panel discussion, and reproduced on page 216, which outlined the goals of each fabricator. Luckily, both presentations and discussions were recorded on audio tape. These recordings were transcribed and used to complete the informal presentations and to append the discussions at the end of each presentation. Considerable effort was expended in clearing the transcribed papers and key discussions with the participants. A few issues could not be resolved quickly, and in the frantic rush of those early years, these proceedings were set aside. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort, we have recently dusted off the manuscript and finished our editing job.

Proceedings of the 1978 international meeting on reduced enrichment for research and test reactors

International Nuclear Information System (INIS)

Travelli, A.

1993-08-01

November 9-10, 1978, marked the first of what has become an annual event - the International Meeting on Reduced Enrichment for Research and Test Reactors (RERTR). The RERTR Program had been started only three months earlier, and the meeting brought together for the first time many people who became major program participants in later years. This first meeting emphasized fuel development, and while it established the basis for all later meetings, it was unique in several respects. It was a time of feeling each other out, and of sharing new ideas, concerns, and hopes. In the absence of an established precedent, a number of participants came with written papers while others made only verbal presentations. Informality added spice and special importance to the discussions at the end of each presentation and, especially, to the panel discussion at the end of the meeting. An important achievement was a consensus on near-, medium-, and long-term density goals for the various fuels. This consensus resulted in a list written on the blackboard at the end of the panel discussion, and reproduced on page 216, which outlined the goals of each fabricator. Luckily, both presentations and discussions were recorded on audio tape. These recordings were transcribed and used to complete the informal presentations and to append the discussions at the end of each presentation. Considerable effort was expended in clearing the transcribed papers and key discussions with the participants. A few issues could not be resolved quickly, and in the frantic rush of those early years, these proceedings were set aside. Believing that the proceedings of this first meeting are important as a historical record of the beginning of the international RERTR effort, we have recently dusted off the manuscript and finished our editing job

Postirradiation examination of high-U-loaded, low-enriched U3O8, UAl2, and U3Si test fuel plates

International Nuclear Information System (INIS)

Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofman, G.L.; Snelgrove, J.L.

1985-01-01

The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded, low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examined, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory performance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior being the slight bowing of some plates at about 80% burnup

Postirradiation examination of high-U-loaded low-enriched U3O8, UAl2, and U3Si test fuel plates

International Nuclear Information System (INIS)

Gomez, J.; Morando, R.; Perez, E.E.; Giorsetti, D.R.; Copeland, G.L.; Hofmann, G.; Snelgrove, J.L.

1984-01-01

The scope of this work is to present an evaluation of the postirradiation examination of the second set of high-U-loaded low-enriched U 3 O 8 , UAl 2 and U 3 Si miniature plates manufactured by the Comision Nacional de Energia Atomica (CNEA) of Argentina, and irradiated and examinated, within the framework of the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Oak Ridge National Laboratory and Argonne National Laboratory. This paper includes fabrication details of the plates, their irradiation history and the results of postirradiation examination which are compared to those of the previous test and to present results from other laboratories participating in the RERTR Program. Postirradiation examination of these plates showed satisfactory poerformance for the oxides, aluminides and silicides (except for the highest-loaded U 3 Si plate) with the only indication of detrimental behavior during the slight bowing of some plates at about 80% burnup

Neutronic feasibility studies for LEU conversion of the HFR Petten reactor

International Nuclear Information System (INIS)

Hanan, N.A.; Deen, J.R.; Matos, J.E.; Hendriks, J.A.; Thijssen, P.J.M.; Wijtsma, F.J.

2000-01-01

Design and safety analyses to determine an optimum LEU fuel assembly design using U 3 Si 2 -Al fuel with up to 4.8 g/cm 3 for conversion of the HFR Petten reactor were performed by the RERTR program in cooperation with the Joint Research Centre and NRG. Credibility of the calculational methods and models were established by comparing calculations with recent measurements by NRG for a core configuration set up for this purpose. This model and methodology were then used to study various LEU fissile loading and burnable poison options that would satisfy specific design criteria. (author)

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

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

Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® Reactor

International Nuclear Information System (INIS)

Schickler, R.A.; Marcum, W.R.; Reese, S.R.

2013-01-01

Highlights: • The Oregon State TRIGA ® Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA ® Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least-squares technique. The quantification of

Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA{sup ®} Reactor

Energy Technology Data Exchange (ETDEWEB)

Schickler, R.A., E-mail: robert.schickler@oregonstate.edu; Marcum, W.R., E-mail: wade.marcum@oregonstate.edu; Reese, S.R.

2013-09-15

Highlights: • The Oregon State TRIGA{sup ®} Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA{sup ®} Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least

Development and validation of an improved version of the DART code

International Nuclear Information System (INIS)

Taboada, H; Moscarda, M.V.; Markiewicz, M.; Estevez, E.; Rest, J.

2002-01-01

ANL/USDOE and CNEA Argentina have been participating within a SisterLab Program in the area of Low Enriched Uranium Advanced Fuels 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 we developed a visual version of DART named FASTDART for silicide and U-Mo fuels that was presented at the RERTR Meeting in Las Vegas, Nevada. This paper describes several major improvements in the FASTDART code: a thermal calculation subroutine, a fuel particle size distribution subroutine and several visual interfaces for thermal output plotting and particle size input. Using the power history, coolant regime data and fuel dimensions, the new thermal subroutine is able to calculate at each time step the maximum temperature along the z-longitudinal axis as a function of plate/rod morphology (corrosion oxide, cladding, meat, aluminide particle layer, each radial shell of a central fuel particle, and particle center). Calculated temperatures at each time step are coupled to the DART calculation kernel such that swelling processes, volume phase fractions and meat thermal conductivity are calculated synergistically. The new fuel particle size-distribution subroutine is essential in order to determine the evolution of the volume fraction of reaction product. This phase degrades the heat transport by a twofold mechanism: its appearance implies a diminution of aluminium phase and its thermal conductivity is lower than those of fuel and dispersant phase. The new version includes the capability of plotting thermal data output by means of the plate/rod temperature profile at a given irradiation step, and displaying the maximum temperature evolution of each layer. A comparison between the reaction layer thickness and matrix and fuel volume fractions of several RERTR-3 experiment

75 FR 4493 - Natural Resources Defense Council; Denial of Petition for Rulemaking

Science.gov (United States)

2010-01-28

... Test Reactors (RERTR) program. The goal of this program is the conversion of research and test reactors... supporting the RERTR program and stating that the NRC would act expeditiously to review the use of new LEU... scrutinized to verify that the HEU export meets U.S. statutory requirements. In 2004, the RERTR program became...

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.

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

Study on microstructure change of Uranium nitride coated U-7wt%Mo powder by heat treatment

Energy Technology Data Exchange (ETDEWEB)

Cho, Woo Hyoung; Park, Jae Soon; Lee, Hae In; Kim, Woo Jeong; Yang, Jae Ho; Park, Jong Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-10-15

Uranium-molybdenum alloy particle dispersion fuel in an aluminum matrix with a high uranium density has been developed for a high performance research reactor in the RERTR program. In order to retard the fuel-matrix interaction in U-Mo/Al dispersion fuel in which the U-Mo fuel particles were dispersed in Al matrix, nitride layer coated U-Mo fuel particle has been designed and techniques to fabricate nitride-layer coated U-7wt%Mo particles have been developed in our lab. In this study, uranium nitride coated U-Mo particle has heat treatment for several times and degree. And we suggested for interaction layer remedy in U-Mo dispersion fuel. We investigate effect of heat treatment interaction layer evolution on uranium nitride coated U-Mo powder. The EDS and XRD analysis to investigate the phase evolution in uranium nitride coated layer is also a part of the present work

Progress report on R and D results from the advanced neutron source

International Nuclear Information System (INIS)

West, C.D.

1992-01-01

This presentation consists of six parts describing the the following: Oxide Formation; U 3 Si 2 Fuel Performance; Aluminum Irradiation Properties and Code Case; Fuel Plate Hydraulic Stability; Thermal-Hydraulic Test Loop designed to examine the CHF/flow instability limits and thermal hydraulic correlations of the ANS core; Cold Source Design Concept. HFIR results indicate good performance of U 3 Si 2 for ANS conditions. Additional data from HFIR tests, RERTR fuel, and simulation experiments are expected to improve understanding of basic behavior. Further research plans for the cold neutron source are: Test the circulation system, Test beryllium fabricability and properties, Develop and test modified pressure-balanced cryostat, design, if possible, of safe continued operation of the reactor even if cold source refrigeration is lost

Methods and codes for neutronic calculations of the MARIA research reactor

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.; Bretscher, M.M.; Hanan, N.A.; Matos, J.E.

1998-01-01

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6x8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminium. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization. (author)

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)

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)

Specificity in the licensing process of reduced enrichment in the Bulgarian research reactor

International Nuclear Information System (INIS)

Vitkova, Marietta; Gorinov, Ivan

2005-01-01

The presented paper considers some specific questions of the licensing process regarding the reconstruction of the Bulgarian IRT-2000 research reactor, which includes conversion to the low enriched fuel. This specificity has risen as a result of two facts. The design of the reactor reconstruction was made on the basis of the existing fresh 36% highly enriched fuel. But after finishing of the design process, this fresh highly enriched fuel was shipped back to Russia in the framework of the RERTR program. These facts have involved some changes in both - in the licensing and the design processes. Re-analysis of the neutronic and thermal-hydraulic calculations is required to be made on the base of the technical specifications of the new LEU fuel. To facilitate the licensing process the NRA has adopted regulatory acceptance criteria for approval of the reactor core design with LEU fuel. (author)

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

Neutronic calculation of safety parameters for the RP-0 and RP-10 nuclear reactors

OpenAIRE

Lázaro, Gerardo; Deen, James R.; Woodruff, William L.

2002-01-01

Theoretical safety calculations were done with proved codes utilized by the staff of the RERTR program in the HEU to LEU core conversions. The studies were designed to evaluate the reactivity coefficients and kinetics parameters of the reactor involved in the evolution of peak power transients by reactivity insertion accidents. It was done to show the trend of these reactivity coefficients as a function of the core size and fuel depletion for RP10 cores. It was useful to get a better underst...

Russian-American Security Cooperation After St. Petersburg: Challenges and Opportunities

Science.gov (United States)

2007-04-01

each shipment. The third GTRI element, the Reduced Enrichment for Research and Test Reactors ( RERTR ) program, funds efforts to convert the cores of...IAEA in the Russian Research Reactor Return Pro- gram, the RERTR Program, and the Tripartite Initia- tive to secure high-risk radioactive sources...Russia’s involvement in the RERTR , see Charles D. Ferguson, Preventing Catastrophic Nuclear Terrorism, New York: Council on Foreign Relations, March 2006

Reducing The Nuclear Danger

Science.gov (United States)

1995-10-01

off convention • Eliminate the civil use of HEU (includes RERTR ) • Reduce stockpiles of civil HEU and plutonium • Promote alternatives to the...these countries. ANL supports the Department’s Reduced Enrichment for Research and Test Reactor ( RERTR ) Program by providing the technical means to...scientists and engineers at 60 institutes in Russia, Ukraine, Kazakhstan and Belarus. The United States and Russia have agreed to pursue a joint RERTR

A level-playing field for medical isotope production - How to phase-out reliance on HEU

International Nuclear Information System (INIS)

Kuperman, A.J.

1999-01-01

Two decades ago, civilian commerce in highly enriched uranium (HEU) for use as targets in the production of medical isotopes was considered a relatively minor security concern for three reasons. First, the number of producers was small. Second, the amount of HEU involved was small. Third, the amount of HEU was dwarfed by the quantities of HEU in civilian commerce as fuel for nuclear research and test reactors. Now, however, all three variables have changed. First, as the use of medical isotopes has expanded rapidly, production programs are proliferating. Second, as the result of such new producers and the expansion of existing production facilities, the amounts of HEU involved are growing. Third, as the RERTR program has facilitated the phase-out of HEU as fuel in most research and test reactors, the quantities of HEU for isotope production have come to represent a significant percentage of global commerce in this weapons-usable material. Medical isotope producers in several states are cooperating with the RERTR program to convert to low-enriched uranium (LEU) targets within the next few years, and one already relies on LEU for isotope production. However, the three biggest isotope producers - in Canada and the European Union - continue to rely on HEU, creating a double-standard that endangers the goal of the RERTR program. Each of these three producers has expressed economic concerns about being put at a competitive disadvantage if it alone converts. This paper proposes forging a firmer international consensus that all present and future isotope producers should convert to LEU, and calls for codifying such a commitment in a statement of intent to be prepared by producers over the next year. With such a level playing field, no producer would need fear being put at a competitive disadvantage by conversion, or being stigmatized by pressure groups for continued reliance on HEU. The phase-out of all HEU commerce for isotope production could be achieved within about

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.

Progress report on R and D results from the advanced neutron source

Energy Technology Data Exchange (ETDEWEB)

West, C D [Oak Ridge National Laboratory, Oak Ridge, TN (United States)

1992-07-01

This presentation consists of six parts describing the the following: Oxide Formation; U{sub 3}Si{sub 2} Fuel Performance; Aluminum Irradiation Properties and Code Case; Fuel Plate Hydraulic Stability; Thermal-Hydraulic Test Loop designed to examine the CHF/flow instability limits and thermal hydraulic correlations of the ANS core; Cold Source Design Concept. HFIR results indicate good performance of U{sub 3}Si{sub 2} for ANS conditions. Additional data from HFIR tests, RERTR fuel, and simulation experiments are expected to improve understanding of basic behavior. Further research plans for the cold neutron source are: Test the circulation system, Test beryllium fabricability and properties, Develop and test modified pressure-balanced cryostat, design, if possible, of safe continued operation of the reactor even if cold source refrigeration is lost.

Main activities carried out for the conversion of the reactor core TRIGA, from HEU 8.5/70 / LEU 8.5/20 to LEU 30/20; Principales actividades llevadas a cabo para la conversion del nucleo del reactor TRIGA, de HEU 8.5/70 / LEU 8.5/20 a LEU 30/20

Energy Technology Data Exchange (ETDEWEB)

Flores C, J., E-mail: jorge.floresc@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2012-10-15

In agreement with the policies of the global initiative of threats reduction (GTRI), Mexico committed that inside the reduction program of the fuel enrichment in research and test reactors (RERTR), the conversion of the core reactor TRIGA (in the nuclear centre) would be made, to use solely fuel with low enrichment ({<=} 20% U{sup 235}). To support to the execution of this commitment, a series of accords and agreements were established. The Project Agreement and Supply among the IAEA, the United States of America and Mexico was the more relevant. In this work the main activities carried out in the Instituto Nacional de Investigaciones Nucleares (ININ) with this purpose are presented. (Author)

SSPA Equipment Engineering Feasibility Report

Energy Technology Data Exchange (ETDEWEB)

N.E. Woolstenhulme; C.R. Clark

2011-09-01

In response to a demanding reactor conversion schedule, construction of the Shielded Sample Preparation Area (SSPA) was initiated in 2010 to augment the existing capabilities of the Hot Fuel Examination Facility (HFEF). While HFEF is and will remain the workhorse for post irradiation sample preparation, there is currently a large backlog of Post-Irradiation Examination (PIE) experiments caused by numerous competing projects (this backlog is expected to continue for the foreseeable future). HFEF, in its present configuration also lacks the ability to prepare samples suitable for several of the tests that have been identified for the successful conclusion of the RERTR program; these samples require fine detail machining of irradiated fuel plates.

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.

Feasibility study to restart the research reactor RA with a converted core

International Nuclear Information System (INIS)

Matausek, M.V.; Plecas, I.; Marinkovic, N.

1999-01-01

Main options are specified for the future status of the 6.5 MW heavy water research reactor RA. Arguments pro and contra restarting the reactor are presented. When considering the option to restart the RA reactor, possibilities to improve its neutronic parameters, such as neutron flux values and irradiation capabilities, are discussed, as well as the compliance with the worldwide activities of Reduced Enrichment for Research and Test Reactors (RERTR) program. Possibility of core conversion is examined. Detailed reactor physics design calculations are performed for different fuel types and uranium loading. For different fuel management schemes results are presented for the effective multiplication factor, power distribution, fuel burnup and consumption. It is shown that, as far as reactor core parameters are considered, conversion to lower enrichment fuel could be easily accomplished. However, conversion to the lower enrichment could only be justified if combined with improvement of some other reactor attributes. (author)

Design study of eventual core conversion for the research reactor RA

International Nuclear Information System (INIS)

Matausek, M. V.; Marinkovic, N.

1998-01-01

Main options are specified for the future status of the 6.5 MW heavy water research reactor RA. Arguments pro and contra restarting the reactor are presented. When considering the option to restart the RA reactor, possibilities to improve its neutronic parameters, such as neutron flux values and irradiation capabilities are discussed, as well as the compliance with the worldwide activities of Reduced Enrichment for Research and Test Reactors (RERTR) program. Possibility of core conversion is examined. Detailed reactor physics design calculations are performed for different fuel types and uranium loading. For different fuel management schemes results are presented for the effective, multiplication factor, power distribution, fuel burnup and consumption. It is shown that, as far as reactor core parameters are considered, conversion to lower enrichment fuel could be easily accomplished. However, conversion to the lower enrichment could only be justified if combined with improvement of some other reactor attributes. (author)

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.

Fluxes at experiment facilities in HEU and LEU designs for the FRM-II

International Nuclear Information System (INIS)

Hanan, N. A.

1998-01-01

An Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm 3 and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime(50 days) and the same neutron flux performance (8 x 10 14 n/cm 2 -s in the reflector). LEU silicide fuel with 4.5 g/cm 3 has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Several issues that were raised by TUM have been addressed in Refs. 1-3. The conclusions of these analyses are summarized below. This paper addresses four additional issues that have been raised in several forums, including Ref 4: heat generation in the cold neutron source (CNS), the gamma and fast neutron fluxes which are components of the reactor noise in neutron scattering experiments in the experiment hall of the reactor, a fuel cycle length difference, and the reactivity worth of the beam tubes and other experiment facilities. The results show that: (a) for the same thermal neutron flux, the neutron and gamma heating in the CNS is smaller in the LEU design than in the HEU design, and cold neutron fluxes as good or better than those of the HEU design can be obtained with the LEU design; (b) the gamma and fast neutron components of the reactor noise in the experiment hall are about the same in both designs; (c) the fuel cycle length is 50 days for both designs; and (d) the absolute value of the reactivity worth of the beam tubes and other experiment facilities is smaller in the LEU design, allowing its fuel cycle length to be increased to 53 or 54 days. Based on the excellent results for the Alternative LEU Design that were obtained in all analyses, the RERTR Program reiterates its conclusion that there are no major technical

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.

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.

Supply of enriched uranium for research reactors

Energy Technology Data Exchange (ETDEWEB)

Mueller, H. [NUKEM GmbH, Alzenau (Germany)

1997-08-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

Supply of enriched uranium for research reactors

International Nuclear Information System (INIS)

Mueller, H.

1997-01-01

Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA's ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future

TEM Characterization of High Burn-up Microstructure of U-7Mo Alloy

Energy Technology Data Exchange (ETDEWEB)

Jian Gan; Brandon Miller; Dennis Keiser; Adam Robinson; James Madden; Pavel Medvedev; Daniel Wachs

2014-04-01

As an essential part of global nuclear non-proliferation effort, the RERTR program is developing low enriched U-Mo fuels (< 20% U-235) for use in research and test reactors that currently employ highly enriched uranium fuels. One type of fuel being developed is a dispersion fuel plate comprised of U-7Mo particles dispersed in Al alloy matrix. Recent TEM characterizations of the ATR irradiated U-7Mo dispersion fuel plates include the samples with a local fission densities of 4.5, 5.2, 5.6 and 6.3 E+21 fissions/cm3 and irradiation temperatures of 101-136?C. The development of the irradiated microstructure of the U-7Mo fuel particles consists of fission gas bubble superlattice, large gas bubbles, solid fission product precipitates and their association to the large gas bubbles, grain subdivision to tens or hundreds of nanometer size, collapse of bubble superlattice, and amorphisation. This presentation will describe the observed microstructures specifically focusing on the U-7Mo fuel particles. The impact of the observed microstructure on the fuel performance and the comparison of the relevant features with that of the high burn-up UO2 fuels will be discussed.

Nuclear characteristics evaluation for Kyoto University Research Reactor with low-enriched uranium core

Energy Technology Data Exchange (ETDEWEB)

Nakajima, Ken; Unesaki, Hironobu [Kyoto University Research Reactor Institute, Kumatori-cho Sennan-gun Osaka (Japan)

2008-07-01

A project to convert the fuel of Kyoto University Research Reactor (KUR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) is in progress as a part of RERTR program. Prior to the operation of LEU core, the nuclear characteristics of the core have been evaluated to confirm the safety operation. In the evaluation, nuclear parameters, such as the excess reactivity, shut down margin control rod worth, reactivity coefficients, were calculated, and they were compared with the safety limits. The results of evaluation show that the LEU core is able to satisfy the safety requirements for operation, i.e. all the parameters satisfy the safety limits. Consequently, it was confirmed that the LEU fuel core has the proper nuclear characteristics for the safety operation. (authors)

Kr ion irradiation study of the depleted-uranium alloys

Science.gov (United States)

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

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

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

Reduced enrichment for research and test reactors: Proceedings

Energy Technology Data Exchange (ETDEWEB)

1988-05-01

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

Reduced enrichment for research and test reactors: Proceedings

International Nuclear Information System (INIS)

1988-05-01

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

Worldwide Emerging Environmental Issues Affecting the U.S. Military. April 2004 - April 2005

Science.gov (United States)

2005-04-01

Reactors ( RERTR ) held in Vienna, November 8-11. [November 2004. Military Implications and Sources] Revival of Nuclear Power in Asia Poses Security...Military Implications: If not already done, the military should contact the RERTR Program managed by the Argonne National Laboratory of the U.S

Cooperative Threat Reduction for a New Era

Science.gov (United States)

2004-09-01

Reactor From Inr-Pitesti,” Institute for Nuclear Research (Romania), Papers Presented by ANL at the RERTR Meeting, 1997 <http://www.td.anl.gov/Programs... RERTR /Analysis97/CToma-abs.html> as of 30 July 2004. 51 This program often works in conjunction with the second proposal above. Although the

Fabrication of high-uranium-loaded U{sub 3}O{sub 8}-Al developmental fuel plates

Energy Technology Data Exchange (ETDEWEB)

Copeland, G L; Martin, M M [Oak Ridge National Laboratory, TN (United States)

1983-08-01

A common plate-type fuel for Research and Test Reactors (RERTR) is U{sub 3}0{sub 8} dispersed in aluminum and clad with an aluminum alloy. There is an impetus to reduce the {sup 235}U enrichment from above 90% to below 20% for these fuels to lessen the risk of diversion of the uranium for non-peaceful uses. Thus, the uranium content of the fuel plates has to be increased to maintain the performance of the reactors. This paper describes work at ORNL to determine the maximal uranium loading for these fuels that can be fabricated with commercially proven materials and techniques and that can be expected to perform satisfactorily in service. We fabricated developmental fuel plates with cores containing from 60 to 100 wt U{sub 3}0{sub 8} in aluminum encapsulated in 6061 aluminum alloy and evaluated them for aspects of fabricability, nondestructive testing, and expected performance. We recommend 75 wt U{sub 3}0{sub 8}-Al 3.1 Mg U/m{sup 3}) as the highest loading in the initial irradiation test. This upper limit is based on a qualitative assessment of the mechanical integrity of the core made by using current fabrication techniques and materials. As the oxide loading is increased beyond this point, planar areas and extensive stringers of oxide and voids develop, which leave little strength in the thickness direction. Fuel plates may then blister over these areas as fission gases collect during irradiation. Current size plates are easily fabricable to the 75 wt % U{sub 3}0{sub 8}-Al core loading by current fabrication techniques. Dogboning is a potential problem at this loading for some applications; however, this can be easily solved by using tapered compact ends. Current nondestructive radiography and transmission x-ray scanning are applicable to the highly loaded plates. Ultrasonic testing for non-bonds is marginal because of the abrupt change in conductance at the cladding-core interface. Plate thickness can be increased if desired; we fabricated 75 wt % plates with

Neutronic calculations for the conversion of the University of Florida Training Reactor from HEU to LEU fuel

International Nuclear Information System (INIS)

Dugan, E.T.; Diaz, N.J.; Kniedler, G.S.

1983-01-01

The University of Florida Training Reactor (UFTR) is located on the University of Florida campus in Gainesville, Florida. The reactor is the Argonaut type, heterogeneous in design and currently fueled with 93% enriched, uranium-aluminum alloy MTR plate-type fuel. Investigations are being performed to examine te feasibility of replacing the highly-enriched fuel of the current UFTR with 4.8% enriched, cylindrical pin SPERT fuel. The SPERT fuel is stainless steel clad and contains uranium dioxide (UO 2 ) pellets. On a broad spectrum, training reactor conversion from high enrichment uranium (HEU) to low enrichment uranium (LEU) fueled facilities has been a continuing concern in the International Atomic Energy Agency (IAEA) and significant work has been done in this area by the Argonne RERTR Program. The International Atomic Energy Agency cites three reasons for reactor conversion to low-enriched uranium. The main reason is the desire to reduce the proliferation potential of research reactor fuels. The second is to increase the assurance of continued fuel availability in the face of probable restrictions on the supply of highly-enriched uranium. The third reason is the possible reduction in requirements for physical security measures during fabrication, transportation, storage and use. This same IAEA report points out that the three reasons stated for the conversion of the fuel of research reactors are interrelated and cannot be considered individually. The concerns of the Nuclear Engineering Sciences Department at the University of Florida relating to the HEU fuel of the UFTR coincide with those of the International Atomic Energy Agency. The primary reason for going to low-enriched pin-type fuel is the concern with proliferation provoked by the highly-enriched plate fuel which has led to tighter security of nuclear facilities such as the UFTR. A second reason for changing to the pin-type fuel is because of difficulties that are being encountered in the supply of the

Neutronic calculations for the conversion of the University of Florida Training Reactor from HEU to LEU fuel

Energy Technology Data Exchange (ETDEWEB)

Dugan, E T; Diaz, N J [Department of Nuclear Engineering Sciences, University of Florida, Gainesville, FL (United States); Kniedler, G S [Reactor Analysis Group, TVA, Chattanooga, TN (United States)

1983-09-01

The University of Florida Training Reactor (UFTR) is located on the University of Florida campus in Gainesville, Florida. The reactor is the Argonaut type, heterogeneous in design and currently fueled with 93% enriched, uranium-aluminum alloy MTR plate-type fuel. Investigations are being performed to examine te feasibility of replacing the highly-enriched fuel of the current UFTR with 4.8% enriched, cylindrical pin SPERT fuel. The SPERT fuel is stainless steel clad and contains uranium dioxide (UO{sub 2}) pellets. On a broad spectrum, training reactor conversion from high enrichment uranium (HEU) to low enrichment uranium (LEU) fueled facilities has been a continuing concern in the International Atomic Energy Agency (IAEA) and significant work has been done in this area by the Argonne RERTR Program. The International Atomic Energy Agency cites three reasons for reactor conversion to low-enriched uranium. The main reason is the desire to reduce the proliferation potential of research reactor fuels. The second is to increase the assurance of continued fuel availability in the face of probable restrictions on the supply of highly-enriched uranium. The third reason is the possible reduction in requirements for physical security measures during fabrication, transportation, storage and use. This same IAEA report points out that the three reasons stated for the conversion of the fuel of research reactors are interrelated and cannot be considered individually. The concerns of the Nuclear Engineering Sciences Department at the University of Florida relating to the HEU fuel of the UFTR coincide with those of the International Atomic Energy Agency. The primary reason for going to low-enriched pin-type fuel is the concern with proliferation provoked by the highly-enriched plate fuel which has led to tighter security of nuclear facilities such as the UFTR. A second reason for changing to the pin-type fuel is because of difficulties that are being encountered in the supply of

The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

Energy Technology Data Exchange (ETDEWEB)

Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

2008-07-15

This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

SEM Characterization of the High Burn-up Microstructure of U-7Mo Alloy

Energy Technology Data Exchange (ETDEWEB)

Dennis D. Keiser, Jr.; Jan-Fong Jue; Jian Gan; Brandon Miller; Adam Robinson; Pavel Medvedev; James Madden; Dan Wachs; M. Teague

2014-04-01

During irradiation, the microstructure of U-7Mo evolves until at a fission density near 5x1021 f/cm3 a high-burnup microstructure exists that is very different than what was observed at lower fission densities. This microstructure is dominated by randomly distributed, relatively large, homogeneous fission gas bubbles. The bubble superlattice has collapsed in many microstructural regions, and the fuel grain sizes, in many areas, become sub-micron in diameter with both amorphous fuel and crystalline fuel present. Solid fission product precipitates can be found inside the fission gas bubbles. To generate more information about the characteristics of the high-fission density microstructure, three samples irradiated in the RERTR-7 experiment have been characterized using a scanning electron microscope equipped with a focused ion beam. The FIB was used to generate samples for SEM imaging and to perform 3D reconstruction of the microstructure, which can be used to look for evidence of possible fission gas bubble interlinkage.

Development of a high density fuel based on uranium-molybdenum alloys with high compatibility in high temperatures

OpenAIRE

OLIVEIRA, FABIO B.V. de

2014-01-01

Este trabalho tem como objetivo o desenvolvimento de um combustível nuclear de alta densidade e baixo enriquecimento com base na liga ?-UMo, para aplicações nas quais seja necessário desempenho satisfatório a altas temperaturas, considerando-se a sua utilização na forma de dispersão. Para tanto, partiu-se da análise dos resultados dos testes RERTR (sigla em inglês para \\"Reduced Enrichment of Research and Test Reactors\\") e de alguns trabalhos teóricos envolvendo a elaboração de ligas metaest...

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.

Kr ion irradiation study of the depleted-uranium alloys

Energy Technology Data Exchange (ETDEWEB)

Gan, J., E-mail: Jian.Gan@inl.go [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Keiser, D.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Miller, B.D. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Kirk, M.A.; Rest, J. [Argonne National Laboratory, 9700 South Cass Ave., Argonne, IL 60439 (United States); Allen, T.R. [University of Wisconsin, 1500 Engineering Drive, Madison, WI 53706 (United States); Wachs, D.M. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si){sub 3}, (U, Mo)(Al, Si){sub 3}, UMo{sub 2}Al{sub 20}, U{sub 6}Mo{sub 4}Al{sub 43} and UAl{sub 4}. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 {sup o}C to ion doses up to 2.5 x 10{sup 19} ions/m{sup 2} ({approx}10 dpa) with an Kr ion flux of 10{sup 16} ions/m{sup 2}/s ({approx}4.0 x 10{sup -3} dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

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.

Neutronic and thermo-hydraulic design of LEU core for Japan Research Reactor 4

International Nuclear Information System (INIS)

Arigane, Kenji; Watanabe, Shukichi; Tsuruta, Harumichi

1988-04-01

As a part of the Reduced Enrichment Research and Test Reactor (RERTR) program in JAERI, the enrichment reduction for Japan Research Reactor 4 (JRR-4) is in progress. A fuel element using a 19.75 % enriched UAlx-Al dispersion type with a uranium density of 2.2 g/cm 3 was designed as the LEU fuel and the neutronic and thermo-hydraulic performances of the LEU core were compared with those of the current HEU core. The results of the neutronic design are as follows: (1) the excess reactivity of the LEU core becomes about 1 % Δk/k less, (2) the thermal neutron flux in the fuel region decreases about 25 % on the average, (3) the thermal neutron fluxes in the irradiation pipes are almost the same and (4) the core burnup lifetime becomes about 20 % longer. The thermo-hydraulic design also shows that: (1) the fuel plate surface temperature decreases about 10 deg C due to the increase of the number of fuel plates and (2) the temperature margin with respect to the ONB temperature increases. Therefore, it is confirmed that the same utilization performance as the HEU core is attainable with the LEU core. (author)

US enrichment reduction studies

International Nuclear Information System (INIS)

1979-06-01

A major national program, the Reduced Enrichment Research and Test Reactor (RERTR) Program, is currently under way in the U.S., centered at the Argonne National Laboratory (ANL), to reduce the potential of research and test reactor fuels for increasing the proliferation of nuclear explosive devices. The main objective of the program is to provide the technical means by which the uranium enrichment to be used in these reactors can be reduced to less than 20% without significant economic and performance penalties. The criteria, basis and goals of the program are consistent with the results of a number of case studies which have been performed as part of the program

Evaluation of the use of nodal methods for MTR neutronic analysis

Energy Technology Data Exchange (ETDEWEB)

Reitsma, F.; Mueller, E.Z.

1997-08-01

Although modern nodal methods are used extensively in the nuclear power industry, their use for research reactor analysis has been very limited. The suitability of nodal methods for material testing reactor analysis is investigated with the emphasis on the modelling of the core region (fuel assemblies). The nodal approach`s performance is compared with that of the traditional finite-difference fine mesh approach. The advantages of using nodal methods coupled with integrated cross section generation systems are highlighted, especially with respect to data preparation, simplicity of use and the possibility of performing a great variety of reactor calculations subject to strict time limitations such as are required for the RERTR program.

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

Energy Technology Data Exchange (ETDEWEB)

1993-07-01

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

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

International Nuclear Information System (INIS)

1993-07-01

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

Shipment security update - 2003

International Nuclear Information System (INIS)

Patterson, John; Anne, Catherine

2003-01-01

At the 2002 RERTR, NAC reported on the interim measures taken by the U.S. Nuclear Regulatory Commission to enhance the security afforded to shipments of spent nuclear fuel. Since that time, there have been a number of additional actions focused on shipment security including training programs sponsored by the U.S. Department of Transportation and the Electric Power Research Council, investigation by the Government Accounting Office, and individual measures taken by shippers and transportation agents. The paper will present a status update regarding this dynamic set of events and provide an objective assessment of the cost, schedule and technical implications of the changing security landscape. (author)

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.

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.)

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

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)

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

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.

ANL progress in developing an LEU target and process for Mo-99 production: Cooperation with CNEA

International Nuclear Information System (INIS)

Gelis, A.V.; Vandegrift, G.F.; Aase, S.B.; Bakel, A.J.; Falkenberg, J.R.; Regalbuto, M.C.; Quigley, K.J.

2003-01-01

The primary mission of the Reduced Enrichment in Research and Test Reactors (RERTR) Program is to facilitate the conversion of research and test-reactor fuel and targets from high-enriched uranium (HEU) to low-enriched uranium (LEU). One of the current goals at Argonne National Laboratory (ANL) is to assist the Argentine Comision Nacional de Energia Atomica (CNEA) in developing an LEU foil target and a process for 99 Mo production. Specifically addressed in this paper is ANL R and D related to this conversion: (1) designing a prototype production vessel for digesting irradiated LEU foils in alkaline solutions and (2) developing a new digestion method to address all issues related to HEU to LEU conversion. (author)

Fuel performance and operation experience of WWER-440 fuel in improved fuel cycle

International Nuclear Information System (INIS)

Gagarinski, A.; Proselkov, V.; Semchenkov, Yu.

2007-01-01

The paper summarizes WWER-440 second-generation fuel operation experience in improved fuel cycles using the example of Kola NPP units 3 and 4. Basic parameters of fuel assemblies, fuel rods and uranium-gadolinium fuel rods, as well as the principal neutronic parameters and burn-up achieved in fuel assemblies are presented. The paper also contains some data concerning the activity of coolant during operation (Authors)

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.

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

The status of HEU and LEU core conversion activities at the Jamaica SLOWPOKE

Energy Technology Data Exchange (ETDEWEB)

Preston, J.; Grant, C., E-mail: john.preston@uwimona.edu.jm [Univ. of the West Indies, Mona Campus, International Centre for Environmental and Nuclear Sciences, Kingston (Jamaica)

2013-07-01

The SLOWPOKE reactor in Jamaica has been operated by the International Centre for Environmental and Nuclear Sciences, University of the West Indies since 1984, mainly for the purpose of Neutron Activation Analysis. The HEU core with current utilization has another 14 years of operation, before the addition of a large beryllium annulus would be required to further extend the life-time by 15 years. However, in keeping with the spirit of the Reduced Enrichment for Research and Test Reactors (RERTR) program, the decision was taken in 2003 to convert the core from HEU to LEU, inline with those at the Ecole Polytechnic and RMC SLOWPOKE facilities. This paper reports on the current status of the conversion activities, including key fuel manufacture and regulatory issues, which have seen substantial progress during the last year. A timetable for the complete process is given, and provided that the fuel fabrication can be completed in the estimated 18 months, the core conversion should be accomplished by the end of 2014. (author)

Characterization of a U-Mo alloy subjected to direct hydriding of the gamma phase

International Nuclear Information System (INIS)

Balart, Silvia N.; Bruzzoni, Pablo; Granovsky, Marta S.

2003-01-01

The Reduced Enrichment for Research and Test Reactors (RERTR) program has imposed the need to develop plate-type fuel elements based on high density uranium compounds, such as U-Mo alloys. One of the steps in the fabrication of the fuel elements is the pulverization of the fissile material. In the case of the U-Mo alloys, the pulverization can be accomplished through hydriding - dehydriding. Two alternative methods of the hydriding-dehydriding process, namely the selective hydriding in alpha phase (HS-alpha) and the massive hydriding in gamma phase (HM-gamma) are currently being studied at the Comision Nacional de Energia Atomica. The HM-gamma method was reproduced at laboratory scale starting from a U-7 wt % Mo alloy. The hydrided and dehydrided materials were characterized using metallographic techniques, scanning electron microscopy, energy dispersive X-ray analysis and X-ray diffraction. These results are compared with previous results of the HS-alpha method. (author)

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...

Multi-fuel reformers for fuel cells used in transportation. Phase 1: Multi-fuel reformers

Science.gov (United States)

1994-05-01

DOE has established the goal, through the Fuel Cells in Transportation Program, of fostering the rapid development and commercialization of fuel cells as economic competitors for the internal combustion engine. Central to this goal is a safe feasible means of supplying hydrogen of the required purity to the vehicular fuel cell system. Two basic strategies are being considered: (1) on-board fuel processing whereby alternative fuels such as methanol, ethanol or natural gas stored on the vehicle undergo reformation and subsequent processing to produce hydrogen, and (2) on-board storage of pure hydrogen provided by stationary fuel processing plants. This report analyzes fuel processor technologies, types of fuel and fuel cell options for on-board reformation. As the Phase 1 of a multi-phased program to develop a prototype multi-fuel reformer system for a fuel cell powered vehicle, the objective of this program was to evaluate the feasibility of a multi-fuel reformer concept and to select a reforming technology for further development in the Phase 2 program, with the ultimate goal of integration with a DOE-designated fuel cell and vehicle configuration. The basic reformer processes examined in this study included catalytic steam reforming (SR), non-catalytic partial oxidation (POX) and catalytic partial oxidation (also known as Autothermal Reforming, or ATR). Fuels under consideration in this study included methanol, ethanol, and natural gas. A systematic evaluation of reforming technologies, fuels, and transportation fuel cell applications was conducted for the purpose of selecting a suitable multi-fuel processor for further development and demonstration in a transportation application.

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.)

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.)

Uranium silicide activities at Babcock and Wilcox

International Nuclear Information System (INIS)

Noel, W.W.; Freim, J.B.

1983-01-01

Babcock and Wilcox, Naval Nuclear Fuel Division (NNFD) in conjunction with Argonne National Laboratory (ANL) is actively involved in the Reduced Enrichment Research Test Reactor (RERTR) Program to produce low enriched fuel elements for research reactors. B and W and ANL have undertaken a joint effort in which NNFD will fabricate two low enriched uranium (LEU), Oak Ridge Reactor (ORR) elements with uranium silicide fuel furnished by ANL. These elements are being fabricated for irradiation testing at Oak Ridge National Laboratory (ORNL). Concurrently with this program, NNFD is developing and implementing the uranium silicide and uranium aluminide fuel fabrication technology. NNFD is fabricating the uranium silicide ORR elements in a two-phase program, Development and Production. To summarize: 1. Full size fuel plates can be made with U 3 SiAl but the fabricator must prevent oxidation of the compact prior to hot roll bonding; 2. Providing the ANL U 3 Si x irradiation results are successful, NNFD plans to provide two ORR elements during February 1983; 3. NNFD is developing and implementing U 3 Si x and UAI x fuel fabrication technology to be operational in 1983; 4. NNFD can supply U 3 O 8 high enriched uranium (HEU) or low enriched uranium (LEU) research reactor elements; 5. NNFD is capable of providing high quality, cost competitive LEU or HEU research reactor elements to meet the needs of the customer

Features of fuel performance at high fuel burnups

International Nuclear Information System (INIS)

Proselkov, V.N.; Scheglov, A.S.; Smirnov, A.V.; Smirnov, V.P.

2001-01-01

Some features of fuel behavior at high fuel burnups, in particular, initiation and development of rim-layer, increase in the rate of fission gas release from the fuel and increase in the inner gas pressure in the fuel rod are briefly described. Basing on the analysis of the data of post-irradiation examinations of fuel rods of WWER-440 working FA and CR fuel followers, that have been operated for five fuel cycles and got the average fuel burnup or varies as 50MW-day/kgU, a conclusion is made that the WWER-440 fuel burnup can be increased at least to average burnups of 55-58 MW-day/kgU per fuel assembly (Authors)

Fuel Handbook[Wood and other renewable fuels

Energy Technology Data Exchange (ETDEWEB)

Stroemberg, Birgitta [TPS Termiska Processer AB, Nykoeping (SE)] (ed.)

2006-03-15

This handbook on renewable fuels is intended for power and heat producers in Sweden. This fuel handbook provides, from a plant owner's perspective, a method to evaluate different fuels on the market. The fuel handbook concerns renewable fuels (but does not include household waste) that are available on the Swedish market today or fuels that have potential to be available within the next ten years. The handbook covers 26 different fuels. Analysis data, special properties, operating experiences and literature references are outlined for each fuel. [Special properties, operating experiences and literature references are not included in this English version] The handbook also contains: A proposed methodology for introduction of new fuels. A recommendation of analyses and tests to perform in order to reduce the risk of problems is presented. [The recommendation of analyses and tests is not included in the English version] A summary of relevant laws and taxes for energy production, with references to relevant documentation. [Only laws and taxes regarding EU are included] Theory and background to evaluate a fuel with respect to combustion, ash and corrosion properties and methods that can be used for such evaluations. Summary of standards, databases and handbooks on biomass fuels and other solid fuels, and links to web sites where further information about the fuels can be found. The appendices includes: A methodology for trial firing of fuels. Calculations procedures for, amongst others, heating value, flue gas composition, key number and free fall velocity [Free fall velocity is not included in the English version]. In addition, conversion routines between different units for a number of different applications are provided. Fuel analyses are presented in the appendix. (The report is a translation of parts of the report VARMEFORSK--911 published in 2005)

Benefits of barrier fuel on fuel cycle economics

International Nuclear Information System (INIS)

Crowther, R.L.; Kunz, C.L.

1988-01-01

Barrier fuel rod cladding was developed to eliminate fuel rod failures from pellet/cladding stress/corrosion interaction and to eliminate the associated need to restrict the rate at which fuel rod power can be increased. The performance of barrier cladding has been demonstrated through extensive testing and through production application to many boiling water reactors (BWRs). Power reactor data have shown that barrier fuel rod cladding has a significant beneficial effect on plant capacity factor and plant operating costs and significantly increases fuel reliability. Independent of the fuel reliability benefit, it is less obvious that barrier fuel has a beneficial effect of fuel cycle costs, since barrier cladding is more costly to fabricate. Evaluations, measurements, and development activities, however, have shown that the fuel cycle cost benefits of barrier fuel are large. This paper is a summary of development activities that have shown that application of barrier fuel significantly reduces BWR fuel cycle costs

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.)

Framing car fuel efficiency : linearity heuristic for fuel consumption and fuel-efficiency ratings

NARCIS (Netherlands)

Schouten, T.M.; Bolderdijk, J.W.; Steg, L.

2014-01-01

People are sensitive to the way information on fuel efficiency is conveyed. When the fuel efficiency of cars is framed in terms of fuel per distance (FPD; e.g. l/100 km), instead of distance per units of fuel (DPF; e.g. km/l), people have a more accurate perception of potential fuel savings. People

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)

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.)

Fuel performance of DOE fuels in water storage

International Nuclear Information System (INIS)

Hoskins, A.P.; Scott, J.G.; Shelton-Davis, C.V.; McDannel, G.E.

1993-01-01

Westinghouse Idaho Nuclear Company operates the Idaho Chemical Processing Plant (ICPP) at the Idaho National Engineering Laboratory. In April of 1992, the U.S. Department of Energy (DOE) decided to end the fuel reprocessing mission at ICPP. Fuel performance in storage received increased emphasis as the fuel now needs to be stored until final dispositioning is defined and implemented. Fuels are stored in four main areas: an original underwater storage facility, a modern underwater storage facility, and two dry fuel storage facilities. As a result of the reactor research mission of the DOE and predecessor agencies, the Energy Research and Development Administration and the Atomic Energy Commission, many types of nuclear fuel have been developed, used, and assigned to storage at the ICPP. Fuel clad with stainless steel, zirconium, aluminum, and graphite are represented. Fuel matrices include uranium oxide, hydride, carbide, metal, and alloy fuels, resulting in 55 different fuel types in storage. Also included in the fuel storage inventory is canned scrap material

Fuel Supply Defaults for Regional Fuels and Fuel Wizard Tool in MOVES201X

Science.gov (United States)

The fuel supply report documents the data and methodology used to derive the default gasoline, diesel and fuel-blend fuel properties, and their respective fuel market share in MOVES. The default market share of the individual fuels varies by calendar year, seasons, and several do...

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...

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

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

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

Fuel economy of hybrid fuel-cell vehicles

Science.gov (United States)

Ahluwalia, Rajesh K.; Wang, X.; Rousseau, A.

The potential improvement in fuel economy of a mid-size fuel-cell vehicle by combining it with an energy storage system has been assessed. An energy management strategy is developed and used to operate the direct hydrogen, pressurized fuel-cell system in a load-following mode and the energy storage system in a charge-sustaining mode. The strategy places highest priority on maintaining the energy storage system in a state where it can supply unanticipated boost power when the fuel-cell system alone cannot meet the power demand. It is found that downsizing a fuel-cell system decreases its efficiency on a drive cycle which is compensated by partial regenerative capture of braking energy. On a highway cycle with limited braking energy the increase in fuel economy with hybridization is small but on the stop-and-go urban cycle the fuel economy can improve by 27%. On the combined highway and urban drive cycles the fuel economy of the fuel-cell vehicle is estimated to increase by up to 15% by hybridizing it with an energy storage system.

Ammonia as a Suitable Fuel for Fuel Cells

International Nuclear Information System (INIS)

Lan, Rong; Tao, Shanwen

2014-01-01

Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5 wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

Ammonia as a suitable fuel for fuel cells

Directory of Open Access Journals (Sweden)

Rong eLan

2014-08-01

Full Text Available Ammonia, an important basic chemical, is produced at a scale of 150 million tons per year. Half of hydrogen produced in chemical industry is used for ammonia production. Ammonia containing 17.5wt% hydrogen is an ideal carbon-free fuel for fuel cells. Compared to hydrogen, ammonia has many advantages. In this mini-review, the suitability of ammonia as fuel for fuel cells, the development of different types of fuel cells using ammonia as the fuel and the potential applications of ammonia fuel cells are briefly reviewed.

Safety analysis of MOX fuels by fuel performance code

Energy Technology Data Exchange (ETDEWEB)

Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)

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.)

Emergency fuels utilization guidebook. Alternative Fuels Utilization Program

Energy Technology Data Exchange (ETDEWEB)

1980-08-01

The basic concept of an emergency fuel is to safely and effectively use blends of specification fuels and hydrocarbon liquids which are free in the sense that they have been commandeered or volunteered from lower priority uses to provide critical transportation services for short-duration emergencies on the order of weeks, or perhaps months. A wide variety of liquid hydrocarbons not normally used as fuels for internal combustion engines have been categorized generically, including limited information on physical characteristics and chemical composition which might prove useful and instructive to fleet operators. Fuels covered are: gasoline and diesel fuel; alcohols; solvents; jet fuels; kerosene; heating oils; residual fuels; crude oils; vegetable oils; gaseous fuels.

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...

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

Apparatus for loading fuel pellets in fuel rods

International Nuclear Information System (INIS)

Tedesco, R.J.

1976-01-01

An apparatus is disclosed for loading fuel pellets into fuel rods for a nuclear reactor including a base supporting a table having grooves therein for holding a multiplicity of pellets. Multiple fuel rods are placed in alignment with grooves in the pellet table and a guide member channels pellets from the table into the corresponding fuel rods. To effect movement of pellets inside the fuel rods without jamming, a number of electromechanical devices mounted on the base have arms connected to the lower surface of the fuel rod table which cyclically imparts a reciprocating arc motion to the table for moving the fuel pellets longitudinally of and inside the fuel rods. These electromechanical devices include a solenoid having a plunger therein connected to a leaf type spring, the arrangement being such that upon energization of the solenoid coil, the leaf spring moves the fuel rod table rearwardly and downwardly, and upon deenergization of the coil, the spring imparts an upward-forward movement to the table which results in physical displacement of fuel pellets in the fuel rods clamped to the table surface. 8 claims, 6 drawing figures

Hardened over-coating fuel particle and manufacture of nuclear fuel using its fuel particle

International Nuclear Information System (INIS)

Yoshimuda, Hideharu.

1990-01-01

Coated-fuel particles comprise a coating layer formed by coating ceramics such as silicon carbide or zirconium carbide and carbons, etc. to a fuel core made of nuclear fuel materials. The fuel core generally includes oxide particles such as uranium, thorium and plutonium, having 400 to 600 μm of average grain size. The average grain size of the coated-fuel particle is usually from 800 to 900 μm. The thickness of the coating layer is usually from 150 to 250 μm. Matrix material comprising a powdery graphite and a thermosetting resin such as phenol resin, etc. is overcoated to the surface of the coated-fuel particle and hardened under heating to form a hardened overcoating layer to the coated-fuel particle. If such coated-fuel particles are used, cracks, etc. are less caused to the coating layer of the coated-fuel particles upon production, thereby enabling to prevent the damages to the coating layer. (T.M.)

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)

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

Energy Technology Data Exchange (ETDEWEB)

1993-07-01

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

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

International Nuclear Information System (INIS)

1993-07-01

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

Fuel Cell Power Plants Renewable and Waste Fuels

Science.gov (United States)

2011-01-13

logo, Direct FuelCell and “DFC” are all registered trademarks (®) of FuelCell Energy, Inc. Applications •On-site self generation of combined heat... of FuelCell Energy, Inc. Fuels Resources for DFC • Natural Gas and LNG • Propane • Biogas (by Anaerobicnaerobic Digestion) - Municipal Waste...FUEL RESOURCES z NATURAL GAS z PROPANE z DFC H2 (50-60%) z ETHANOL zWASTE METHANE z BIOGAS z COAL GAS Diversity of Fuels plus High Efficiency

Fuel-cycle cost comparisons with oxide and silicide fuels

International Nuclear Information System (INIS)

Matos, J.E.; Freese, K.E.

1982-01-01

This paper addresses fuel cycle cost comparisons for a generic 10 MW reactor with HEU aluminide fuel and with LEU oxide and silicide fuels in several fuel element geometries. The intention of this study is to provide a consistent assessment of various design options from a cost point of view. Fuel cycle cost benefits could result if a number of reactors were to utilize fuel elements with the same number or different numbers of the same standard fuel plate. Data are presented to quantify these potential cost benefits. This analysis shows that there are a number of fuel element designs using LEU oxide or silicide fuels that have either the same or lower total fuel cycle costs than the HEU design. Use of these fuels with the uranium densities considered requires that they are successfully demonstrated and licensed

Fuel temperature characteristics of the 37-element and CANFLEX fuel bundle

International Nuclear Information System (INIS)

Bae, Jun Ho; Rho, Gyu Hong; Park, Joo Hwan

2009-10-01

This report describes the fuel temperature characteristics of CANFLEX fuel bundles and 37-element fuel bundles for a different burnup of fuel. The program was consisted for seeking the fuel temperature of fuel bundles of CANFLEX fuel bundles and 37-element fuel bundles by using the method in NUCIRC. Fuel temperature has an increasing pattern with the burnup of fuel for CANFLEX fuel bundles and 37-element fuel bundles. For all the case of burnup, the fuel temperature of CANFLEX fuel bundles has a lower value than that of 37-element fuel bundles. Especially, for the high power channel, the CANFLEX fuel bundles show a lower fuel temperature as much as about 75 degree, and the core averaged fuel temperature has a lower fuel temperature of about 50 degree than that of 37-element fuel bundles. The lower fuel temperature of CANFLEX fuel bundles is expected to enhance the safety by reducing the fuel temperature coefficient. Finally, for each burnup of CANFLEX fuel bundles and 37-element fuel bundles, the equation was present for predicting the fuel temperature of a bundle in terms of a coolant temperature and bundle power

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

Checklist for transition to new highway fuel(s).

Energy Technology Data Exchange (ETDEWEB)

Risch, C.; Santini, D.J. (Energy Systems)

2011-12-15

Transportation is vital to the U.S. economy and society. As such, U.S. Presidents have repeatedly stated that the nation needs to reduce dependence on petroleum, especially for the highway transportation sector. Throughout history, highway transportation fuel transitions have been completed successfully both in United States and abroad. Other attempts have failed, as described in Appendix A: Historical Highway Fuel Transitions. Planning for a transition is critical because the changes can affect our nation's ability to compete in the world market. A transition will take many years to complete. While it is tempting to make quick decisions about the new fuel(s) of choice, it is preferable and necessary to analyze all the pertinent criteria to ensure that correct decisions are made. Doing so will reduce the number of changes in highway fuel(s). Obviously, changes may become necessary because of occurrences such as significant technology breakthroughs or major world events. With any and all of the possible transitions to new fuel(s), the total replacement of gasoline and diesel fuels is not expected. These conventional fuels are envisioned to coexist with the new fuel(s) for decades, while the revised fuel and vehicle infrastructures are implemented. The transition process must analyze the needs of the primary 'players,' which consist of the customers, the government, the fuel industry, and the automotive industry. To maximize the probability of future successes, the prime considerations of these groups must be addressed. Section 2 presents a succinct outline of the Checklist. Section 3 provides a brief discussion about the groupings on the Checklist.

75 FR 37733 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

Science.gov (United States)

2010-06-30

... Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program AGENCY... direct final rule to amend the Renewable Fuel Standard program requirements on May 10, 2010. Because EPA... Fuel Standard program requirements, published on May 10, 2010. We stated in that direct final rule that...

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...

Alternate-Fueled Flight: Halophytes, Algae, Bio-, and Synthetic Fuels

Science.gov (United States)

Hendricks, R. C.

2012-01-01

Synthetic and biomass fueling are now considered to be near-term aviation alternate fueling. The major impediment is a secure sustainable supply of these fuels at reasonable cost. However, biomass fueling raises major concerns related to uses of common food crops and grasses (some also called "weeds") for processing into aviation fuels. These issues are addressed, and then halophytes and algae are shown to be better suited as sources of aerospace fuels and transportation fueling in general. Some of the history related to alternate fuels use is provided as a guideline for current and planned alternate fuels testing (ground and flight) with emphasis on biofuel blends. It is also noted that lessons learned from terrestrial fueling are applicable to space missions. These materials represent an update (to 2009) and additions to the Workshop on Alternate Fueling Sustainable Supply and Halophyte Summit at Twinsburg, Ohio, October 17 to 18, 2007.

Increased fuel burn-up and fuel cycle equilibrium

International Nuclear Information System (INIS)

Debes, M.

2001-01-01

Improvement of nuclear competitiveness will rely mainly on increased fuel performance, with higher burn-up, and reactors sustained life. Regarding spent fuel management, the EDF current policy relies on UO 2 fuel reprocessing (around 850 MTHM/year at La Hague) and MOX recycling to ensure plutonium flux adequacy (around 100 MTHM/year, with an electricity production equivalent to 30 TWh). This policy enables to reuse fuel material, while maintaining global kWh economy with existing facilities. It goes along with current perspective to increase fuel burn-up up to 57 GWday/t mean in 2010. The following presentation describes the consequences of higher fuel burn-up on fuel cycle and waste management and implementation of a long term and global equilibrium for decades in spent fuel management resulting from this strategy. (author)

IRT - Sofia conversion feasibility study experience 2002-2009

Energy Technology Data Exchange (ETDEWEB)

Belousov, S.I.; Apostolov, T.G. [Institute for Nuclear Research and Nuclear Energy of Bulgarian Academy of Science, Tsarigradsko 72, 1784 Sofia (Bulgaria)

2010-07-01

A joint conversion feasibility study concerning the IRT - Sofia research reactor between INRNE and the RERTR Program at ANL was initiated in 2002. The initial steps studies (up to 2006) were mainly focused on neutronics properties significant for reactor application and safety analyses. Thermal hydraulic, accident analyses as well as additional neutronics study required were performed after that (up to 2010). The obtained results show that the IRT-4M LEU fuel assemblies (19.75% {sup 235}U enrichment) are appropriate for IRT-Sofia conversion (IRT-Sofia was initially designed for the IRT-2M HEU fuel assemblies with 36% {sup 235}U enrichment). The results obtained in the frames of the joint study show that the IRT-Sofia operation even with usage of only one pump in the primary circuit meets all safety requirements at power level up to 1000 kW and that safety is maintained for accident transients. Presented results of analyses (neutronics, thermal hydraulic, and accident) and accumulated experience for the IRT-Sofia will be useful for other research reactors where conversion from IRT-2M (HEU) to IRT-4M (LEU) fuel is underway and/or foreseen. (authors)

Nuclear criticality safety parameter evaluation for uranium metallic alloy

Energy Technology Data Exchange (ETDEWEB)

Sanchez, Andrea; Abe, Alfredo, E-mail: andreasdpz@hotmail.com, E-mail: abye@uol.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Energia Nuclear

2013-07-01

Nuclear criticality safety during fuel fabrication process, transport and storage of fissile and fissionable materials requires criticality safety analysis. Normally the analysis involves computer calculations and safety parameters determination. There are many different Criticality Safety Handbooks where such safety parameters for several different fissile mixtures are presented. The handbooks have been published to provide data and safety principles for the design, safety evaluation and licensing of operations, transport and storage of fissile and fissionable materials. The data often comprise not only critical values, but also subcritical limits and safe parameters obtained for specific conditions using criticality safety calculation codes such as SCALE system. Although many data are available for different fissile and fissionable materials, compounds, mixtures, different enrichment level, there are a lack of information regarding a uranium metal alloy, specifically UMo and UNbZr. Nowadays uranium metal alloy as fuel have been investigated under RERTR program as possible candidate to became a new fuel for research reactor due to high density. This work aim to evaluate a set of criticality safety parameters for uranium metal alloy using SCALE system and MCNP Monte Carlo code. (author)

High density fuels using dispersion and monolithic fuel

Energy Technology Data Exchange (ETDEWEB)

Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia, E-mail: dsgomes@ipen.br, E-mail: teixeira@ipen.br, E-mail: alfredo@ctmsp.mar.mil.br, E-mail: rafael.orm@gmail.com, E-mail: claudia.giovedi@ctmsp.mar.mil.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Universidade de São Paulo (USP), SP (Brazil). Departamento de Engenharia Naval e Oceânica

2017-07-01

Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

High density fuels using dispersion and monolithic fuel

International Nuclear Information System (INIS)

Gomes, Daniel S.; Silva, Antonio T.; Abe, Alfredo Y.; Muniz, Rafael O.R.; Giovedi, Claudia; Universidade de São Paulo

2017-01-01

Fuel plates used in high-performance research reactors need to be converted to low-enrichment uranium fuel; the fuel option based on a monolithic formulation requires alloys to contain 6 - 10 wt% Mo. In this case, the fuel plates are composed of the metallic alloy U-10Mo surrounded by a thin zirconium layer encapsulated in aluminum cladding. This study reviewed the physical properties of monolithic forms. The constraints produced during the manufacturing process were analyzed and compared to those of dispersed fuel. The bonding process used for dispersion fuels differs from the techniques applied to foil bonding used for pure alloys. The quality of monolithic plates depends on the fabrication method, which usually involves hot isostatic pressing and the thermal annealing effect of residual stress, which degrades the uranium cubic phase. The preservation of the metastable phase has considerable influence on fuel performance. The physical properties of the foil fuel under irradiation are superior to those of aluminum-dispersed fuels. The fuel meat, using zirconium as the diffusion barrier, prevents the interaction layer from becoming excessively thick. The problem with dispersed fuel is breakaway swelling with a medium fission rate. It has been observed that the fuel dispersed in aluminum was minimized in monolithic forms. The pure alloys exhibited a suitable response from a rate at least twice as much as the fission rate of dispersions. The foils can support fissile material concentration combined with a reduced swelling rate. (author)

Fuel rod behaviour at high burnup WWER fuel cycles

International Nuclear Information System (INIS)

Medvedev, A.; Bogatyr, S.; Kouznetsov, V.; Khvostov, G.; Lagovsky; Korystin, L.; Poudov, V.

2003-01-01

The modernisation of WWER fuel cycles is carried out on the base of complete modelling and experimental justification of fuel rods up to 70 MWd/kgU. The modelling justification of the reliability of fuel rod and fuel rod with gadolinium is carried out with the use of certified START-3 code. START-3 code has a continuous experimental support. The thermophysical and strength reliability of WWER-440 fuel is justified for fuel rod and pellet burnups 65 MWd/kgU and 74 MWd/U, accordingly. Results of analysis are demonstrated by the example of uranium-gadolinium fuel assemblies of second generation under 5-year cycle with a portion of 6-year assemblies and by the example of successfully completed pilot operation of 5-year cycle fuel assemblies during 6 years at unit 3 of Kolskaja NPP. The thermophysical and strength reliability of WWER-1000 fuel is justified for a fuel rod burnup 66 MWd/kgU by the example of fuel operation under 4-year cycles and 6-year test operation of fuel assemblies at unit 1 of Kalininskaya NPP. By the example of 5-year cycle at Dukovany NPP Unit 2 it was demonstrated that WWER fuel rod of a burnup 58 MWd/kgU ensure reliable operation under load following conditions. The analysis has confirmed sufficient reserves of Russian fuel to implement program of JSC 'TVEL' in order to improve technical and economical parameters of WWER fuel cycles

78 FR 62462 - Regulation of Fuels and Fuel Additives: Modifications to Renewable Fuel Standard Program

Science.gov (United States)

2013-10-22

... renewable fuel is defined as fuel produced from renewable biomass that is used to replace or reduce the quantity of fossil fuel present in home heating oil or jet fuel.\\3\\ In essence, additional renewable fuel... of ``home heating oil.'' EPA determined that this term was ambiguous, and defined it by incorporating...

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)

Low contaminant formic acid fuel for direct liquid fuel cell

Science.gov (United States)

Masel, Richard I [Champaign, IL; Zhu, Yimin [Urbana, IL; Kahn, Zakia [Palatine, IL; Man, Malcolm [Vancouver, CA

2009-11-17

A low contaminant formic acid fuel is especially suited toward use in a direct organic liquid fuel cell. A fuel of the invention provides high power output that is maintained for a substantial time and the fuel is substantially non-flammable. Specific contaminants and contaminant levels have been identified as being deleterious to the performance of a formic acid fuel in a fuel cell, and embodiments of the invention provide low contaminant fuels that have improved performance compared to known commercial bulk grade and commercial purified grade formic acid fuels. Preferred embodiment fuels (and fuel cells containing such fuels) including low levels of a combination of key contaminants, including acetic acid, methyl formate, and methanol.

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.)

Fuel quality issues in stationary fuel cell systems.

Energy Technology Data Exchange (ETDEWEB)

Papadias, D.; Ahmed, S.; Kumar, R. (Chemical Sciences and Engineering Division)

2012-02-07

Fuel cell systems are being deployed in stationary applications for the generation of electricity, heat, and hydrogen. These systems use a variety of fuel cell types, ranging from the low temperature polymer electrolyte fuel cell (PEFC) to the high temperature solid oxide fuel cell (SOFC). Depending on the application and location, these systems are being designed to operate on reformate or syngas produced from various fuels that include natural gas, biogas, coal gas, etc. All of these fuels contain species that can potentially damage the fuel cell anode or other unit operations and processes that precede the fuel cell stack. These detrimental effects include loss in performance or durability, and attenuating these effects requires additional components to reduce the impurity concentrations to tolerable levels, if not eliminate the impurity entirely. These impurity management components increase the complexity of the fuel cell system, and they add to the system's capital and operating costs (such as regeneration, replacement and disposal of spent material and maintenance). This project reviewed the public domain information available on the impurities encountered in stationary fuel cell systems, and the effects of the impurities on the fuel cells. A database has been set up that classifies the impurities, especially in renewable fuels, such as landfill gas and anaerobic digester gas. It documents the known deleterious effects on fuel cells, and the maximum allowable concentrations of select impurities suggested by manufacturers and researchers. The literature review helped to identify the impurity removal strategies that are available, and their effectiveness, capacity, and cost. A generic model of a stationary fuel-cell based power plant operating on digester and landfill gas has been developed; it includes a gas processing unit, followed by a fuel cell system. The model includes the key impurity removal steps to enable predictions of impurity breakthrough

Fuel cells for electricity generation from carbonaceous fuels

Energy Technology Data Exchange (ETDEWEB)

Ledjeff-Hey, K; Formanski, V; Roes, J [Gerhard-Mercator- Universitaet - Gesamthochschule Duisburg, Fachbereich Maschinenbau/Fachgebiet Energietechnik, Duisburg (Germany); Heinzel, A [Fraunhofer Inst. for Solar Energy Systems (ISE), Freiburg (Germany)

1998-09-01

Fuel cells, which are electrochemical systems converting chemical energy directly into electrical energy with water and heat as by-products, are of interest as a means of generating electricity which is environmentally friendly, clean and highly efficient. They are classified according to the electrolyte used. The main types of cell in order of operating temperature are described. These are: alkaline fuel cells, the polymer electrolyte membrane fuel cell (PEMFC); the phosphoric acid fuel cell (PAFC); the molten carbonate fuel cell (MCFC); the solid oxide fuel cell (SOFC). Applications depend on the type of cell and may range from power generation on a large scale to mobile application in cars or portable systems. One of the most promising options is the PEM-fuel cell stack where there has been significant improvement in power density in recent years. The production from carbonaceous fuels and purification of the cell fuel, hydrogen, is considered. Of the purification methods available, hydrogen separation by means of palladium alloy membranes seems particular effective in reducing CO concentrations to the low levels required for PEM cells. (UK)

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

Fuel cracking in relation to fuel oxidation in support of an out-reactor instrumented defected fuel experiment