Nuclear criticality assessment of LEU and HEU fuel element storage

International Nuclear Information System (INIS)

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

1984-01-01

Criticality aspects of storing LEU (20%) and HEU (93%) fuel elements have been evaluated as a function of 235 U loading, element geometry, and fuel type. Silicide, oxide, and aluminide fuel types have been evaluated ranging in 235 U loading from 180 to 620 g per element and from 16 to 23 plates per element. Storage geometry considerations have been evaluated for fuel element separations ranging from closely packed formations to spacings of several centimeters between elements. Data are presented in a form in which interpolations may be made to estimate the eigenvalue of any fuel element storage configuration that is within the range of the data. (author)

Full size U-10Mo monolithic fuel foil and fuel plate fabrication-technology development

International Nuclear Information System (INIS)

Moore, G.A.; Jue, J-F.; Rabin, B.H.; Nilles, M.J.

2010-01-01

Full-size U-10Mo foils are being developed for use in high density LEU monolithic fuel plates. The application of a zirconium barrier layer to the foil is performed using a hot co-rolling process. Aluminium clad fuel plates are fabricated using Hot Isostatic Pressing (HIP) or a Friction Bonding (FB) process. An overview is provided of ongoing technology development activities, including: the co-rolling process, foil shearing/slitting and polishing, cladding bonding processes, plate forming, plate-assembly swaging, and fuel plate characterization. Characterization techniques being employed include, Ultrasonic Testing (UT), radiography, and microscopy. (author)

Reprocessing of LEU silicide fuel at Dounreay

International Nuclear Information System (INIS)

Cartwright, P.

1996-01-01

UKAEA have recently reprocessed two LEU silicide fuel elements in their MTR fuel reprocessing plant at Dounreay. The reprocessing was undertaken to demonstrate UKAEA's commitment to the world-wide research reactor communities future needs. Reprocessing of LEU silicide fuel is seen as a waste treatment process, resulting in the production of a liquid feed suitable for conditioning in a stable form of disposal. The uranium product from the reprocessing can be used as a blending feed with the HEU to produce LEU for use in the MTR cycle. (author)

The whole-core LEU silicide fuel demonstration in the JMTR

Energy Technology Data Exchange (ETDEWEB)

Aso, Tomokazu; Akashi, Kazutomo; Nagao, Yoshiharu [Japan Atomic Energy Research Institute, Ibaraki-ken (Japan)] [and others

1997-08-01

The JMTR was fully converted to LEU silicide (U{sub 3}Si{sub 2}) fuel with cadmium wires as burnable absorber in January, 1994. The reduced enrichment program for the JMTR was initiated in 1979, and the conversion to MEU (enrichment ; 45%) aluminide fuel was carried out in 1986 as the first step of the program. The final goal of the program was terminated by the present LEU conversion. This paper describes the results of core physics measurement through the conversion phase from MEU fuel core to LEU fuel core. Measured excess reactivities of the LEU fuel cores are mostly in good agreement with predicted values. Reactivity effect and burnup of cadmium wires, therefore, were proved to be well predicted. Control rod worth in the LEU fuel core is mostly less than that in the MEU fuel core. Shutdown margin was verified to be within the safety limit. There is no significant difference in temperature coefficient of reactivity between the MEU and LEU fuel cores. These results verified that the JMTR was successfully and safely converted to LEU fuel. Extension of the operating cycle period was achieved and reduction of spend fuel elements is expected by using the fuel with high uranium density.

Performance of PARR-1 with LEU Fuel

International Nuclear Information System (INIS)

Pervez, S.; Latif, M.; Bokhari, I.H.; Bakhtyar, S.

2005-01-01

Pakistan Research Reactor (PARR-1) went critical in 1965 with HEU fuel. The reactor core was converted to LEU fuel with power upgradation from 5 MW to 10 MW in 1992. The reactor has been operated with LEU fuel for about 10,000 hours and has produced about 66,000 MWh energy up to now. Average burn up of the irradiated fuel is about 42 %. The fuel performance during the last 12 years has been excellent. Post irradiation visual inspection of the fuel has revealed no abnormality. During operation there have been no signs of releases in the pool water establishing the full integrity of this fuel. The reactor has been mainly utilized for radioisotope production, beam tube experiments including neutron diffraction studies, neutron radiography etc. Studies have been completed to operate the reactor with a mixed core (HEU + LEU) to utilize the less burned HEU fuel elements. A major project of production of fission Moly using PARR-1 is in the final stages. (author)

The ORR Whole-Core LEU Fuel Demonstration

International Nuclear Information System (INIS)

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

1990-01-01

The ORR Whole-Core LEU Fuel Demonstration, conducted as part of the US Reduced Enrichment Research and Test Reactor Program, has been successfully completed. Using commercially-fabricated U 3 Si 2 -Al 20%-enriched fuel elements (4.8 g U/cc) and fuel followers (3.5 g U/cc), the 30-MW Oak Ridge Research Reactor was safely converted from an all-HEU core, through a series of HEU/LEU mixed transition cores, to an all-LEU core. There were no fuel element failures and average discharge burnups were measured to be as high as 50% for the standard elements and 75% for the fuel followers. Experimental results for burnup-dependent critical configurations, cycle-averaged fuel element powers, and fuel-element-averaged 235 U burnups validated predictions based on three-dimensional depletion calculations. Calculated values for plutonium production and isotopic mass ratios as functions of 235 U burnup support the corresponding measured quantities. In general, calculations for reaction rate distributions, control rod worths, prompt neutron decay constants, and isothermal temperature coefficients were found to agree with corresponding measured values. Experimentally determined critical configurations for fresh HEU and LEU cores radially reflected with water and with beryllium are well-predicted by both Monte Carlo and diffusion calculations. 17 refs

Whole-core LEU fuel demonstration in the ORR

International Nuclear Information System (INIS)

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

1985-01-01

A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U 3 Si 2 at 4.8 Mg U/m 3 and shim rod fuel followers will contain U 3 Si 2 at 3.5 Mg U/m 3 . Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U 3 Si 2 fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worths, cycle length, fuel discharge burnup, gamma heating rates, β/sub eff/l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed

Core thermohydraulic design with LEU fuels for upgraded research reactor, JRR-3

Energy Technology Data Exchange (ETDEWEB)

Sudo, Y; Ando, H; Ikawa, H; Ohnishi, N [Department of Research Reactor Operation, Japan Atomic Energy Research Institute (JAERI), 319-11 Tokai-Mura, Ibaraki-Ken (Japan)

1985-07-01

This paper presents the outline of core thermohydraulic design and analysis of the research reactor, JRR-3, which is to be upgraded to a 20 MWt pool-type, light water-cooled reactor with 20% LEU plate-type fuels. The major feature of core thermohydraulics of the upgraded JRR-3 is that core flow is a downflow at the condition of normal operation, with which fuel plates are exposed to a severer condition than with an upflow in case of operational transients and accidents. The core thermo-hydraulic design was, therefore, done for the condition of normal operation so that fuel plates may have enough safety margin both against the onset of nucleate boiling not to allow the nucleate boiling anywhere in the core and against the initiation of DNB, and the safety margin for these were evaluated. The core velocity thus designed is at the optimum condition where fuel plates have the maximum margin against the onset of nucleate boiling. The core thermohydraulic characteristics were also clarified for the natural circulation cooling mode. (author)

Full core operation in JRR-3 with LEU fuels

International Nuclear Information System (INIS)

Murayama, Y.; Issiki, M.

1995-01-01

The new JRR-3 a 20MWT swimming pool type research reactor, is made up of plate type LEU fuel elements with U-Al x fuel at 2.2 gU/cm 3 . Reconstruction work for the new JR-3 was a good success, and common operation started in November 1990, and 7 cycles (26 days operation/cycle) have passed. We have no experience in using such a high uranium density fuel element with aluminide fuel. So we plan to examine the condition of the irradiated fuel elements with three methods, that is, measurement of the value of FFD in operation, observation of external view of the fuels in refueling work and postirradiation examination after maximum burn-up will be established. In the results of the first two methods, the fuel elements of JRR-3 is burned up normally and have no evidence of failure. (author)

Radiological consequence analysis with HEU and LEU fuels

Energy Technology Data Exchange (ETDEWEB)

Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

1984-01-01

A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables.

Radiological consequence analysis with HEU and LEU fuels

International Nuclear Information System (INIS)

Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

1984-01-01

A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables

The conversion of NRU from HEU to LEU fuel

International Nuclear Information System (INIS)

Sears, D.F.; Atfield, M.D.; Kennedy, I.C.

1990-01-01

The program at Chalk River Nuclear Laboratories (CRNL) to develop and test low-enriched uranium fuel (LEU, 3 Si, USiAl, USi Al and U 3 Si 2 (U-3.96 wt% Si; U-3.5 wt% Si-1.5 wt% AL; U-3.2 wt%; Si-3 wt% Al; U-7.3 wt% Si, respectively). Fuel elements were fabricated with uranium loadings suitable for NRU, 3.15 gU/cm 3 , and for NRX, 4.5 gU/cm 3 , and were irradiated under normal fuel-operating conditions. Eight experimental irradiations involving 100 mini-elements and 84 full-length elements (7X12-element rods) were completed to qualify the LEU fuel and the fabrication technology. Post irradiation examinations confirmed that the performance of the LEU fuel, and that of a medium enrichment uranium (MEU, 45% U-235) alloy fuel tested as a back-up, was comparable to the HEU fuel. The uranium silicide dispersion fuel swelling was approximately linear up to burnups exceeding NRU's design terminal burnup (80 at%). NRU was partially converted to LEU fuel when the first 31 prototype fuel rods manufactured with industrial scale production equipment were installed in the reactor. The rods were loaded in NRU at a fuelling rate of about two rods per week over the period 1988 September to December. This partial LEU core (one third of a full NRU core) has allowed the reactor engineers and physicists to evaluate the bulk effects of the LEU conversion on NRU operations. As expected, the irradiation is proceeding without incident

LEU fuel cycle analyses for the Belgian BR2 Research Reactor

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.

1988-01-01

Equilibrium fuel cycle characteristics were calculated for reference HEU and two proposed LEU fuel cycles using an 11-group diffusion-theory neutron flux solution in hexagonal-Z geometry. The diffusion theory model was benchmarked with a detailed Monte Carlo core model. The two proposed LEU fuel designs increased the 235 U loading 20% and the fuel meat volume 51%. The first LEU design used 10 B as a burnable absorber. Either proposed LEU fuel element would provide equilibrium fuel cycle characteristics similar to those of the HEU fuel cycle. Irradiation rates of Co control followers and Ir disks in the center of the core were reduced 6 ± 1% in the LEU equilibrium core compared to reference HEU core. 11 refs., 4 figs., 5 tabs

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

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.

A radiological consequence analysis with HEU and LEU fuels

International Nuclear Information System (INIS)

Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

1985-01-01

A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and nonsite specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. (author)

The whole-core LEU fuel demonstration in the ORR

International Nuclear Information System (INIS)

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

1985-01-01

A whole-core demonstration of LEU fuel in the ORR is expected to begin during November 1985. Fuel elements will contain U 3 Si 2 at 4.8 Mg U/m 3 and shim rod fuel followers will contain U 3 Si 2 at 3.5 Mg U/m 3 . Fuel fabrication is underway at B and W, CERCA, and NUKEM, with shipments scheduled to commence in October. The primary objectives of the demonstration are to provide data for validation of LEU and mixed-core fuel cycle calculations and to provide a large-scale demonstration of the acceptable performance of production-line U 3 Si 2 fuel elements. It is planned to approach the full LEU core through a series of mixed cores. Measurements to be made include flux distribution, reactivity swing, control rod worth, cycle length, fuel discharge burn-up, gamma heating rate, β eff /l, and isothermal temperature coefficient. Measurements will also be made on fresh LEU and fresh HEU critical configurations. Preliminary safety approval has been received and the final safety assessment is being reviewed. Key issues being addressed in the safety assessment are fuel performance, radiological consequences, margin to burnout and transient behavior. The LEU core is comparable in all safety aspects to the HEU core and the transition core is only marginally worse owing to higher power seeking factors. (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

LEU-plate irradiation at FRJ-2 (DIDO) under the German AF-programme

Energy Technology Data Exchange (ETDEWEB)

Groos, E; Krug, W; Seferiadis, J; Thamm, G

1985-07-01

10 LEU fuel plates (8 with uranium silicides max. U-density 6.1 g/cm{sup 3}) have been irradiated at FRJ-2 (DIDO) of KFA-Juelich till end of October 1984 during 321 full power days up to max. burnup of 2.41x10{sup 27} fissions/m{sup 3} without major interruptions and troubles. PIE began recently in KFA hot cells. Visual inspections revealed no damage or greater deformation for the majority of the plates, but red/brown coloured layers (partially peeled off) on the cladding over the fuel. Aluminium (oxide) is the chief constituent of the layer with smaller portions of Ni and Fe the latter causing the red/brown colour. The major part of the layer ({approx}50 {mu}m) most probably has been formed during 20 h immediately after experiment start-up under abnormal conditions of the coolant water. Gamma scanning has been completed. Dimensional measurements are under way confirming first observations of severe swelling (pillowing) of 1 plate. Density and blister testing as well as metallography and burnup analysis remain to be accomplished end of 1985/beginning of 1986. (author)

Neutronic study on conversion of SAFARI-1 to LEU silicide fuel

International Nuclear Information System (INIS)

Ball, G.; Pond, R.; Hanan, N.; Matos, J.

1995-01-01

This paper marks the initial study into the technical and economic feasibility of converting the SAFARI-1 reactor in South Africa to LEU silicide fuel. Several MTR assembly geometries and LEU uranium densities have been studied and compared with MEU and HEU fuels. Two factors of primary importance for conversion of SAFARI-1 to LEU fuel are the economy of the fuel cycle and the performance of the incore and excore irradiation positions

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)

Fabrication, fabrication control and in-core follow up of 4 LEU leader fuel elements based on U3Si2 in RECH-1

International Nuclear Information System (INIS)

Chavez, J.C.; Barrera, M.; Olivares, L.; Lisboa, J.

1999-01-01

The RECH-1 MTR reactor has been converted from HEU to MEU (45% enrichment) and the decision to a LEU (20% enrichment) conversion was taken some years ago. This LEU conversion decision involved a local fuel development and fabrication based on U 3 Si 2 -Al dispersion fuel, and a fabrication qualification stage that resulted in four fuel elements fully complying with established fabrication standards for this type of fuel. This report-presents relevant points of these four leaders fuel elements fabrication, in particular a fuel plate core homogeneity control development. A summary of the intended in core follow-up studies for the leaders fuel elements is also presented here. (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)

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

Progress on LEU very high density fuel and target development in Argentina

International Nuclear Information System (INIS)

Balart, S.; Cabot, P.; Calzetta, O.; Duran, A.; Garces, J.; Hermida, J.D.; Manzini, A.; Pasqualini, E.; Taboada, H.

2006-01-01

Since last RRFM meeting, CNEA has continued on new LEU fuel and target development activities. Main goals are the plan to convert our RA-6 reactor from HEU to a new LEU core, 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, to optimize techniques to recover U from silicide scrap samples as cold test for radiowaste separation for final conditioning of silicide spent fuels. and to improve the diffusion of LEU target and radiochemical technology for radioisotope production. Future plans include: - Completion of the RA-6 reactor conversion to LEU; - Improvement on fuel development and production facilities to implement new technologies, including NDT techniques to assess bonding quality; - Irradiation of miniplates and full scale fuel assembly at RA-3 and plans to perform irradiation on higher power and temperature regime reactors; - Optimization of LEU target and radiochemical techniques for radioisotope production. (author)

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

Progress in safety evaluation for the JMTR core conversion to LEU fuel

International Nuclear Information System (INIS)

Sakurai, F.; Komori, Y.; Saito, J.; Komukai, B.; Ando, H.; Nakata, H.; Sakakura, A.; Niiho, S.; Saito, M.; Futamura, Y.

1991-01-01

The JMTR (50 MWt) has been in steady operation with MEU fuel since July 1986. The effort is still continued to convert the core from MEU to LEU fuel. The LEU silicide fuel element at 4.8 gU/cm 3 with Cd wires as burnable absorbers has been selected in order to achieve upgraded fuel cycle performance of extended cycle length and reduced control rod movement operation. The neutronic calculation methods (diffusion theory model) developed for the LEU core with Cd wires was benchmarked with a detailed Monte Carlo model and verified experimentally using the critical facility, JMTRC. Hydraulic tests of the LEU silicide fuel element with Cd wires were completed with satisfactory results, and measurements of release/born (R/B) ratios of FPs of silicide fuel at high temperature are in progress. (orig.)

Re-qualification of MTR-type fuel plates fabrication process

International Nuclear Information System (INIS)

Elseaidy, I.M.; Ghoneim, M.M.

2010-01-01

The fabricability issues with increased uranium loading due to use low enrichment of uranium (LEU), i.e. less than 20 % of U 235 , increase the problems which occur during compact manufacturing, roll bonding of the fuel plates, potential difficulty in forming during rolling process, mechanical integrity of the core during fabrication, potential difficulty in meat homogeneity, and the ability to fabricate plates with thicker core as a means of increasing total uranium loading. To produce MTR- type fuel plates with high uranium loading (HUL) and keep the required quality of these plates, many of qualification process must be done in the commissioning step of fuel fabrication plant. After that any changing of the fabrication parameters, for example changing of any of the raw materials, devises, operators, and etc., a re- qualification process should be done in order to keep the quality of produced plates. Objective of the present work is the general description of the activities to be accomplished for re-qualification of manufacturing MTR- type nuclear fuel plates. For each process to be re-qualified, a detailed of re-qualification process were established. (author)

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)

The University of Missouri Research Reactor HEU to LEU conversion project status

Energy Technology Data Exchange (ETDEWEB)

McKibben, James C; Kutikkad, Kiratadas; Foyto, Leslie P; Peters, Nickie J; Solbrekken, Gary L; Kennedy, John [University of Missouri Research Reactor, Missouri (United States); Stillman, John A; Feldman, Earl E; Tzanos, Constantine P; Stevens, John G [Argonne National Laboratory, Argonne, Illinois (United States)

2012-03-15

The University of Missouri Research Reactor (MURR) is one of five U.S. high performance research and test reactors that are actively collaborating with the U.S. Department of Energy (DOE) to find a suitable low-enriched uranium (LEU) fuel replacement for the currently required highly-enriched uranium (HEU) fuel. A conversion feasibility study based on U-10Mo monolithic LEU fuel was completed in 2009. It was concluded that the proposed LEU fuel assembly design, in conjunction with an increase in power level from 10 to 12 MWth, will (1) maintain safety margins during operation, (2) allow operating fuel cycle lengths to be maintained for efficient and effective use of the facility, and (3) preserve an acceptable level and spectrum of key neutron fluxes to meet the scientific mission of the facility. The MURR and Argonne National Laboratory (ANL) team is continuing to work toward realization of the conversion. The 'Preliminary Safety Analysis Report Methodologies and Scenarios for LEU Conversion of MURR' was completed in June 2011. This report documents design parameter values critical to the Fuel Development (FD), Fuel Fabrication Capability (FFC) and Hydromechanical Fuel Test Facility (HMFTF) projects. The report also provides a preliminary evaluation of safety analysis techniques and data that will be needed to complete the fuel conversion Safety Analysis Report (SAR), especially those related to the U-10Mo monolithic LEU fuel. Specific studies are underway to validate the proposed path to an LEU fuel conversion. Coupled fluid-structure simulations and experiments are being conducted to understand the hydrodynamic plate deformation risk for 0.965 mm (38 mil) thick fuel plates. Methodologies that were recently developed to answer the U.S. Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) regarding the MURR 2006 relicensing submittal will be used in the LEU conversion effort. Transition LEU fuel elements that will have a minimal impact on

Comparison of the FRM-II HEU design with an alternative LEU design

International Nuclear Information System (INIS)

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

2004-01-01

The FRM-II reactor design of the Technical University of Munich has a compact core that utilizes fuel plates containing highly-enriched uranium (HEU, 93%). This paper presents an alternative core design utilizing low-enriched uranium (LEU, 3 that provides nearly the same neutron flux for experiments as the HEU design, but has a less favourable fuel cycle economy. If an LEU fuel with a uranium density of 6.0 - 6.5 g/cm 3 . were developed, the alternative design would provide the same neutron flux and use the same number of cores per year as the HEU design. The results of this study show that there are attractive possibilities for using LEU fuel instead of HEU fuel in the FRM-II. Further optimization of the LEU design and near-term availability of LEU fuel with a uranium density greater than 4.8 g/cm 3 would enhance the performance of the LEU core. The REKIR Program is ready to exchange information with the Technical University of Munich to resolve any differences that may exist and to identify design modifications that would optimize reactor performance utilizing LEU fuel. (author)

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)

HEU and Leu FueL Shielding Comparative Study Applied for Spent Fuel Transport

International Nuclear Information System (INIS)

Margeanu, C.A.; Margeanu, S.; Barbos, D.

2009-01-01

INR Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies. The dual core concept involves the operation of a 14 MW TRIGA steady-state, high flux research and material testing reactor at one end of a large pool, and the independent operation of an annular-core pulsing reactor (TRIGA-ACPR) at the other end of the pool. The steady-state reactor is mostly used for long term testing of power reactor fuel components (pellets, pins, subassemblies and fuel assemblies) followed by post-irradiation examination. Following the general trend to replace the He fuel type (High Enriched Uranium) by Leu fuel type (Low Enriched Uranium), in the light of international agreements between IAEA and the states using He fuel in their nuclear reactors, Inr Past's have been accomplished the TRIGA research reactor core full conversion on May 2006. The He fuel repatriation in US in the frame of Foreign Research Reactor Spent Nuclear Fuel Return Programme effectively started in 1999, the final stage being achieved in summer of 2008. Taking into account for the possible impact on the human and environment, in all activities associated to nuclear fuel cycle, the spent fuel or radioactive waste characteristics must be well known. Shielding calculations basic tasks consist in radiation doses calculation, in order to prevent any risks both for personnel protection and impact on the environment during the spent fuel manipulation, transport and storage. The paper is a comparative study of Leu and He fuel utilization effects for the shielding analysis during spent fuel transport. A comparison against the measured data for He spent fuel, available from the last stage of the spent fuel repatriation, is presented. All the geometrical and material data related on the spent fuel shipping cask were considered according to the Nac-Lt Cask approved model. The shielding analysis estimates radiation doses to shipping cask wall surface

Neutronics Study on LEU Nuclear Thermal Rocket Fuel Options

Energy Technology Data Exchange (ETDEWEB)

Venneri, Paolo; Kim, Yong Hee [KAIST, Daejeon (Korea, Republic of); Howe, Steven [CSNR, Idaho (United States)

2014-10-15

This has resulted in a non-trivial simplification of the tasks needed to develop such an engine and the quick initial development of the concept. There are, however, a series of key core-design choices that are currently under scrutiny in the field that have to be resolved in order for the LEU-NTR to be fully developed. The most important of these is the choice of fuel: carbide composite or tungsten cermet. This study presents a first comparison of the two fuel types specifically in the neutronic application to the LEU-NTR, keeping in mind the unique neutronic environment and the system requirements of the system. The scope of the study itself is limited to a neutronics study of the two fuels and only a cursory overview of the material properties of the fuels themselves... The results of this study have led to two major conclusions. First of all is that the carbide composite fuel is, from a neutronics standpoint, a much better fuel. It has a low absorption cross-section, is inherently a strong moderator, is able to achieve a higher reactivity using smaller amounts of fissile material, and can potentially enable a smaller reactor. Second is that despite its neutronic difficulties (high absorption, inferior moderating abilities, and lower k-infinity values) the tungsten cermet fuel is still able to perform satisfactorily in an LEU-NTR, largely due to its ability to have an extremely high fuel loading.

Fuel cycle flexibility in Advanced Heavy Water Reactor (AHWR) with the use of Th-LEU fuel

International Nuclear Information System (INIS)

Thakur, A.; Singh, B.; Pushpam, N.P.; Bharti, V.; Kannan, U.; Krishnani, P.D.; Sinha, R.K.

2011-01-01

The Advanced Heavy Water Reactor (AHWR) is being designed for large scale commercial utilization of thorium (Th) and integrated technological demonstration of the thorium cycle in India. The AHWR is a 920 MW(th), vertical pressure tube type cooled by boiling light water and moderated by heavy water. Heat removal through natural circulation and on-line fuelling are some of the salient features of AHWR design. The physics design of AHWR offers considerable flexibility to accommodate different kinds of fuel cycles. Our recent efforts have been directed towards a case study for the use of Th-LEU fuel cycle in a once-through mode. The discharged Uranium from Th-LEU cycle has proliferation resistant characteristics. This paper gives the initial core, fuel cycle characteristics and online refueling strategy of Th-LEU fuel in AHWR. (author)

Innovative nuclear thermal rocket concept utilizing LEU fuel for space application

International Nuclear Information System (INIS)

Nam, Seung Hyun; Venneri, Paolo; Choi, Jae Young; Jeong, Yong Hoon; Chang, Soon Heung

2015-01-01

Space is one of the best places for humanity to turn to keep learning and exploiting. A Nuclear Thermal Rocket (NTR) is a viable and more efficient option for human space exploration than the existing Chemical Rockets (CRs) which are highly inefficient for long-term manned missions such as to Mars and its satellites. NERVA derived NTR engines have been studied for the human missions as a mainstream in the United States of America (USA). Actually, the NERVA technology has already been developed and successfully tested since 1950s. The state-of-the-art technology is based on a Hydrogen gas (H_2) cooled high temperature reactor with solid core utilizing High-Enriched Uranium (HEU) fuel to reduce heavy metal mass and to use fast or epithermal neutron spectrums enabling simple core designs. However, even though the NTR designs utilizing HEU is the best option in terms of rocket performance, they inevitably provoke nuclear proliferation obstacles on all Research and Development (R and D) activities by civilians and non-nuclear weapon states, and its eventual commercialization. To surmount the security issue to use HEU fuel for a NTR, a concept of the innovative NTR engine, Korea Advanced NUclear Thermal Engine Rocket utilizing Low-Enriched Uranium fuel (KANUTER-LEU) is presented in this paper. The design goal of KANUTER-LEU is to make use of a LEU fuel for its compact reactor, but does not sacrifice the rocket performance relative to the traditional NTRs utilizing HEU. KANUTER-LEU mainly consists of a fission reactor utilizing H_2 propellant, a propulsion system and an optional Electricity Generation System as a bimodal engine. To implement LEU fuel for the reactor, the innovative engine adopts W-UO_2 CERMET fuel to drastically increase uranium density and thermal neutron spectrum to improve neutron economy in the core. The moderator and structural material selections also consider neutronic and thermo-physical characteristics to reduce non-fission neutron loss and

A neutronics study of LEU fuel options for the HFR-Petten

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.

1985-01-01

The standard HEU fuel cycle characteristics are compared with those of several different LEU fuel cycles in the new vessel configuration. The primary design goals were to provide similar reactivity performance and neutron flux profiles with a minimal increase in 235 U loading. The fuel cycle advantages of Cd burnable absorbers over 10 B are presented. The LEU fuel cycle requirements were calculated also for an extended 32-day cycle and for a reload batch size reduction from six to five standard elements for the standard 26-day cycle. The effects of typical in-core experiments upon neutron flux profiles and fuel loading requirements are also presented. (author)

Status of LEU fuel development and conversion of NRU

International Nuclear Information System (INIS)

Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

1991-01-01

This paper reviews the status of the LEU conversion program and the progress made in the fuel development program over the last year. The results from post-irradiation examinations of prototype NRU fuel rods containing Al-U 3 Si dispersion fuel, and of mini-elements containing Al-U 3 Si 2 dispersion fuel, are presented. (orig.)

Mixed core conversion study with HEU and LEU fuels

International Nuclear Information System (INIS)

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

1984-01-01

The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically

Fuel cycle cost comparisons with oxide and silicide fuels

Energy Technology Data Exchange (ETDEWEB)

Matos, J E; Freese, K E [RERTR Program, Argonne National Laboratory (United States)

1983-09-01

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. The status of the development and demonstration of the oxide and silicide fuels are presented in several papers in these proceedings. Routine utilization of these fuels with the uranium densities considered here requires that they are successfully demonstrated and licensed. Thermal-hydraulic safety margins, shutdown margins, mixed cores, and transient analyses are not addressed here, but analyses of these safety issues are in progress for a limited number of the most promising design options. 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 is 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. All safety criteria for the reactor with these fuel element designs need to be satisfied as well. With LEU oxide fuel, 31 g U/cm{sup 3} 1 and 0.76 mm--thick fuel meat, elements with 18-22 plates 320-391 g {sup 235}U) result in the same or lower total costs than with the HEU element 23 plates, 280 g {sup 235}U). Higher LEU loadings (more plates per element) are needed for larger excess reactivity requirements. However, there is little cost advantage to using more than 20 of these plates per element. Increasing the fuel meat thickness from 0.76 mm to 1.0 mm with 3.1 g U/cm{sup 3} in the design with 20 plates per element could result in significant cost reductions if the

Comparison of the FRM-II HEU design with an alternative LEU design. Attachment

International Nuclear Information System (INIS)

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

2004-01-01

After presentation of the foregoing paper by Dr. Nelson Hanan of Argonne National Laboratory (ANL) proposing an alternative LEU core with one fuel ring and a power level of 33 MW, a presentation was made by Dr. Klaus Boning of the Technical University of Munich comparing the FRM-II HEU design with an LEU design by Tlm that had two fuel rings and a power level of 40 MW. Dr. Boning raised the following issues concerning the use of LEU fuel in FRM-H reactor designs: (1) qualification of HEU and LEU silicide fuels, (2) gamma heating in the heavy water reflector, (3) the radiological consequences of hypothetical accidents, and (4) cost and schedule. These issues are addressed in this Attachment. In his presentation, Dr. Hanan mentioned that ANL was also investigating other LEU designs. This work led to a second alternative LEU design that has the same neutron flux performance (8 x 10 14 n/cm 2 /s peak neutron flux in the reflector) and the same fuel lifetime (50 full power days) as the HEU design, but uses LEU silicide fuel with a uranium density of only 4.5 g/cm 3 . This design was achieved by using a fuel plate that has a fuel meat thickness of 0.76 mm, a cladding thickness of 0.38 mm, and a water channel gap of 2.2 mm. A comparison is shown of the main characteristics of this second alternative LEU design with those of the FRM-II HEU design. The ANL core again has one fuel ring with the same dimensions. With this LEU design, a two stage process is no longer necessary because LEU silicide fuel with a uranium density of 4.5 g/cm 3 is fully qualified, licensable, and available now for use in a high flux reactor such as the FRM-II

Fission product release from TRIGA-LEU reactor fuels

International Nuclear Information System (INIS)

Baldwin, N.L.; Foushee, F.C.; Greenwood, J.S.

1980-01-01

Due to present international concerns over nuclear proliferation, TRIGA reactor fuels will utilize only low-enriched uranium (LEU) (enrichment <20%). This requires increased total uranium loading per unit volume of fuel in order to maintain the appropriate fissile loading. Tests were conducted to determine the fractional release of gaseous and metallic fission products from typical uranium-zirconium hydride TRIGA fuels containing up to 45 wt-% uranium. These tests, performed in late 1977 and early 1978, were similar to those conducted earlier on TRIGA fuels with 8.5 wt-% U. Fission gas release measurements were made on prototypic specimens from room temperature to 1100 deg. C in the TRIGA King Furnace Facility. The fuel specimens were irradiated in the TRIGA reactor at a low power level. The fractional releases of the gaseous nuclides of krypton and xenon were measured under steady-state operating conditions. Clean helium was used to sweep the fission gases released during irradiation from the furnace into a standard gas collection trap for gamma counting. The results of these tests on TRIGA-LEU fuel agree well with data from the similar, earlier tests on TRIGA fuel. The correlation used to calculate the release of fission products from 8.5 wt-% U TRIGA fuel applies equally well for U contents up to 45 wt-%. (author)

Establishing a LEU MTR fuel manufacturing facility in South Africa

International Nuclear Information System (INIS)

Jamie, R.W.; Kocher, A.

2010-01-01

The South African MTR Fuel Manufacturing Facility was established in the 1970's to supply SAFARI-1 with Fuel Elements and Control Rods. South African capability was developed in parallel with the uranium enrichment program to meet the needs of the Reactor. Further to the July 2005 decision by the South African Governmnent to convert both SAFARI-1 and the Fuel Plant to LEU, the SAFARI-1 phase has been successfully completed and Necsa has commenced with the conversion of the MTR Fuel Manufacturing Facility. In order to establish, validate and qualify the facility, Necsa has entered into a co-operation and technology transfer agreement with AREVA CERCA, the French manufacturer of Research Reactor fuel elements. Past experiences, conversion challenges and the status of the MTR Fuel Facility Project are discussed. On-going co-operation with AREVA CERCA to implement the local manufacture of LEU fuel is explained and elaborated on. (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

Status of LEU fuel development and conversion of NRU

International Nuclear Information System (INIS)

Sears, D.F.; Herbert, L.N.; Vaillancourt, K.D.

1989-11-01

The status of the low-enrichment uranium (LEU) fuel development and NRU conversion program at Chalk River Nuclear Laboratories is reviewed. Construction of a new fuel fabrication facility is essentially completed and installation of LEW fuel manufacturing equipment has begun. The irradiation of 31 prototype Al-61 wt% U 3 Si dispersion fuel rods, approximately one third of a full NRU core, is continuing without incident. Recent post-irradiation examination of spent fuel rods revealed that the prototype LEU fuel achieved the design burnup (80 at%) in excellent condition, confirming that the Al-U 3 Si 2 dispersion fuel to complement out Al-U 3 Si capability. Three full-size NRU rods containing Al-U 3 Si 2 dispersion fuel have been fabricated for a qualification irradiation in NRU. Post-irradiation examinations of mini-elements containing Al-U 3 Si 2 fuel revealed that the U 3 Si 2 behaved similarly to U 3 Si 2 fuel revealed that the U 3 Si 2 particles and the aluminum matrix, and fission gas bubbles up to 10 μm in diameter, could be seen in the particles after 60 at% and 80 at% burnup. The mini-elements contained a variety of silicide particle sizes; however, no significant swelling dependence on particle size distribution was observed

Preparation results for lifetime test of conversion LEU fuel in plutonium production reactors

International Nuclear Information System (INIS)

Vatulin, A.; Stetskiy, Yu.; Kukharkin, N.; Kalougin, A.; Gavrilov, P.; Ivanov, A.

1999-01-01

The program of converting Russian production reactors for the purpose to stop their plutonium fabrication is currently in progress. The program also provides for operation of these reactors under the conversion mode with using of low-enriched fuel (LEU). LEU fuel elements were developed and activities related to their preparation for reactor tests were carried out. (author)

Interim Report on Mixing During the Casting of LEU-10Mo Plates in the Triple Plate Molds

Energy Technology Data Exchange (ETDEWEB)

Aikin, Jr., Robert M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-04-12

LEU-10%Mo castings are commonly produced by down blending unalloyed HEU with a DU-12.7%Mo master-alloy. This work uses process modeling to provide insight into the mixing of the unalloyed uranium and U-Mo master alloy during melting and mold filling of a triple plate casting. Two different sets of situations are considered: (1) mixing during mold filling from a compositionally stratified crucible and (2) convective mixing of a compositionally stratified crucible during mold heating. The mold filling simulations are performed on the original Y-12 triple plate mold and the horizontal triple plate mold.

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

International Nuclear Information System (INIS)

Travelli, A.

1983-01-01

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

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

A mixed core conversion study with HEU and LEU fuels

International Nuclear Information System (INIS)

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

1985-01-01

The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically. (author)

Status of fuel element technology for plate type dispersion fuels with high uranium density

International Nuclear Information System (INIS)

Hrovat, M.; Huschka, H.; Koch, K.H.; Nazare, S.; Ondracek, G.

1983-01-01

A number of about 20 Material Test and Research Reactors in Germany and abroad is supplied with fuel elements by the company NUKEM. The power of these reactors differs widely ranging from up to about 100 MW. Consequently, the uranium density of the fuel elements in the meat varies considerably depending on the reactor type and is usually within the range from 0.4 to 1.3 g U/cm 3 if HEU is used. In order to convert these reactors to lower uranium enrichment (19.75% 235-U) extensive work is carried out at NUKEM since about two years with the goal to develop fuel elements with high U-density. This work is sponsored by the German Ministry for Research and Technology in the frame of the AF-program. This paper reports on the present state of development for fuel elements with high U-density fuels at NUKEM is reported. The development works were so far concentrated on UAl x , U 3 O 8 and UO 2 fuels which will be described in more detail. In addition fuel plates with new fuels like e.g. U-Si or U-Fe compounds are developed in collaboration with KfK. The required uranium densities for some typical reactors with low, medium, and high power are listed allowing a comparison of HEU and LEU uranium density requirements. The 235-U-content in the case of LEU is raised by 18%. Two different meat thicknesses are considered: Standard thickness of 0.5 mm; and increased thickness of 0.76 mm. From this data compilation the objective follows: in the case of conversion to LEU (19.75% 235-U-enrichment), uranium densities have to be made available up to 24 gU/cm 3 meat for low power level reactors, up to 33 gU/cm 3 meat for medium power level reactors, and between 5.75 and 7.03 g/cm 3 meat for high power level reactors according to this consideration

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

LEU fuel development at CERCA

International Nuclear Information System (INIS)

Durand, Jean Pierre; Ottone, J.C.; Mahe, M.; Ferraz, G.

1998-01-01

The aim of this paper is to detail the recent progress on both U 3 Si 2 high loaded fuels and new γ phase fuels. Concerning high density density silicide plates up to 6 g Ut/cm 3 , the CEA irradiation programme is completed. Data are still under analysis but one can state that the behaviour was globally similar to conventional fuels known in SILOE and OSIRIS reactors. From the new γ fuel point of view, after demonstration feasibility in 1997 of U Mo thermally stable plates loaded up to 8.3 g Ut/cm3, CERCA has analysed the technical ability of quality inspection means assuming that is of an utmost interest for the insurance of a proper use of high performances fuel in reactors. There are mainly two differences between U Mo fuels (and more generally γ fuels) and conventional ones. Firstly, X-ray diffraction analysis on the fuel powder are needed because the chemical analysis is not sufficient to characterise the γ structure requested. Secondly, the physical limits of the Ultrasonic inspection have been reached due to transitory effect between the meat and the edges. Therefore this technic can not applied in the transitory areas. From that knowledge, the manufacture specifications for a plate dedicated to an irradiation plan can be discussed with a clearer view of the main differences with the U 3 Si 2 fuel reference. (author)

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)

Production of leu high density fuels at Babcock and Wilcox

International Nuclear Information System (INIS)

Freim, J.B.

1983-01-01

A large number of fuel elements of all types are produced for both international and domestic customers by Nuclear Fuel Division of Babcock and Wilcox. A brief history of the division, included previous and present research reactor fuel element fabrication experience is discussed. The manufacturing facilities are briefly described. The fabrication of LEU fuels and economic analysis of the production are included. (A.J.)

Comparison of the parameters of the IR-8 reactor with different fuel assembly designs with LEU fuel

International Nuclear Information System (INIS)

Vatulin, A.; Stetsky, Y.; Dobrikova, I.

1999-01-01

The estimation of neutron-physical, heat and hydraulic parameters of the IR-8 research reactor with low enriched uranium (LEU) fuel was performed. Two fuel assembly (FA) designs were reviewed: IRT-4M with the tubular type fuel elements and IRT-MR with the rod type fuel elements. UO 2 -Al dispersion 19.75% enrichment fuel is used in both cases. The results of the calculations were compared with main parameters of the reactor, using the current IRT-3M FA with 90% high enriched uranium (HEU) fuel. The results of these comparisons showed that during the LEU conversion of the reactor the cycle length, excess reactivity and peak power of the IRT-MR type FA are higher than for the IRT-3M type FA and IRT-4M type FA. (author)

Greek research reactor performance characteristics after addition of beryllium reflector and LEU fuel

International Nuclear Information System (INIS)

Deen, J.R.; Snelgrove, J.L.; Papastergiou, C.

1992-01-01

The GRR-1 is a 5-MW pool-type, light-water-moderated and-cooled reactor fueled with MTR-type fuel elements. Recently received Be reflector blocks will soon be added to the core to add additional reactivity until fresh LEU fuel arrives. REBUS-3 xy fuel cycle analyses, using burnup dependent cross sections, were performed to assist in fuel management decisions for the water- and Be-reflected HEU nonequilibrium cores. Cross sections generated by EPRI-CELL have been benchmarked to identical VIM Monte Carlo models. The size of the Be-reflected LEU core has been reduced to 30 elements compared to 35 for the HEU water-reflected core, and an equilibrium cycle calculation has been performed

Neutronic analysis of a reference LEU core for Pakistan research reactor using oxide fuel

International Nuclear Information System (INIS)

Akhtar, K.M.; Qazi, M.K.; Bokhari, I.H.; Khan, L.A.; Pervez, S.

1988-07-01

Neutronic analysis of a 10 MW reference core for PARR, having 28 fresh LEU fuel elements arranged in a 6x5 configuration has been carried out using standard computer codes WIMS-D, EXTERMINATOR-II, and CITATION. Total nuclear power peaking of 3.2 has bee found to occur in the fuel plate adjacent to the water filled central flux trap at the depth of 43.8 cm from the top of the active core. Replacement of water in central flux trap with an aluminum block, having a 50 mm diameter water filled irradiation channel changes the flux profiles in fuel, core side flux trap and reflector. The thermal flux in the central flux trap decreases by about 53%. Therefore some of the fuel elements will have to be removed and the new configuration has to be analysed to determine the first operating core. However, after achieving some burn-up and confirmation from thermal hydraulic analysis, the core configuration analysed, will be the final working core. (orig./A.B.)

Fuel element burnup determination in HEU-LEU mixed TRIGA research reactor core

International Nuclear Information System (INIS)

Zagar, Tomaz; Ravnik, Matjaz

2000-01-01

This paper presents the results of a burnup calculations and burnup measurements for TRIGA FLIP HEU fuel elements and standard TRIGA LEU fuel elements used simultaneously in small TRIGA Mark II research reactor in Ljubljana, Slovenija. The fuel element burnup for approximately 15 years of operation was calculated with two different in house computer codes TRIGAP and TRIGLAV (both codes are available at OECD NEA Data Bank). The calculation is performed in one-dimensional radial geometry in TRIGAP and in two-dimensional (r,φ) geometry in TRIGLAV. Inter-comparison of results shows important influence of in-core water gaps, irradiation channels and mixed rings on burnup calculation accuracy. Burnup of 5 HEU and 27 LEU fuel elements was also measured with reactivity method. Measured and calculated burnup values are inter-compared for these elements (author)

Further data of silicide fuel for the LEU conversion of JMTR

International Nuclear Information System (INIS)

Saito, M.; Futamura, Y.; Nakata, H.; Ando, H.; Sakurai, F.; Ooka, N.; Sakakura, A.; Ugajin, M.; Shirai, E.

1990-01-01

Silicide fuel data for the safety assessment of the JMTR LEU fuel conversion are being measured. The data include fission product release, thermal properties, behaviour under accident conditions, and metallurgical characteristics. The methods used in the experiments are discussed. Results of fission products release at high temperature are described. The release of iodine from the silicide fuel is considerably lower than for U-Al alloy fuel

Greenfield Alternative Study LEU-Mo Fuel Fabrication Facility

Energy Technology Data Exchange (ETDEWEB)

Washington Division of URS

2008-07-01

This report provides the initial “first look” of the design of the Greenfield Alternative of the Fuel Fabrication Capability (FFC); a facility to be built at a Greenfield DOE National Laboratory site. The FFC is designed to fabricate LEU-Mo monolithic fuel for the 5 US High Performance Research Reactors (HPRRs). This report provides a pre-conceptual design of the site, facility, process and equipment systems of the FFC; along with a preliminary hazards evaluation, risk assessment as well as the ROM cost and schedule estimate.

Neutronic analysis of the JMTR with LEU fuel and burnable poison

Energy Technology Data Exchange (ETDEWEB)

Nagaoka, Yoshiharu; Oyamada, Rokuro [Japan Atomic Energy Research Institute, Oarai-machi Ibaraki-ken (Japan); Matos, J E; Woodruff, W L [Argonne National Laboratory, Argonne, IL (United States)

1985-07-01

The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (author)

Neutronic analysis of the JMTR with LEU fuel and burnable poison

International Nuclear Information System (INIS)

Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

1985-01-01

The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed (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)

Conversion and start up of Tehran Research Reactor with LEU fuel

International Nuclear Information System (INIS)

Zaker, M.

2004-01-01

The MW Tehran Research Reactor, Highly Enriched Uranium (HEU) fuel has been converted to Low Enriched Uranium (LEU) fuel using U 3 0 8 -Al with less than 20% enriched uranium. Measured value of excess reactivity, control rod worth and other parameters indicate good agreement with computational predictions. (author)

TRIGA high wt -% LEU fuel development program. Final report

International Nuclear Information System (INIS)

West, G.B.

1980-07-01

The principal purpose of this work was to investigate the characteristics of TRIGA fuel where the contained U-235 was in a relatively high weight percent (wt %) of LEU (low enriched uranium - enrichment of less than 20%) rather than a relatively low weight percent of HEU (high enriched uranium). Fuel with up to 45 wt % U was fabricated and found to be acceptable after metallurgical examinations, fission product retention tests and physical property examinations. Design and safety analysis studies also indicated acceptable prompt negative temperature coefficient and core lifetime characteristics for these fuels

Post-irradiation studies of test plates for low enriched fuel elements for research reactors

International Nuclear Information System (INIS)

Groos, E.; Buecker, H.J.; Derz, H.; Schroeder, R.

1988-07-01

In developing new fuels for research reactor elements that allow the use of low enriched uranium (LEU) 3 Si 2 , U 3 Si 1.5 , U 3 Si 1.3 and U 3 Si. Even up to high burnup rates (80% fifa) U 3 Si 2 was proved to be a reliable fuel that according to the test results achieved to date complies with all necessary requirements above all with respect to dimensional stability. U 3 Si showed significant changes of the fuel microstructure associated with considerably higher fuel swelling, that will probably exclude its use in research reactor operation. The irradiation of U 3 Si 1.3 and U 3 Si 1.5 plates had to be terminated untimely. Up to a burnup of 40% fifa these plates behaved quite well. An extrapolation to higher burnup rates, however only seems to be possible with reservations. (orig./HP) [de

Techno-economic study on conversion of SAFARI-1 to LEU silicide fuel

International Nuclear Information System (INIS)

Ball, G.; Malherbe, F.J.

2004-01-01

This paper marks the conclusion of the techno-economic study into the conversion of SAFARI-1 reactor in South Africa to LEU silicide fuel. Several different fuel types were studied and their characteristics compared to the current HEU fuel. The technical feasibility of operating SAFARI-1 with the different fuels as well as the overall economic impact of the fuels is discussed and conclusions drawn.(author)

Main results and status of the development of LEU fuel for Russian research reactors

International Nuclear Information System (INIS)

Vatulin, A.; Morozov, A.; Suprun, V.; Dobrikova, I.

2005-01-01

VNIINM develops low enrichment uranium (LEU) fuel on base U-Mo alloys and a novel design of pin-type fuel elements. The development is carried out both for existing reactors, and for new advanced designs of reactors. The work is carried on the following main directions: - irradiate LEU U-Mo dispersion fuel (the uranium density up to 6,0 g/cm 3 ) in two Russian research reactors: MIR (RIAR, Dimitrovgrad) as pin type fuel mini-elements and in WWR-M (PINP, Gatchina) within full-scaled fuel assembly (FA) with pin type fuel elements; - finalize development of design and fabrication process of IRT type FA with pin type fuel elements; - develop methods of reducing of U-Mo fuel --Al matrix interaction under irradiation; - develop fabricating methods of fuel elements on base of monolithic U-Mo fuel. The paper generally reviews the results of calculation, design and technology investigations accomplished by now. (author)

Neutronic analysis of the JMTR with LEU fuel and burnable poison

International Nuclear Information System (INIS)

Nagaoka, Yoshiharu; Oyamada, Rokuro; Matos, J.E.; Woodruff, W.L.

1984-01-01

The results of neutronics calculations are presented for the JMTR equilibrium core with LEU silicide fuel, boron and cadmium burnable poisons in the sideplates, and a cycle length of 24 days instead of 11 days with the current HEU fuel. The data indicate that several options are feasible provided that silicide fuels with high uranium densities are successfully demonstrated and licensed. 2 refs., 10 figs., 5 tabs

Neutronic design of a LEU [low enriched uranium] core for the Ohio State University research reactor

International Nuclear Information System (INIS)

Seshadri, M.D.; Aybar, H.S.; Aldemir, T.

1987-01-01

The 10 kw HEU fuelled Ohio State University Reactor (OSURR) will be upgraded to operate at 500 kW with standardized 125 g 235 U LEU U 3 Si 2 fuel plates. An earlier scoping study based on two-dimensional diffusion calculations has identified the potential LEU core configurations for the conversion/upgrade of OSURR using the standardized plates in a 16-plate (+ 2 dummy plates) standard and 10-scoping study is improved for a more precise determination of the excess reactivities and safety rod worths for these potential configurations. Comparison of the results obtained by the improved model to experimental results and to the results of full-core Monte Carlo simulations shows excellent agreement. The results also indicate that the conversion/upgrade of OSURR can be realized with three possible LEU core configurations while maintaining a cold, clean shutdown margin of 1.57-1.91 % Δ k/k, depending on the configuration used. (Author)

SMOPY, a new NDA tool for safeguards of LEU and MOX spent fuel

International Nuclear Information System (INIS)

Lebrun, A.; Merelli, M.; Szabo, J.-L.; Huver, M.; Arenas-Carrasco, J.

2001-01-01

Upon IAEA request, the French support program to IAEA Safeguards has developed a new device for control of the irradiated LEU and MOX fuels. The Safeguards Mox Python (SMOPY) is the achievement of a 4 years R and D program supported by CEA and COGEMA in partnership with Eurisys Mesures. The SMOPY system is based on the combination of 2 NDA techniques (passive neutron and room temperature gamma spectrometry) and on line interpretation tools (automatic gamma spectrum interpretation, depletion code EVO). Through the measurement managing software, all this contributes to the fully automatic measurement, interpretation and characterization of any kind of spent fuel. The device is transportable (50 kg, 60 cm) and is composed of four parts: 1. the measurement head with one high efficiency fission chamber and a micro room temperature gamma spectrometric probe; 2. the carrier which carries the measurement head. The carrier bottom fits the racks for accurate positioning and its top fits operator's fuel moving tool; 3. the portable electronic cabinet which includes both neutron and gamma electronic cards; 4. the portable PC which gets inspectors data, controls the measurement, get measured values, interprets them and immediately provides the inspector with worthwhile info for appropriate on the field decisions. Main features of SMOPY are: Discrimination of MOX versus LEU irradiated fuels in any case (conservative case is one cycle MOX versus three cycles LEU after short cooling time); Full characterization of irradiated LEU (burnup, cooling time, Pu amounts ...); Partial Defect Test on LEU fuels. A first version of SMOPY has been tested in industrial condition during summer 2000. This tests shown a need of shielding improvement around the gamma detector. A new version has been build a will be qualified during a new field test and then the system will be ready for routine operation in IAEA and commercial delivery. After giving details about the system itself, this paper

Thermal-hydraulic analysis of research reactor core with different LEU fuel types using RELAP5

Energy Technology Data Exchange (ETDEWEB)

El-Sahlamy, Neama M. [Nuclear and Radiological Regulatory Authority, Cairo (Egypt)

2017-11-15

In the current work, comparisons between the core performances when using different LEU fuels are done. The fuels tested are UA1{sub X}-A1, U{sub 3}O{sub 8}-Al, and U{sub 3}Si{sub 2}-Al fuels with 19.7 % enrichment. Calculations are done using RELAP5 code to evaluate the thermal-hydraulic performance of the IAEA benchmark 10 MW reactor. First, a reassessment of the slow reactivity insertion transient with UA1{sub X}-A1 LEU fuel to compare the results with those reported in the IAEA TECDOC [1]. Then, comparisons between the thermal-hydraulic core performances when using the three LEU fuels are done. The assessment is performed at initial power of 1.0 W. The reactor power is calculated using the RELAP5 point kinetic model. The reactivity feedback, from changes in water density and fuel temperature, is considered for all cases. From the results it is noticed that U{sub 3}Si{sub 2}-Al fuel gives the best fuel performance since it has the minimum value of peak fuel temperature and the minimum peak clad surface temperature, as operating parameters. Also, it gives the maximum value of the Critical Heat Flux Ratio and the lowest tendency to flow instability occurrence.

Neutronics analysis of the proposed 25-MW leu TRIGA Multipurpose Research Reactor

International Nuclear Information System (INIS)

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

1982-01-01

More than two years ago the government of Indonesia announced plans to purchase a research reactor for the Puspiptek Research Center in Serpong Indonesia to be used for isotope production, materials testing, neutron physics measurements, and reactor operator training. Reactors using low-enriched uranium (LEU) plate-type and rod-type fuel elements were considered. This paper deals with the neutronic evaluation of the rod-type 25-MW LEU TRIGA Multipurpose Research Reactor (MPRR) proposed by the General Atomic Company of the United States of America

Safety analysis of JMTR LEU fuel core, (3)

International Nuclear Information System (INIS)

Tsuchida, Noboru; Shiraishi, Tadao; Takahashi, Yutaka; Inada, Seiji; Saito, Minoru; Futamura, Yoshiaki; Kitano, Kyoshiro.

1992-10-01

Dose analysis in the safety evaluation and the site evaluation were performed for the JMTR core conversion from MEU fuel to LEU fuel. In the safety evaluation, the effective dose equivalents for the public surrounding the site were estimated in fuel handling accident and flow blockage to coolant channel which were selected as the design basis accidents with release of radioactive fission products to the environment. In the site evaluation, the flow blockage to coolant channel was selected as siting basis events, since this accident had the possibility of spreading radioactive release. Maximum exposure doses for the public were estimated assuming large amounts of fission products to release. It was confirmed that risk of radiation exposure of the public is negligible and the siting is appropriate. (author)

Chilean fuel elements fabrication progress report

International Nuclear Information System (INIS)

Baeza, J.; Contreras, H.; Chavez, J.; Klein, J.; Mansilla, R.; Marin, J.; Medina, R.

1993-01-01

Due to HEU-LEU core conversion necessity for the Chilean MTR reactors, the Fuel Elements Plant is being implemented to LEU nuclear fuel elements fabrication. A glove box line for powder-compact processing designed at CCHEN, which supposed to operate under an automatic control system, is at present under initial tests. Results of first natural uranium fuel plates manufacturing runs are shown

A comparative and predictive study of the annual fuel cycle costs for HEU and LEU fuels in the High Flux Reactor, Petten, 1985-1993

Energy Technology Data Exchange (ETDEWEB)

Moss, R L; May, P

1985-07-01

The internationally agreed constraint on availability of supply of HEU fuels to Research and Test Reactors has necessitated that a cost analysis be carried out to determine the financial effect of converting the core of the HFR from HEU to LEU fuels. A computer program, written at Petten and based on information extracted from studies in Europe and the USA, identifies the major cost variables to be manufacturing, uranium, reprocessing and transport costs. Comparison between HEU and LEU cores have been carried out and includes the effects of inflation and exchange rate fluctuations. Conversion of the HFR core to LEU fuels is shown to be financially disadvantageous. (author)

A comparative and predictive study of the annual fuel cycle costs for HEU and LEU fuels in the High Flux Reactor, Petten, 1985-1993

International Nuclear Information System (INIS)

Moss, R.L.; May, P.

1985-01-01

The internationally agreed constraint on availability of supply of HEU fuels to Research and Test Reactors has necessitated that a cost analysis be carried out to determine the financial effect of converting the core of the HFR from HEU to LEU fuels. A computer program, written at Petten and based on information extracted from studies in Europe and the USA, identifies the major cost variables to be manufacturing, uranium, reprocessing and transport costs. Comparison between HEU and LEU cores have been carried out and includes the effects of inflation and exchange rate fluctuations. Conversion of the HFR core to LEU fuels is shown to be financially disadvantageous. (author)

Irradiation Experiment Conceptual Design Parameters for NBSR Fuel Conversion

Energy Technology Data Exchange (ETDEWEB)

Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Brown, N. R. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Baek, J. S [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Cheng, L. Y. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.; Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States). Nuclear Science and Technology Dept.

2014-04-30

It has been proposed to convert the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, from high-enriched uranium (HEU) fuel to low-Enriched uranium (LEU) fuel. The motivation to convert the NBSR to LEU fuel is to reduce the risk of proliferation of special nuclear material. This report is a compilation of relevant information from recent studies related to the proposed conversion using a metal alloy of LEU with 10 w/o molybdenum. The objective is to inform the design of the mini-plate and full-size-Plate irradiation experiments that are being planned. This report provides relevant dimensions of the fuel elements, and the following parameters at steady state: average and maximum fission rate density and fission density, fuel temperature distribution for the plate with maximum local temperature, and two-dimensional heat flux profiles of fuel plates with high power densities. The latter profiles are given for plates in both the inner and outer core zones and for cores with both fresh and depleted shim arms (reactivity control devices). A summary of the methodology to obtain these results is presented. Fuel element tolerance assumptions and hot channel factors used in the safety analysis are also given.

Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

International Nuclear Information System (INIS)

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

1995-01-01

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the technical specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort. (author)

Status report on conversion of the Georgia Tech Research Reactor to low enrichment fuel

International Nuclear Information System (INIS)

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

1991-01-01

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. The US Department of Energy is funding a program to compare reactor performance with high and low enrichment fuels. The goals of the program are: (1) to amend the SAR and the Technical Specifications of the GTRR so that LEU U 3 Si 2 -Al dispersion fuel plates can replace the current HEU U-Al alloy fuel, and (2) to optimize the LEU core such that maximum value neutron beams can be extracted for possible neutron capture therapy application. This paper presents a status report on the LEU conversion effort

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

A comparison of the radiological consequences of a HEU and LEU fueled research reactor

International Nuclear Information System (INIS)

Kollas, J.G.

1985-01-01

An analysis of the design basis accident radiological consequences of the HEU and LEU fueled Greek Research Reactor is presented. Doses and individual cancer risk from exposure to the passing radioactive cloud are estimated up to a distance of 20 km from the reactor site. Collective exposure and latent health effects are estimated for the total Athens area of 3081000 inhabitants. The results indicate that the plutonium isotopes buildup in the LEU fuel does not increase appreciably the consequences in respect to the HEU fueled reactor. The plutonium impact concerns mainly bone effects and secondly lung and whole body effects. The contribution to the limiting thyroid dose and the corresponding thyroid effects is insignificant. (author)

Irradiation of MEU and LEU test fuel elements in DR 3

International Nuclear Information System (INIS)

Haack, K.

1984-01-01

Irradiation of three MEU and three LEU fuel elements in the Danish reactor DR 3. Thermal and fast neutron flux density scans of the core have been made and the results, related to the U235-content of each fuel element, are compared with the values from HEU fuel elements. The test elements were taken to burn-up percentages of 50-60%. Reactivity values of the test elements at charge and at discharge have been measured and the values are compared with those of HEU fuel elements. (author)

Experience in producing LEU fuel elements for the RSG-GAS

International Nuclear Information System (INIS)

Suripto, A.; Soentono, S.

1991-01-01

To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U 3 O 8 -Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

Fuel Management Strategies for a Possible Future LEU Core of a TRIGA Mark II Vienna

Energy Technology Data Exchange (ETDEWEB)

Khan, R.; Villa, M.; Steinhauser, G.; Boeck, H. [Vienna University of Technology-Atominstitut (Austria)

2011-07-01

The Vienna University of Technology/Atominstitut (VUT/ATI) operates a TRIGA Mark II research reactor. It is operated with a completely mixed core of three different types of fuel. Due to the US fuel return program, the ATI have to return its High Enriched Uranium (HEU) fuel latest by 2019. As an alternate, the Low Enrich Uranium (LEU) fuel is under consideration. The detailed results of the core conversion study are presented at the RRFM 2011 conference. This paper describes the burn up calculations of the new fuel to predict the future burn up behavior and core life time. It also develops an effective and optimized fuel management strategy for a possible future operation of the TRIGA Mark II with a LEU core. This work is performed by the combination of MCNP5 and diffusion based neutronics code TRIGLAV. (author)

LEU-fueled SLOWPOKE-2 modelling with MCNP4A

International Nuclear Information System (INIS)

Pierre, J.R.M.; Bonin, H.W.J.

1996-01-01

Following the commissioning of the Low Enrichment Uranium (LEU) Fueled SLOWPOKE-2 research reactor at Royal Military College,excess reactivity measurements were conducted over a range of temperature and power. Given the advance in computer technology, the use of Monte Carlo N-Particle Transport Code System MCNP 4A appeared possible for the simulation of the LEU-fueled SLOWPOKE-2 reactor core, and this work demonstrates that this is indeed the case. MCNP 4A is a full three dimensional program allowing the user to enter a large amount of complexity. The limit on the geometry complexity is the computing time required to achieve a reasonable standard deviation. To this point several models of the SLOWPOKE-2 have been developed giving some insight on the sensitivity of the code. MCNP4A can use various cross section libraries. The aim of this work is to calculate accurately the reactivity of the core and reproduce The temperature trend of the reactivity. The model preserved as much as possible the details of the core and facility in order to allow further study in the flux mapping

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)

Fuel element burnup measurements for the equilibrium LEU silicide RSG GAS (MPR-30) core under a new fuel management strategy

International Nuclear Information System (INIS)

Pinem, Surian; Liem, Peng Hong; Sembiring, Tagor Malem; Surbakti, Tukiran

2016-01-01

Highlights: • Burnup measurement of fuel elements comprising the new equilibrium LEU silicide core of RSG GAS. • The burnup measurement method is based on a linear relationship between reactivity and burnup. • Burnup verification was conducted using an in-house, in-core fuel management code BATAN-FUEL. • A good agreement between the measured and calculated burnup was confirmed. • The new fuel management strategy was confirmed and validated. - Abstract: After the equilibrium LEU silicide core of RSG GAS was achieved, there was a strong need to validate the new fuel management strategy by measuring burnup of fuel elements comprising the core. Since the regulatory body had a great concern on the safety limit of the silicide fuel element burnup, amongst the 35 burnt fuel elements we selected 22 fuel elements with high burnup classes i.e. from 20 to 53% loss of U-235 (declared values) for the present measurements. The burnup measurement method was based on a linear relationship between reactivity and burnup where the measurements were conducted under subcritical conditions using two fission counters of the reactor startup channel. The measurement results were compared with the declared burnup evaluated by an in-house in-core fuel management code, BATAN-FUEL. A good agreement between the measured burnup values and the calculated ones was found within 8% uncertainties. Possible major sources of differences were identified, i.e. large statistical errors (i.e. low fission counters’ count rates), variation of initial U-235 loading per fuel element and accuracy of control rod indicators. The measured burnup of the 22 fuel elements provided the confirmation of the core burnup distribution planned for the equilibrium LEU silicide core under the new fuel management strategy.

Heat-transfer analysis of the existing HEU and proposed LEU cores of Pakistan research reactor

International Nuclear Information System (INIS)

Khan, L.A.; Nabbi, R.

1987-02-01

In connection with conversion of Pakistan Research Reactor (PARR) from the use of Highly Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel, steady-state thermal hydraulic analysis of both existing HEU and proposed LEU cores has been carried out. Keeping in mind the possibility of power upgrading, the performance of proposed LEU core, under 10 MW operating conditions, has also been evaluated. Computer code HEATHYD has been used for this purpose. In order to verify the reliability of the code, IAEA benchmark 2 MW reactor was analyzed. The cooling parameters evaluated include: coolant velocity, critical velocity, pressure drop, temperature distribution in the core, heat fluxes at onset of nucleate boiling, flow instability and burnout and corresponding safety margins. From the results of the study it can be concluded that the conversion of the core to LEU fuel will result in higher safety margins, as compared to existing HEU core, mainly because the increased number of fuel plates in the proposed design will reduce the average heat flux significantly. Anyhow upgrading of the reactor power to 10 MW will need the flow rate to be adjusted between 850 to 900 m 3 /hr, to achieve reasonable safety margins, at least, comparable with the existing HEU core. (orig.)

Facility safeguards at an LEU fuel fabrication facility in Japan

International Nuclear Information System (INIS)

Kuroi, H.; Osabe, T.

1984-01-01

A facility description of a Japanese LEU BWR-type fuel fabrication plant focusing on safeguards viewpoints is presented. Procedures and practices of MC and A plan, measurement program, inventory taking, and the report and record system are described. Procedures and practices of safeguards inspection are discussed and lessons learned from past experiences are reviewed

Fuel cell cooler-humidifier plate

Science.gov (United States)

Vitale, Nicholas G.; Jones, Daniel O.

2000-01-01

A cooler-humidifier plate for use in a proton exchange membrane (PEM) fuel cell stack assembly is provided. The cooler-humidifier plate combines functions of cooling and humidification within the fuel cell stack assembly, thereby providing a more compact structure, simpler manifolding, and reduced reject heat from the fuel cell. Coolant on the cooler side of the plate removes heat generated within the fuel cell assembly. Heat is also removed by the humidifier side of the plate for use in evaporating the humidification water. On the humidifier side of the plate, evaporating water humidifies reactant gas flowing over a moistened wick. After exiting the humidifier side of the plate, humidified reactant gas provides needed moisture to the proton exchange membranes used in the fuel cell stack assembly. The invention also provides a fuel cell plate that maximizes structural support within the fuel cell by ensuring that the ribs that form the boundaries of channels on one side of the plate have ends at locations that substantially correspond to the locations of ribs on the opposite side of the plate.

Low enrichment fuel conversion for Iowa State University

International Nuclear Information System (INIS)

Rohach, A.F.; Hendrickson, R.A.

1990-08-01

Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. The LEU fuel has not been received. The HEU fuel assemblies for the UTR-10 reactor will not fit into any current research reactor shipping containers; therefore, the fuel assemblies must be disassembled and the fuel shipped as fuel plates. Procedures and practices have been developed so that the fuel assemblies will be disassembled in a shielded environment

Transition from HEU to LEU fuel in Romania's 14-MW TRIGA reactor

International Nuclear Information System (INIS)

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

1995-01-01

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 x 5 square array of HEU U (10 wt% - ZrH - Er 2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incoloy. With a total inventory of 35 HEU fuel clusters, burnup, considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an average 235 U burnup in the range from 50 to 62%. Because of the U.S. policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% 235 U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations. (author)

Comparison of thermohydraulic and nuclear aspects in a standard HEU core and a typical LEU core for the HFR Petten. A case study

International Nuclear Information System (INIS)

Pruimboom, H.; Tas, A.

1985-01-01

Within the framework of the RERTR program various HEU-LEU core calculations have been performed by ANL in a cooperative effort with ECN and JRC Petten. The main purpose of this work has been to gain competence in analysing HEU-LEU core conversion for high power Materials Testing Reactors and to assist in a possible HEU-LEU conversion of the HFR Petten. For reference purposes the present HFR standard core (HEU) in the 'old' vessel geometry was calculated at first. As a next step the new vessel geometry and the increased fuel weights were taken into account. Subsequently various LEU HFR core options have been analysed. Main parameters in the LEU study were the uranium loading in the meat, the fuel type, the thickness of the meat, the number of fuel plates per element and the type of burnable poison applied. Though the study has not yet been completed, one of its striking preliminary results concerns the increased power peaking in the LEU fuel elements as compared with the HEU situation. A preliminary analysis of the thermal characteristics of a typical LEU core as compared with a standard HEU core has been made and is presented in the paper. A short survey of the various HEU and LEU calculations is given. The thermal safety analysis procedure for the HFR, as based on the flow instability criterion, is clarified. Finally, the thermal comparison HEU versus LEU and the resulting conclusions are presented. (author)

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

Irradiation of full size UMo plates

International Nuclear Information System (INIS)

Vacelet, H.; Lavastre, Y.; Grasse, M.; Sacristan, P.; Languille, A.

1999-01-01

An important development program for a UMo MTR fuel has been launched in France. The goal of the French working group is to develop a high performing and reprocessable fuel before the end of the US return policy. This paper is focussed on the fabrication of full-sized UMo plates with LEU (Low Enriched Enrichment) and their irradiation in OSIRIS reactor which was started on the 22nd of September. The results of the plates inspection are presented here as well as the irradiation conditions. (author)

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

Preliminary Results of Ancillary Safety Analyses Supporting TREAT LEU Conversion Activities

Energy Technology Data Exchange (ETDEWEB)

Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2015-10-01

Report (FSAR) [3]. Depending on the availability of historical data derived from HEU TREAT operation, results calculated for the LEU core are compared to measurements obtained from HEU TREAT operation. While all analyses in this report are largely considered complete and have been reviewed for technical content, it is important to note that all topics will be revisited once the LEU design approaches its final stages of maturity. For most safety significant issues, it is expected that the analyses presented here will be bounding, but additional calculations will be performed as necessary to support safety analyses and safety documentation. It should also be noted that these analyses were completed as the LEU design evolved, and therefore utilized different LEU reference designs. Preliminary shielding, neutronic, and thermal hydraulic analyses have been completed and have generally demonstrated that the various LEU core designs will satisfy existing safety limits and standards also satisfied by the existing HEU core. These analyses include the assessment of the dose rate in the hodoscope room, near a loaded fuel transfer cask, above the fuel storage area, and near the HEPA filters. The potential change in the concentration of tramp uranium and change in neutron flux reaching instrumentation has also been assessed. Safety-significant thermal hydraulic items addressed in this report include thermally-induced mechanical distortion of the grid plate, and heating in the radial reflector.

Development of new ORIGEN2 data library sets for research reactors with light water cooled oxide and silicide LEU (20 w/o) fuels based on JENDL-3.3 nuclear data

International Nuclear Information System (INIS)

Liem, Peng Hong; Sembiring, Tagor Malem

2013-01-01

Highlights: • We developed new ORIGEN2 data library sets for research reactors based on JENDL-3.3. • The sets cover oxide and silicide LEU fuels with meat density up to 4.74 g U/cm 3 . • Two kinds of data library sets are available: fuel region and non-fuel regions. • We verified the new data library sets with other codes. • We validated the new data library against a non-destructive test. -- Abstract: New sets of ORIGEN2 data library dedicated to research/testing reactors with light water cooled oxide and silicide LEU fuel plates based on JENDL-3.3 nuclear data were developed, verified and validated. The new sets are considered to be an extension of the most recent release of ORIGEN2.2UPJ code, i.e. the ORLIBJ33 library sets. The newly generated ORIGEN2 data library sets cover both oxide and silicide LEU fuels with fuel meat density range from 2.96 to 4.74 g U/cm 3 used in the present and future operation of the Indonesian 30 MWth RSG GAS research reactor. The new sets are expected applicable also for other research/testing reactors which utilize similar fuels or have similar neutron spectral indices. In addition to the traditional ORIGEN2 library sets for fuel depletion analyses in fuel regions, in the new data library sets, new ORIGEN2 library sets for irradiation/activation analyses were also prepared which cover all representative non-fuel regions of RSG GAS such as reflector elements, irradiation facilities, etc. whose neutron spectra are significantly softer than fuel regions. Verification with other codes as well as validation with a non-destructive test result showed promising results where a good agreement was confirmed

The use of U3Si2 dispersed in aluminum in plate-type fuel elements for research and test reactors

International Nuclear Information System (INIS)

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

1987-10-01

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

Foreign research reactor spent nuclear fuel inventories containing HEU and LEU of US-origin

International Nuclear Information System (INIS)

Matos, J.E.

1995-01-01

This paper provides estimates of the quantities and types of foreign research reactor spent nuclear fuel containing HEU and LEU of US-origin that are anticipated during the period beginning in January 1996 and extending for 10-15 years

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)

Status of HEU-LEU conversion of FRJ-2

International Nuclear Information System (INIS)

Damm, G.; Nabbi, R.

2002-01-01

The operator of the German FRJ-2 research reactor, 'Research Center Juelich', has participated from the beginning in the RERTR programme and made comprehensive contributions to the test and use of LEU fuel for HEU-LEU-conversion measures. The originally planned time scale for the conversion of FRJ-2 was significantly delayed because of a change of the manufacturer of the LEU fuel elements and a 4 years shutdown of the reactor for refurbishment purposes. In the meantime the new LEU fuel elements are qualified and tested in the reactor. In the moment calculations for the safety report are made and it is planned to apply for the license of FRJ-2 operation with LEU fuel at the beginning of 2003. In order to get most reliable results a sophisticated computational method based on a MCNP model coupled with the depletion code BURN was developed for reactor physical calculations, core conversion studies and fuel element performance analysis and applied to the mixed and LEU core. The licensing schedule and results of latest calculations for the conversion study will be presented. The simulations shows that the thermal flux in the LEU core is about 19% resulting in a lower burnup rate. But in the reflector area around the core and in the center of the cold n source the neutron flux reduction remains limited to 6%. Due to a harder neutron spectrum in the LEU core the kinetic and safety related parameters are slightly reduced. Using the ORIGEN code it could be shown that the increase of the total fission products inventory amounts to about 6% compared to a HEU core. As a consequence of the high amount of U-238, the amount of U-235 in the LEU core has to be about 27% higher than in the HEU core but the U-235 burnup is approx. 5% lower due to the contribution of fissile plutonium. (author)

Modeling RERTR experimental fuel plates using the PLATE code

International Nuclear Information System (INIS)

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

2003-01-01

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

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1993-01-01

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

Instrumentation of fuel elements and fuel plates

International Nuclear Information System (INIS)

Durand, J.P.; Fanjas, Y.

1994-01-01

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

Selection of LEU/Th reference fuel for the HTGR-SC/C lead plant

International Nuclear Information System (INIS)

Turner, R.F.; Neylan, A.J.; Baxter, A.M.; McEachern, D.W.; Stansfield, O.M.

1983-05-01

This paper describes the reference fuel materials for the high-temperature gas-cooled reactor (HTGR) plant for steam cycle/cogeneration (SC/C). A development and testing program carried out in 1978 through 1982 led to the selection of coated fuel particles of uranium-oxycarbide (UCO) for fissile materials and thorium oxide (ThO 2 ) for fertiel materials. Low-enriched uranium (LEU) is the enrichment basis for the HTGR-SC/C application. While UC 2 and UO 2 would also meet the essential criteria for fissile fuel, the UCO, alternative was selected on the basis of improved performance, economics, and process conditions

LEU fuel fabrication program for the RECH-1 reactor. Status report

International Nuclear Information System (INIS)

Chavez, J.C.; Barrera, M.; Jimenez, O.; Lisboa, J.; Marin, J.

2000-01-01

In 1995 a 50 LEU U 3 Si 2 fuel elements fabrication program for the RECH-1 research reactor was established at the Comision Chilena de Energia Nuclear, CCHEN. After a fabrication process qualification stage, in 1998, four elements were early delivered to the reactor in order to start an irradiation qualification stage. The irradiation has reached an estimated 10% burn-up and no fabrication problems have been detected up to this burn-up level. During 1999 and up to the first quarter of 2000, 19 fuel elements were produced and 7 fuel elements are expected for the end of 2000. This report presents an updated summary of the main results obtained in this fuel fabrication program. A summary of other activities generated by this program, such as in core follow-up of the four leader fuel elements, ISO 9001 implementation for the fabrication process and a fabrication and qualification optimization planning, is also presented here. (author)

A neutronic feasibility study for LEU conversion of the High Flux Beam Reactor (HFBR)

International Nuclear Information System (INIS)

Pond, R.B.; Hanan, N.A.; Matos, J.E.

1997-01-01

A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm 3 in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept. (author)

Key considerations in the conversion to LEU of a Mo-99 commercially producing reactor: SAFARI-1 of South Africa

International Nuclear Information System (INIS)

Stumpf, W.E.; Vermaak, A.P.; Ball, G.

2000-01-01

Apart from the technological demands and considerations associated with the conversion of a Mo-99 commercially producing reactor to LEU, a number of commercial challenges also need to be addressed. This is particularly the case when the reactor is primarily used as a source for the production, on an uninterrupted basis, of significant quantities of Mo-99 to satisfy long term commitments to a range of global customers. This paper highlights key business considerations which are applicable in the conversion process of firstly, reactor fuel to LEU and secondly target plates for Mo-99, also to LEU, using the SAFARI-1 reactor in South Africa as a typical example of such a commercially utilized reactor. (author)

Key considerations in the conversion to LEU of a Mo-99 commercially producing reactor: SAFARI-1 of South Africa

Energy Technology Data Exchange (ETDEWEB)

Stumpf, W E; Vermaak, A P; Ball, G [NECSA, PO Box 582, Pretoria (South Africa)

2000-10-01

Apart from the technological demands and considerations associated with the conversion of a Mo-99 commercially producing reactor to LEU, a number of commercial challenges also need to be addressed. This is particularly the case when the reactor is primarily used as a source for the production, on an uninterrupted basis, of significant quantities of Mo-99 to satisfy long term commitments to a range of global customers. This paper highlights key business considerations which are applicable in the conversion process of firstly, reactor fuel to LEU and secondly target plates for Mo-99, also to LEU, using the SAFARI-1 reactor in South Africa as a typical example of such a commercially utilized reactor. (author)

Neutronic calculations of PARR-1 cores using LEU silicide fuel

International Nuclear Information System (INIS)

Arshad, M.; Bakhtyar, S.; Hayat, T.; Salahuddin, A.

1991-08-01

Detailed neutronic calculations have been carried out for different PARR-1 cores utilizing low enriched uranium (LEU) silicide fuel and operating at an upgraded power of 9 MW. The calculations include the search for critical loadings in open and stall ends of the pool, neutronic analysis of the first full equilibrium core and calculations cores. The burnup study of inventory have also been carried out. Further, the reactivity coefficients of the first full power operation core are evaluated for use in the accident analysis. 14 figs. (author)

Status of high-density fuel plate fabrication

International Nuclear Information System (INIS)

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

1991-01-01

Progress has continued on the fabrication of fuel plates with equivalent fuel zone loadings approaching 9 gU/cm 3 . Through hot isostatic pressing (HIP), successful diffusion bonds have been made with 1100 Al and 6061 Al alloys. Although additional study is necessary to optimize the procedure, these bonds demonstrated the most critical processing step for proof-of-concept hardware. Two types of prototype highly loaded fuel plates have been fabricated. The first is a fuel plate in which 0.030-in. (0.76-mm) uranium compound wires are bonded within an aluminum cladding; the second, a dispersion fuel plate with uniform cladding and fuel zone thickness. The successful fabrication of these fuel plates derives from the unique ability of the HIP process to produce diffusion bonds with minimal deformation. (orig.)

HEU benchmark calculations and LEU preliminary calculations for IRR-1

International Nuclear Information System (INIS)

Caner, M.; Shapira, M.; Bettan, M.; Nagler, A.; Gilat, J.

2004-01-01

We performed neutronics calculations for the Soreq Research Reactor, IRR-1. The calculations were done for the purpose of upgrading and benchmarking our codes and methods. The codes used were mainly WIMS-D/4 for cell calculations and the three dimensional diffusion code CITATION for full core calculations. The experimental flux was obtained by gold wire activation methods and compared with our calculated flux profile. The IRR-1 is loaded with highly enriched uranium fuel assemblies, of the plate type. In the framework of preparation for conversion to low enrichment fuel, additional calculations were done assuming the presence of LEU fresh fuel. In these preliminary calculations we investigated the effect on the criticality and flux distributions of the increase of U-238 loading, and the corresponding uranium density.(author)

Multilateral nonproliferation cooperation: US - Led effort to remove HEU/LEU fresh and spent fuels from the Republic of Georgia to Dounreay, Scotland

International Nuclear Information System (INIS)

Shelton, Thomas A.; Viebrock, James M.; Riedy, Alexander W.; Moses, Stanley D.; Bird, Helen M.

1998-01-01

This paper presents the efforts led by United States for removing HEU/LEU fresh and spent fuel from dhe Republic of Georgia to Dounreay, Scotland. These efforts are resulted from a plan approved by the United States Government, in cooperation with the United Kingdom and Georgia Governments to rapidly retrieve and transport circa 4.3 kilograms of enriched uranium. This material consisted largely of highly enriched uranium (HEU) and a small amount of low enriched uranium (LEU) fresh fuel, as well as about 800 grams of HEU/LEU-based spent fuel from a shutdown IR T-M research reactor on the outskirts of Table's, Georgia. The technical team lead by DOE consisted of HEU handling, packaging and transportation experts from the Oak Ridge Y-12 plant, managed and operated by Lockheed Martin Energy Systems, and fuel handling and transportation experts from Nac International in Norcross, Georgia, United States

Evaluation of Thin Plate Hydrodynamic Stability through a Combined Numerical Modeling and Experimental Effort

Energy Technology Data Exchange (ETDEWEB)

Tentner, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Bojanowski, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Wilson, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Solbrekken, G [Univ. of Missouri, Columbia, MO (United States); Jesse, C. [Univ. of Missouri, Columbia, MO (United States); Kennedy, J. [Univ. of Missouri, Columbia, MO (United States); Rivers, J. [Univ. of Missouri, Columbia, MO (United States); Schnieders, G. [Univ. of Missouri, Columbia, MO (United States)

2017-05-01

An experimental and computational effort was undertaken in order to evaluate the capability of the fluid-structure interaction (FSI) simulation tools to describe the deflection of a Missouri University Research Reactor (MURR) fuel element plate redesigned for conversion to lowenriched uranium (LEU) fuel due to hydrodynamic forces. Experiments involving both flat plates and curved plates were conducted in a water flow test loop located at the University of Missouri (MU), at conditions and geometries that can be related to the MURR LEU fuel element. A wider channel gap on one side of the test plate, and a narrower on the other represent the differences that could be encountered in a MURR element due to allowed fabrication variability. The difference in the channel gaps leads to a pressure differential across the plate, leading to plate deflection. The induced plate deflection the pressure difference induces in the plate was measured at specified locations using a laser measurement technique. High fidelity 3-D simulations of the experiments were performed at MU using the computational fluid dynamics code STAR-CCM+ coupled with the structural mechanics code ABAQUS. Independent simulations of the experiments were performed at Argonne National Laboratory (ANL) using the STAR-CCM+ code and its built-in structural mechanics solver. The simulation results obtained at MU and ANL were compared with the corresponding measured plate deflections.

Neutronic calculations in core conversion of the IAN-R1 research reactor from MTR HEU to TRIGA LEU fuel

International Nuclear Information System (INIS)

Sarta Fuentes, Jose A.; Castiblanco, L.A.

2003-01-01

With cooperation of the International Atomic Energy Agency (IAEA), neutronic calculations were carried out for conversion of the Ian-R1 Reactor from MTR-HEU fuel to TRIGA-LEU fuel. In order to establish a staff for neutronic calculation at the Instituto de Cancan's Nucleares y Energia s Alternatives (INEA) a program was established. This program included training, acquisition of hardware, software and calculation for the core with MTR-HEU fuel , enriched nominally to 93% and calculation for several arrangements with the TRIGA-LEU fuel, enriched to 19.7%. The results were verified and compared with several groups of calculation at the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, and General Atomics (GA) in United States. As a result of this program, several technical reports have been wrote. (author)

Status of LEU conversion program at CRNL

International Nuclear Information System (INIS)

Kennedy, I.C.

1991-01-01

After briefly reviewing the salient features of the NRU Reactor at Chalk River Nuclear Laboratories (CRNL), the progress of our LEU fuel development and testing program is described. The results (to date) of full-size prototype fuel-rod irradiations are reviewed, and the status of the new fuel-fabrication facility on the site is updated. Although development work is proceeding on U 3 Si 2 dispersions, all indications so far are that CRNL's U 3 Si fuel is fully acceptable for reactor operation. Fuel rods from the new fabrication shop will be installed in NRU in 1990, and the complete core conversion of NRU to LEU driver fuel is expected by 1991. (orig.)

MTR fuel plate qualification capabilities at SCK-CEN

International Nuclear Information System (INIS)

Koonen, E.; Jacquet, P.

2002-01-01

In order to enhance the capabilities of BR2 in the field of MTR fuel plate testing, a dedicated irradiation device has been designed. In its basic version this device allows the irradiation of 3 fuel plates. The central fuel plate may be replaced by a dummy plate or a plate carrying dosimeters. A first FUTURE device has been built. A benchmark irradiation has been executed with standard BR2 fuel plates in order to qualify this device. Detailed neutronic calculations were performed and the results compared to the results of the post-irradiation examinations of the plates. These comparisons demonstrate the capability to conduct a fuel plate irradiation program under requested and well-known irradiation conditions. Further improvements are presently being designed in order to extend the ranges of heat flux and surface temperature of the fuel plates that can be handled with the FUTURE device. (author)

Neutronic analysis for conversion of the Ghana Research Reactor-1 facility using Monte Carlo methods and UO{sub 2} LEU fuel

Energy Technology Data Exchange (ETDEWEB)

Anim-Sampong, S.; Akaho, E.H.K.; Maakuu, B.T.; Gbadago, J.K. [Ghana Research Reactor-1 Centre, Dept. of Nuclear Engineering and Materials Science, National Nuclear Research Institute, Ghana Atomic Energy Commission, Legon, Accra (Ghana); Andam, A. [Kwame Nkrumah Univ. of Science and Technology, Dept. of Physics (Ghana); Liaw, J.J.R.; Matos, J.E. [Argonne National Lab., RERTR Programme, Div. of Nuclear Engineering (United States)

2007-07-01

Monte Carlo particle transport methods and software (MCNP) have been applied to the modelling, simulation and neutronic analysis for the conversion of the HEU-fuelled (high enrichment uranium) core of the Ghana Research Reactor-1 (GHARR-1) facility. The results show that the MCNP model of the GHARR-1 facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR) is good as the simulated neutronic and other reactor physics parameters agree with very well with experimental and zero power results. Three UO{sub 2} LEU (low enrichment uranium) fuels with different enrichments (12.6% and 19.75%), core configurations, core loadings were utilized in the conversion studies. The nuclear criticality and kinetic parameters obtained from the Monte Carlo simulation and neutronic analysis using three UO{sub 2} LEU fuels are in close agreement with results obtained for the reference 90.2% U-Al HEU core. The neutron flux variation in the core, fission chamber and irradiation channels for the LEU UO{sub 2} fuels show the same trend as the HEU core as presented in the paper. The Monte Carlo model confirms a reduction (8% max) in the peak neutron fluxes simulated in the irradiation channels which are utilized for experimental and commercial activities. However, the reductions or 'losses' in the flux levels neither affects the criticality safety, reactor operations and safety nor utilization of the reactor. Employing careful core loading optimization techniques and fuel loadings and enrichment, it is possible to eliminate the apparent reductions or 'losses' in the neutron fluxes as suggested in this paper. Concerning neutronics, it can be concluded that all the 3 LEU fuels qualify as LEU candidates for core conversion of the GHARR-1 facility.

Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

International Nuclear Information System (INIS)

Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung

2015-01-01

The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW th power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H 2 corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the preliminary design

Manufacturing and investigation of U-Mo LEU fuel granules by hydride-dehydride processing

International Nuclear Information System (INIS)

Stetskiy, Y.A.; Trifonov, Y.I.; Mitrofanov, A.V.; Samarin, V.I.

2002-01-01

Investigations of hydride-dehydride processing for comminution of U-Mo alloys with Mo content in the range 1.9/9.2% have been performed. Some regularities of the process as a function of Mo content have been determined as well as some parameters elaborated. Hydride-dehydride processing has been shown to provide necessary phase and chemical compositions of U-Mo fuel granules to be used in disperse fuel elements for research reactors. Pin type disperse mini-fuel elements for irradiation tests in the loop of 'MIR' reactor (Dmitrovgrad) have been fabricated using U-Mo LEU fuel granules obtained by hydride-dehydride processing. Irradiation tests of these mini-fuel elements loaded to 4 g U tot /cm 3 are planned to start by the end of this year. (author)

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

Analysis of the TREAT LEU Conceptual Design

Energy Technology Data Exchange (ETDEWEB)

Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Kontogeorgakos, D. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Papadias, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mo, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Strons, P. S. [Argonne National Lab. (ANL), Argonne, IL (United States); Fei, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Wright, A. E. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-03-01

Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Management and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO₂-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.

A neutronic feasibility study for LEU conversion of the SAFARI-1 reactor

International Nuclear Information System (INIS)

Pond, R.B.; Hanan, N.A.; Matos, J.E.; Ball, G.

2000-01-01

A neutronic feasibility study to convert the SAFARI-1 reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with NECSA. Comparisons were made of the reactor performance with the current 90% enriched HEU fuel type (UAl) and two 19.75% enriched LEU fuel types (U 3 Si 2 and U7Mo). The thermal fluxes with the LEU fuels were 3 - 9% lower than with the current HEU fuel. For the same fuel assembly design, a uranium density of approximately 4.5 g/cm 3 was required with U 3 Si 2 -Al fuel and a uranium density of about 4.6 g/cm 3 was required with U7Mo-Al fuel to match the 24.6-day cycle of the UAl-alloy fuel with 0.92 gU/cm 3 . The selection of a suitable LEU fuel and the decision to convert SAFARI-1 will be an economic matter that depends upon the fuel type, fuel assembly design, experiment performance and fuel cycle costs. (author)

US Progress on Property Characterization to Support LEU U-10 Mo Monolithic Fuel Development

Energy Technology Data Exchange (ETDEWEB)

Cole, James Irvin [Idaho National Laboratory; Rabin, Barry H [Idaho National Laboratory; Smith, James Arthur [Idaho National Laboratory; Scott, Clark Landon [Idaho National Laboratory; Benefiel, Bradley Curtis [Idaho National Laboratory; Larsen, Eric David [Idaho National Laboratory; Lind, Robert Paul [Idaho National Laboratory; Sell, David Alan [Idaho National Laboratory

2016-03-01

The US High Performance Research Reactor program is pursuing development and qualification of a new high density monolithic LEU fuel to facilitate conversion of five higher power research reactors located in the US (ATR, HFIR, NBSR, MIT and MURR). In order to support fabrication development and fuel performance evaluations, new testing capabilities are being developed to evaluate the properties of fuel specimens. Residual stress and fuel-cladding bond strength are two characteristics related to fuel performance that are being investigated. In this overview, new measurement capabilities being developed to assess these characteristics in both fresh and irradiated fuel are described. Progress on fresh fuel testing is summarized and on-going hot-cell implementation efforts to support future PIE campaigns are detailed. It is anticipated that benchmarking of as-fabricated fuel characteristics will be critical to establishing technical bases for specifications that optimize fuel fabrication and ensure acceptable in-reactor fuel performance.

Preliminary Thermohydraulic Analysis of a New Moderated Reactor Utilizing an LEU-Fuel for Space Nuclear Thermal Propulsion

Energy Technology Data Exchange (ETDEWEB)

Nam, Seung Hyun; Choi, Jae Young; Venneria, Paolo F.; Jeong, Yong Hoon; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of)

2015-10-15

The Korea Advanced NUclear Thermal Engine Rocket utilizing an LEU fuel (KANUTER-LEU) is a non-proliferative and comparably efficient NTR engine with relatively low thrust levels of 40 - 50 kN for in-space transportation. The small modular engine can expand mission versatility, when flexibly used in a clustered engine arrangement, so that it can perform various scale missions from low-thrust robotic science missions to high-thrust manned missions. In addition, the clustered engine system can enhance engine redundancy and ensuing crew safety as well as the thrust. The propulsion system is an energy conversion system to transform the thermal energy of the reactor into the kinetic energy of the propellant to produce the powers for thrust, propellant feeding and electricity. It is mainly made up of a propellant Feeding System (PFS) comprising a Turbo-Pump Assembly (TPA), a Regenerative Nozzle Assembly (RNA), etc. For this core design study, an expander cycle is assumed to be the propulsion system. The EGS converts the thermal energy of the EHTGR in the idle operation (only 350 kW{sub th} power) to electric power during the electric power mode. This paper presents a preliminary thermohydraulic design analysis to explore the design space for the new reactor and to estimate the referential engine performance. The new non-proliferative NTR engine concept, KANUTER-LEU, is under designing to surmount the nuclear proliferation obstacles on allR and Dactivities and eventual commercialization for future generations. To efficiently implement a heavy LEU fuel for the NTR engine, its reactor design innovatively possesses the key characteristics of the high U density fuel with high heating and H{sub 2} corrosion resistances, the thermal neutron spectrum core and also minimizing non-fission neutron loss, and the compact reactor design with protectively cooling capability. To investigate feasible design space for the moderated EHTGR-LEU and resultant engine performance, the

Calculation of plate temperatures in a Mk 4 LEU fuel element

International Nuclear Information System (INIS)

Haack, K.

1988-09-01

A calculation method for estimating the axial temperature distributions of each tube in each of the 26 fuel elements of the DR 3 core is described and demonstrated. With input data for fuel element power, D2O outlet temperature and main D2O circulator combination, a computer code calculates all important temperatures in the fuel element. 11 tabs., 32 ills. 8 refs. (author)

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)

Calculation of plate temperatures in a Mk 4 LEU fuel element

International Nuclear Information System (INIS)

Haack, K.

1991-10-01

A calculation method for estimating the axial temperature distributions of each tube in each of the 26 fuel elements of the DR 3 core is described and demonstrated. With input data for fuel element power, D 2 O outlet temperature and main D 2 O circulator combination, a computer code calculates all important temperatures in the fuel element. Preface to Second Edition Oct. 1991. The second edition is based on the more reliable thermophysical heavy water properties made available by the investigations of Professor J. Bukovsky. The values in the tables are replaced and a new set of fuel element temperature curves is enclosed as an example of the temperature distributions in a low enriched uranium (19,8% 235 U as U 3 Si 2 ). (author) 11 tabs., 32 ills., 9 refs

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

Evolution of fuel plate parameters during deformation in rolling

Energy Technology Data Exchange (ETDEWEB)

Durazzo, M., E-mail: mdurazzo@ipen.br [Nuclear and Energy Research Institute – IPEN/CNEN-SP, São Paulo (Brazil); Vieira, E.; Urano de Carvalho, E.F. [Nuclear and Energy Research Institute – IPEN/CNEN-SP, São Paulo (Brazil); Riella, H.G. [Nuclear and Energy Research Institute – IPEN/CNEN-SP, São Paulo (Brazil); Chemical Engineering Department, Santa Catarina Federal University, Florianópolis (Brazil)

2017-07-15

The Nuclear and Energy Research Institute – IPEN/CNEN-SP routinely produces the nuclear fuel necessary for operating its research reactor, IEA-R1. This fuel consists of fuel plates containing U{sub 3}Si{sub 2}-Al composites as the meat, which are fabricated by rolling. The rolling process currently deployed was developed based on information obtained from literature, which was used as a premise for defining the current manufacturing procedures, according to a methodology with an essentially empirical character. Despite the current rolling process being perfectly stable and highly reproducible, it is not well characterized and is therefore not fully known. The objective of this work is to characterize the rolling process for producing dispersion fuel plates. Results regarding the evolution of the main parameters of technological interest, after each rolling pass, are presented. Some defects that originated along the fuel plate deformation during the rolling process were characterized and discussed. The fabrication procedures for manufacturing the fuel plates are also presented. - Highlights: •Evolution of defects when manufacturing dispersion fuel plates. •Aspects of dispersion fuel plates fabrication. •What happen during the manufacturing of dispersion fuel plates? •Clarifying the deformation of fuel plates by rolling.

HEU/LEU-conversion of BER II successfully finished

International Nuclear Information System (INIS)

Haas, K.; Fischer, C.-O.; Krohn, H.

2000-01-01

The BER II (Berliner Experimental Reactor) research reactor is a swimming pool type reactor located in Berlin, Germany. The reactor operates with a thermal power of 10 MW and is primarily used to produce neutrons for neutron scattering experiments. The conversion from HEU- to LEU-fuel elements began in August, 1997. At the last RERTR Meeting 1999 in Budapest, Hungary, Hahn-Meitner-Institut (HMI) presented a 'Status Report' on the conversion of 10 HEU/LEU mixed cores. In February 2000, HMI finished the HEU/LEU-conversion. Hereby, the first pure LEU-standard-core went into operation. Our second LEU-core just ends its operation at the end of July. The third LEU-core will be built up in the beginning of August. The average burn-up rate was improved from 50 - 55% (HEU) to 60 - 65% (LEU). Therefore, only 14 elements/year are now used instead of 28/year. The following report describes our first steps in building pure LEU-cores from mixed HEU/LEU-cores, as well as our initial experience using the pure LEU-cores. (author)

Pilot plant production at Riso of LEU silicide fuel for the Danish reactor DR3

International Nuclear Information System (INIS)

Toft, P.; Borring, J.; Adolph, E.

1988-01-01

A pilot plant for fabricating LEU silicide fuel elements has been established at Riso National Laboratory. Three test elements for the Danish reactor DR3 have been fabricated, based on 19.88% enriched U 3 Si 2 powder that has been purchased elsewhere. The pilot plant has been set up and 3 test elements fabricated without any major difficulties

A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR).

Energy Technology Data Exchange (ETDEWEB)

Hanan, N. A.

1998-01-14

A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm{sup 3} or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.

Fuel loading and homogeneity analysis of HFIR design fuel plates loaded with uranium silicide fuel

International Nuclear Information System (INIS)

Blumenfeld, P.E.

1995-08-01

Twelve nuclear reactor fuel plates were analyzed for fuel loading and fuel loading homogeneity by measuring the attenuation of a collimated X-ray beam as it passed through the plates. The plates were identical to those used by the High Flux Isotope Reactor (HFIR) but were loaded with uranium silicide rather than with HFIR's uranium oxide fuel. Systematic deviations from nominal fuel loading were observed as higher loading near the center of the plates and underloading near the radial edges. These deviations were within those allowed by HFIR specifications. The report begins with a brief background on the thermal-hydraulic uncertainty analysis for the Advanced Neutron Source (ANS) Reactor that motivated a statistical description of fuel loading and homogeneity. The body of the report addresses the homogeneity measurement techniques employed, the numerical correction required to account for a difference in fuel types, and the statistical analysis of the resulting data. This statistical analysis pertains to local variation in fuel loading, as well as to ''hot segment'' analysis of narrow axial regions along the plate and ''hot streak'' analysis, the cumulative effect of hot segment loading variation. The data for all twelve plates were compiled and divided into 20 regions for analysis, with each region represented by a mean and a standard deviation to report percent deviation from nominal fuel loading. The central regions of the plates showed mean values of about +3% deviation, while the edge regions showed mean values of about -7% deviation. The data within these regions roughly approximated random samplings from normal distributions, although the chi-square (χ 2 ) test for goodness of fit to normal distributions was not satisfied

Irradiation of novel MTR fuel plates in BR2

International Nuclear Information System (INIS)

Verboomen, B.; Aoust, Th.; Beeckmans De Westmeerbeeck, A.; De Raedt, Ch.

2000-01-01

Since the end of 1999, novel MTR fuel plates with very high-density meat are being irradiated in BR2. The purpose of the irradiation is to investigate the behaviour of these fuel plates under very severe reactor operation conditions. The novel fuel plates are inserted in two standard six-tube BR2 fuel elements in the locations normally occupied by the standard outer fuel plates. The irradiation in BR2 was prepared by carrying out detailed neutron Monte Carlo calculations of the whole BR2 core containing the two experimental fuel elements for various positions in the reactor and for various azimuthal orientations of the fuel elements. Comparing the thus determined fission density levels and azimuthal profiles in the new MTR fuel plates irradiated in the various channels allowed the experimenters to choose the most appropriate BR2 channel and the most appropriate fuel element orientation. (author)

Flow-induced plastic collapse of stacked fuel plates

Energy Technology Data Exchange (ETDEWEB)

Davis, D C; Scarton, H A

1985-03-01

Flow-induced plastic collapse of stacked fuel plate assemblies was first noted in experimental reactors such as the ORNL High Flux Reactor Assembly and the Engineering Test Reactor (ETR). The ETR assembly is a stack of 19 thin flat rectangular fuel plates separated by narrow channels through which a coolant flows to remove the heat generated by fission of the fuel within the plates. The uranium alloyed plates have been noted to buckle laterally and plastically collapse at the system design coolant flow rate of 10.7 m/s, thus restricting the coolant flow through adjacent channels. A methodology and criterion are developed for predicting the plastic collapse of ETR fuel plates. The criterion is compared to some experimental results and the Miller critical velocity theory.

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

Finite element analysis of advanced neutron source fuel plates

International Nuclear Information System (INIS)

Luttrell, C.R.

1995-08-01

The proposed design for the Advanced Neutron Source reactor core consists of closely spaced involute fuel plates. Coolant flows between the plates at high velocities. It is vital that adjacent plates do not come in contact and that the coolant channels between the plates remain open. Several scenarios that could result in problems with the fuel plates are studied. Finite element analyses are performed on fuel plates under pressure from the coolant flowing between the plates at a high velocity, under pressure because of a partial flow blockage in one of the channels, and with different temperature profiles

Nuclear and radiological safety in the substitution process of the fuel HEU to LEU 30/20 in the Reactor TRIGA Mark III of the ININ

International Nuclear Information System (INIS)

Hernandez G, J.

2012-10-01

Inside the safety initiative in the international ambit, with the purpose of reducing the risks associated with the use of high enrichment nuclear fuels (HEU) for different proposes to the peaceful uses of the nuclear energy, Mexico contributes by means of the substitution of the high enrichment fuel HEU for low enrichment fuel LEU 30/20 in the TRIGA Mark III Reactor, belonging to Instituto Nacional de Investigaciones Nucleares (ININ). The conversion process was carried out by means of the following activities: analysis of the proposed core, reception and inspection of the fuel LEU 30/20, the discharge of the fuels of the mixed reactor core, shipment of the fuels HEU fresh and irradiated to the origin country, reload activities with the fuels LEU 30/20 and parameters measurement of the core operation. In order to maintaining the personnel's integrity and infrastructure associated to the Reactor, during the whole process the measurements of nuclear and radiological safety were controlled to detail, in execution with the license requirements of the installation. This work describes the covering activities and radiological inspections more relevant, as well as the measurements of radiological control implemented with base in the estimate of the equivalent dose of the substitution process. (Author)

Modeling Thermal and Stress Behavior of the Fuel-clad Interface in Monolithic Fuel Mini-plates

International Nuclear Information System (INIS)

Miller, Gregory K.; Medvedev, Pavel G.; Burkes, Douglas E.; Wachs, Daniel M.

2010-01-01

As part of the Global Threat Reduction Initiative, a fuel development and qualification program is in process with the objective of qualifying very high density low enriched uranium fuel that will enable the conversion of high performance research reactors with operational requirements beyond those supported with currently available low enriched uranium fuels. The high density of the fuel is achieved by replacing the fuel meat with a single monolithic low enriched uranium-molybdenum fuel foil. Doing so creates differences in the mechanical and structural characteristics of the fuel plate because of the planar interface created by the fuel foil and cladding. Furthermore, the monolithic fuel meat will dominate the structural properties of the fuel plate rather than the aluminum matrix, which is characteristic of dispersion fuel types. Understanding the integrity and behavior of the fuel-clad interface during irradiation is of great importance for qualification of the new fuel, but can be somewhat challenging to determine with a single technique. Efforts aimed at addressing this problem are underway within the fuel development and qualification program, comprised of modeling, as-fabricated plate characterization, and post-irradiation examination. An initial finite element analysis model has been developed to investigate worst-case scenarios for the basic monolithic fuel plate structure, using typical mini-plate irradiation conditions in the Advanced Test Reactor. Initial analysis shows that the stress normal to the fuel-clad interface dominates during irradiation, and that the presence of small, rounded delaminations at the interface is not of great concern. However, larger and/or fuel-clad delaminations with sharp corners can create areas of concern, as maximum principal cladding stress, strain, displacement, and peak fuel temperature are all significantly increased. Furthermore, stresses resulting from temperature gradients that cause the plate to bow or buckle in

Followup calculations for the UVAR LEU conversion

International Nuclear Information System (INIS)

Rydin, R.; Hosticka, B.; Burns, T; Hubbard, T.; Mulder, R

2004-01-01

The UVAR reactor was successfully converted to LEU fuel in April 1994. Void coefficient measurements were made on the 4- by-4 fully-graphite-reflected LEU-1 core configuration, and an isothermal temperature coefficient measurement was made on the operational 4-by-5 partially-graphite-reflected LEU-2 core configuration. Both of these experiments have now been modeled in their critical configurations using the 3DBUM code. The LEU cores were also modeled using the Monte Carlo code MCNP in order to obtain a neutron/gamma source for BNCT filter design calculations. Advanced BNCT filters have been evaluated using both MCNP and the discrete ordinates code DORT. The results indicate that the UVAR would be an ideal source for the BNCT treatment of brain tumors. (author)

Followup calculations for the UVAR LEU conversion

International Nuclear Information System (INIS)

Rydin, R.A.; Hosticka, B.; Burns, T.

1995-01-01

The UVAR reactor was successfully converted to LEU fuel in April 1994. Void coefficient measurements were made on the 4-by-4 fully-graphite-reflected LEU-1 core configuration, and an isothermal temperature coefficient measurement was made on the operational 4-by-5 partially-graphite-reflected LEU-2 core configuration. Both of these experiments have now been modeled in their critical configurations using the 3DBUM code. The LEU cores were also modeled using the Monte Carlo code MCNP in order to obtain a neutron/gamma source for BNCT filter design calculations. Advanced BNCT filters have been evaluated using both MCNP and the discrete ordinates code DORT. The results indicate that the UVAR would be an ideal source for the BNCT treatment of brain tumors

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)

MTR fuel plate qualification in OSIRIS reactor

International Nuclear Information System (INIS)

Sacristan, P.; Boulcourt, P.; Naury, S.; Marchard, L.; Carcreff, H.; Noirot, J.

2005-01-01

Qualification of new MTR fuel needs the irradiation in research reactors under representative neutronic, heat flux and thermohydraulic conditions. The experiments are performed in France in the OSIRIS reactor by irradiating MTR full size fuel plates in the IRIS device located in the reactor core. The fuel plates are easily removed from the device during the shutdown of the reactor for performing thickness measurements along the plates by means of a swelling measurement device. Beside the calculation capabilities, the experimental platform includes: the ISIS neutron mock-up for the measurement of neutron flux distribution along the plates; the γ spectrometry for the purpose of measuring the activities of the radionuclides representative of the power and the burnup and to compare with the neutronic calculation. Owing to the experience feedback, a good agreement is observed between calculation and measurement; destructive post irradiation examinations in the LECA facility (Cadarache). New irradiations with the IRIS device and at higher heat flux are under preparation for qualification of MTR fuels. (author)

Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

Directory of Open Access Journals (Sweden)

Aldawahra Saadou

2015-06-01

Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

Fuel cell end plate structure

Science.gov (United States)

Guthrie, Robin J.; Katz, Murray; Schroll, Craig R.

1991-04-23

The end plates (16) of a fuel cell stack (12) are formed of a thin membrane. Pressure plates (20) exert compressive load through insulation layers (22, 26) to the membrane. Electrical contact between the end plates (16) and electrodes (50, 58) is maintained without deleterious making and breaking of electrical contacts during thermal transients. The thin end plate (16) under compressive load will not distort with a temperature difference across its thickness. Pressure plate (20) experiences a low thermal transient because it is insulated from the cell. The impact on the end plate of any slight deflection created in the pressure plate by temperature difference is minimized by the resilient pressure pad, in the form of insulation, therebetween.

Swelling of U-7Mo/Al-Si dispersion fuel plates under irradiation – Non-destructive analysis of the AFIP-1 fuel plates

Energy Technology Data Exchange (ETDEWEB)

Wachs, D.M., E-mail: daniel.wachs@inl.gov [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Robinson, A.B.; Rice, F.J. [Idaho National Laboratory, Characterization and Advanced PIE Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Kraft, N.C.; Taylor, S.C. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Lillo, M. [Idaho National Laboratory, Nuclear Systems Design and Analysis Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Woolstenhulme, N.; Roth, G.A. [Idaho National Laboratory, Nuclear Fuels and Materials Division, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-08-01

Extensive fuel-matrix interactions leading to plate pillowing have proven to be a significant impediment to the development of a suitable high density low-enriched uranium molybdenum alloy (U-Mo) based dispersion fuel for high power applications in research reactors. The addition of silicon to the aluminum matrix was previously demonstrated to reduce interaction layer growth in mini-plate experiments. The AFIP-1 project involved the irradiation, in-canal examination, and post-irradiation examination of two fuel plates. The irradiation of two distinct full size, flat fuel plates (one using an Al-2wt%Si matrix and the other an Al-4043 (∼4.8 wt% Si) matrix) was performed in the INL ATR reactor in 2008–2009. The irradiation conditions were: ∼250 W/cm{sup 2} peak Beginning Of Life (BOL) power, with a ∼3.5e21 f/cm{sup 3} peak burnup. The plates were successfully irradiated and did not show any pillowing at the end of the irradiation. This paper reports the results and interpretation of the in-canal and post-irradiation non-destructive examinations that were performed on these fuel plates. It further compares additional PIE results obtained on fuel plates irradiated in contemporary campaigns in order to allow a complete comparison with all results obtained under similar conditions. Except for a brief indication of accelerated swelling early in the irradiation of the Al-2Si plate, the fuel swelling is shown to evolve linearly with the fission density through the maximum burnup.

Preliminary investigation of the use of monolithic U-Mo fuel in the MIT reactor

International Nuclear Information System (INIS)

Newton, Thomas H. Jr.; Kazimi, Mujid S.; Pilat, Edward E.; Xu Zhiwen

2003-01-01

Studies have begun on the use of monolithic LEU U-Mo fuel in the MIT Nuclear Research Reactor (MITR-II) using the Monte Carlo Transport code MCNP. These studies have included model benchmarking, LEU fuel optimization, burnup evaluation, in-core facility design, and determination of safety attributes. Benchmarking studies on the initial core have shown favorable agreement between the calculated and measured reactivity worths of the six control blades. In addition, optimization studies on LEU U7Mo MITR-II fuel have shown that an arrangement of ten to twelve plates per fuel element would have initial reactivity values and thermal neutron fluxes comparable to the current HEU core. Burnup studies which have been made using the MCODE depletion program will be presented. Safety attributes such as temperature coefficients, shutdown margins, and coolant subcooled margin are under evaluation. (author)

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)

Bipolar plates for PEM fuel cells

Science.gov (United States)

Middelman, E.; Kout, W.; Vogelaar, B.; Lenssen, J.; de Waal, E.

The bipolar plates are in weight and volume the major part of the PEM fuel cell stack, and are also a significant contributor to the stack costs. The bipolar plate is therefore a key component if power density has to increase and costs must come down. Three cell plate technologies are expected to reach targeted cost price levels, all having specific advantages and drawbacks. NedStack has developed a conductive composite materials and a production process for fuel cell plates (bipolar and mono-polar). The material has a high electric and thermal conductivity, and can be processed into bipolar plates by a proprietary molding process. Process cycle time has been reduced to less than 10 s, making the material and process suitable for economical mass production. Other development work to increase material efficiency resulted in thin bipolar plates with integrated cooling channels, and integrated seals, and in two-component bipolar plates. Total thickness of the bipolar plates is now less than 3 mm, and will be reduced to 2 mm in the near future. With these thin integrated plates it is possible to increase power density up to 2 kW/l and 2 kW/kg, while at the same time reducing cost by integrating other functions and less material use.

Drying studies of simulated DOE aluminum plate fuels

International Nuclear Information System (INIS)

Lords, R.E.; Windes, W.E.; Crepeau, J.C.; Sidwell, R.W.

1996-01-01

Experiments have been conducted to validate the Idaho National Engineering Laboratory (INEL) drying procedures for preparation of corroded aluminum plate fuel for dry storage in an existing vented (and filtered) fuel storage facility. A mixture of hydrated aluminum oxide bound with a clay was used to model the aluminum corrosion product and sediment expected in these Department of Energy (DOE) owned fuel types. Previous studies demonstrated that the current drying procedures are adequate for removal of free water inside the storage canister and for transfer of this fuel to a vented dry storage facility. However, using these same drying procedures, the simulated corrosion product was found to be difficult to dry completely from between the aluminum clad plates of the fuel. Another related set of experiments was designed to ensure that the fuel would not be damaged during the drying process. Aluminum plate fuels are susceptible to pitting damage on the cladding that can result in a portion of UAl x fuel meat being disgorged. This would leave a water-filled void beneath the pit in the cladding. The question was whether bursting would occur when water in the void flashes to steam, causing separation of the cladding from the fuel, and/or possible rupture. Aluminum coupons were fabricated to model damaged fuel plates. These coupons do not rupture or sustain any visible damage during credible drying scenarios

Dual fuel gradients in uranium silicide plates

Energy Technology Data Exchange (ETDEWEB)

Pace, B.W. [Babock and Wilcox, Lynchburg, VA (United States)

1997-08-01

Babcock & Wilcox has been able to achieve dual gradient plates with good repeatability in small lots of U{sub 3}Si{sub 2} plates. Improvements in homogeneity and other processing parameters and techniques have allowed the development of contoured fuel within the cladding. The most difficult obstacles to overcome have been the ability to evaluate the bidirectional fuel loadings in comparison to the perfect loading model and the different methods of instilling the gradients in the early compact stage. The overriding conclusion is that to control the contour of the fuel, a known relationship between the compact, the frames and final core gradient must exist. Therefore, further development in the creation and control of dual gradients in fuel plates will involve arriving at a plausible gradient requirement and building the correct model between the compact configuration and the final contoured loading requirements.

A neutronic feasibility study for LEU conversion of the WWR-M reactor at Gatchina

International Nuclear Information System (INIS)

Petrov, Yu. V.; Erykalov, A.N.; Onegin, M.S.

2000-01-01

In this report we present the results of computations of the full scale reactor core with HEU (90%), MEU (36%) and LEU (19.75%) fuel. The reactor computer model for the MCU RFFI Monte Carlo code includes all peculiarities of the core. Calculations show that a uranium density of 3.3gU/cm 3 of MEU (36%) fuel and 8/25gU/cm 3 of LEU (19.75%) in WWR-M5 fuel assembly (FA) geometry is required to match the fuel cycle length of the HEU (90%) case with the same end of cycle (EOEC) excess reactivity. For the equilibrium fuel cycle the fuel burnup and poisoning, the fast and thermal neutron fluxes, the reactivity worth of control rods were calculated for the reference case with HEU (90%) FA and for the MEU and LEU FA. The relative accuracy of this neutronic feasibility study of fuel enrichment reduction of the WWR-M reactor in Gatchina is sufficient to start the fabrication feasibility study of MEU (36%) WWR-M5 fuel assemblies. At the present stage of technology it seems hardly possible to manufacture LEU (19.75%) fuel elements in WWR-M5 geometry due to too high uranium density. Only a future R and D can solve the problem. (author)

Accelerating the design and testing of LEU fuel assemblies for conversion of Russian-designed research reactors outside Russia

International Nuclear Information System (INIS)

Matos, J.E

2003-01-01

This paper identifies proposed geometries and loading specifications of LEU tube-type and pin-type test assemblies that would be suitable for accelerating the conversion of Russian-designed research reactors outside of Russia if these fuels are manufactured, qualified by irradiation testing, and made commercially available in Russia. (author)

A neutronic feasibility study for LEU conversion of the IR-8 research reactor

International Nuclear Information System (INIS)

Deen, J.R.; Hanan, N.A.; Matos, J.E.; Egorenkov, P.M.; Nasonov, V.A.

1998-01-01

Equilibrium fuel cycle comparisons for the IR-8 research reactor were made for HEU (90%), HEU (36%), and LEU (19.75%) fuel assembly (FA) designs using three dimensional multi-group diffusion theory models benchmarked to detailed Monte Carlo models of the reactor. Comparisons were made of changes in reactivity, cycle length, average 235 U discharge burnup, thermal neutron flux, and control rod worths for the 90% and 36% enriched IRT-3M fuel assembly and the 19.75% enriched IRT-4M fuel assembly with the same fuel management strategy. The results of these comparisons showed that a uranium density of 3.5 g/cm 3 in the fuel meat would be required in the LEU IRT-4M fuel assembly to match the cycle length of the HEU (90%) IRT-3M FA and an LEU density of 3.7 g/cm 3 is needed to match the cycle length of the HEU (36%) IRT-3M FA. (author)

Fabrication and irradiation testing of LEU [low enriched uranium] fuels at CRNL status as of 1987 September

International Nuclear Information System (INIS)

Sears, D.F.; Berthiaume, L.C.; Herbert, L.N.

1987-01-01

The current status of Chalk River Nuclear Laboratories' (CRNL) program to develop and test low-enriched uranium (LEU), proliferation-resistant fuels for use in research reactors is reviewed. CRNL's fuel manufacturing process has been qualified by the successful demonstration irradiation of 7 full-size rods in the NRU reactor. Now industrial-scale production equipment has been commissioned, and a fuel-fabrication campaign for 30 NRU rods and a MAPLE-X core is underway. Excess capacity could be used for commercial fuel fabrication. In the irradiation testing program, mini-elements with deliberately included core surface defects performed well in-reactor, swelling by only 7 to 8 vol% at 93 atomic percent burnup of the original U-235. The additional restraint provided by the aluminium cladding which flowed into the defects during extrusion contributed to this good performance. Mini-elements containing a variety of particle size distributions were also successfully irradiated to 93 at% burnup in NRU, as part of a study to establish the optimum particle size distribution. Swelling was found to be proportional to the percentage of fines (<44μm particles) contained in the cores. The mini-elements containing the composition normally used at CRNL had swollen by 5.8 vol%, and mini-elements with a much higher percentage of fines had swollen by 6.8 vol%, at 93 at% burnup. Also, a program to develop LEU targets for Mo-99 production, via the technology developed to fabricate dispersed silicide fuel, has started, and preliminary scoping studies are underway. (Author)

Performance and economic penalties of some LEU [low enriched uranium] conversion options for the Australian Reactor HIFAR

International Nuclear Information System (INIS)

McCulloch, D.B.; Robinson, G.S.

1987-01-01

Performance calculations for the conversion of HIFAR to low enriched uranium (LEU) fuel have been extended to a wide range of 235 U loadings per fuel element. Using a simple approximate algorithm for the likely costs of LEU compared with highly enriched uranium (HEU) fuel elements, the increases in annual fuelling costs for LEU compared with HEU fuel are examined for a range of conversion options involving different performance penalties. No significant operational/safety problems were found for any of the options canvassed. (Author)

Neutronic analysis of HEU to LEU conversion calculation for AEOI 5 MW pool-type MTR fuel research reactor core

International Nuclear Information System (INIS)

Pazirandeh, A.; Lutz, D.; Bartsch, G.

1987-07-01

The possibility of converting HEU(93%) fuel to LEU(20%) fuel without or with slight alteration to the fuel element geometry is discussed. The fuel density varies between 1.7 to 4.1 g U-235/cm. In cross section generation a unit cell with an extra zone to account for extra Al and water was considered. In burnup calculations a sequential shuffling pattern was assumed with fixed position control fuel elements. A cross section data set in 45 energy groups were generated using RSYST/CGM system using the cross section library JFET. Then for 2D-diffusion calculations homogenized and condensed 5 energy group cross sections were prepared. (orig./HP)

Fuel performance analysis for the HAMP-1 mini plate test

Energy Technology Data Exchange (ETDEWEB)

Cho, Byoung Jin; Tahka, Y. W.; Yim, J. S.; Lee, B. H. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

U-7wt%Mo/Al- 5wt%Si dispersion fuel with 8gU/cm{sup 3} is chosen to achieve more efficiency and higher performance than the conventional U{sub 3}Si{sub 2} fuel. As part of the fuel qualification program for the KiJang research reactor (KJRR), three irradiation tests with mini-plates are on the way at the High-flux Advanced Neutron Application Reactor (HANARO). The first test among three HANARO Mini-Plate Irradiation tests (HAMP-1, 2, 3) has completed. PLATE code has been initially developed to analyze the thermal performance of high density U-Mo/Al dispersion fuel plates during irradiation [1]. We upgraded the PLATE code with the latest irradiation results which were implemented by corrosion, thermal conductivity and swelling model. Fuel performance analysis for HAMP-1 was conducted with updated PLATE. This paper presents results of performance evaluation of the HAMP-1. Maximum fuel temperature was obtained 136 .deg., which is far below the preset limit of 200 .deg. for the irradiation test. The meat swelling and corrosion thickness was also confirmed that the developed fuel would behave as anticipated.

Thermal hydraulic analysis of the JMTR improved LEU-core

Energy Technology Data Exchange (ETDEWEB)

Tabata, Toshio; Nagao, Yoshiharu; Komukai, Bunsaku; Naka, Michihiro; Fujiki, Kazuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Takeda, Takashi [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokai, Ibaraki (Japan)

2003-01-01

After the investigation of the new core arrangement for the JMTR reactor in order to enhance the fuel burn-up and consequently extend the operation period, the ''improved LEU core'' that utilized 2 additional fuel elements instead of formerly installed reflector elements, was adopted. This report describes the results of the thermal-hydraulic analysis of the improved LEU core as a part of safety analysis for the licensing. The analysis covers steady state, abnormal operational transients and accidents, which were described in the annexes of the licensing documents as design bases events. Calculation conditions for the computer codes were conservatively determined based on the neutronic analysis results and others. The results of the analysis, that revealed the safety criteria were satisfied on the fuel temperature, DNBR and primary coolant temperature, were used in the licensing. The operation license of the JMTR with the improved LEU core was granted in March 2001, and the reactor operation with new core started in November 2001 as 142nd operation cycle. (author)

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

Energy Technology Data Exchange (ETDEWEB)

Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showing agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm² and temporary heat flux limit of 600 W/cm². This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.

Creep analysis of fuel plates for the Advanced Neutron Source

International Nuclear Information System (INIS)

Swinson, W.F.; Yahr, G.T.

1994-11-01

The reactor for the planned Advanced Neutron Source will use closely spaced arrays of fuel plates. The plates are thin and will have a core containing enriched uranium silicide fuel clad in aluminum. The heat load caused by the nuclear reactions within the fuel plates will be removed by flowing high-velocity heavy water through narrow channels between the plates. However, the plates will still be at elevated temperatures while in service, and the potential for excessive plate deformation because of creep must be considered. An analysis to include creep for deformation and stresses because of temperature over a given time span has been performed and is reported herein

Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Wehe, D K [Oak Ridge National Laboratory, Oak Ridge, TN (United States); King, J S; Lee, J C; Martin, W R [Department of Nuclear Engineering, University of Michigan, Ann Arbor, MI (United States)

1985-07-01

The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from meV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional {sup 238}U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above MeV. (author)

Development of hold down plate of INGLE fuel assembly

International Nuclear Information System (INIS)

Kim, Hyeong Koo; Kim, Kyu Tae

1996-07-01

Hold down plate for the INGLE fuel which has been designed for high performance in the standpoints of thermal margin and structural integrity compared to current fuel for YGN 3/4 and UCN 3/4 has been developed and its structural integrity has been verified based on the eh stress analysis. The design feature of the developed hold down plate has not only perfect compatibility with the reactor internals of Korea standard reactor, but also brand-new locking mechanism between upper tie plate and guide tubes. This locking mechanism introduced to the INGLE fuel provides very simple and reliable reconstitutability. In this report, finite element stress analysis with the aid of the ANSYS code as a solver and the MSC/PATRAN code as a pre and post processor were performed to verify structural integrity of the hold down plate considering various load cases which seem to be applied to the hold down plate during its lifetime. Based on the analysis results, the developed hold down plate for INGLE fuel sustains structural integrity under considered load conditions. 3 tabs., 16 figs., 9 refs. (Author)

Evaluation of plate type fuel elements by eddy current test method

International Nuclear Information System (INIS)

Frade, Rangel Teixeira

2015-01-01

Plate type fuel elements are used in MTR research nuclear reactors. The fuel plates are manufactured by assembling a briquette containing the fissile material inserted in a frame, with metal plates in both sides of the set, to act as a cladding. This set is rolled under controlled conditions in order to obtain the fuel plate. In Brazil, this type of fuel is manufactured by IPEN and used in the IEA-R1 reactor. After fabrication of three batches of fuel plates, 24 plates, one of them is taken, in order to verify the thickness of the cladding. For this purpose, the plate is sectioned and the thickness measurements are carried out by using optical microscopy. This procedure implies in damage of the plate, with the consequent cost. Besides, the process of sample preparation for optical microscopy analysis is time consuming, it is necessary an infrastructure for handling radioactive materials and there is a generation of radioactive residues during the process. The objective of this study was verify the applicability of eddy current test method for nondestructive measurement of cladding thickness in plate type nuclear fuels, enabling the inspection of all manufactured fuel plates. For this purpose, reference standards, representative of the cladding of the fuel plates, were manufactured using thermomechanical processing conditions similar to those used for plates manufacturing. Due to no availability of fuel plates for performing the experiments, the presence of the plate’s core was simulated using materials with different electrical conductivities, fixed to the thickness reference standards. Probes of eddy current testing were designed and manufactured. They showed high sensitivity to thickness variations, being able to separate small thickness changes. The sensitivity was higher in tests performed on the reference standards and samples without the presence of the materials simulating the core. For examination of the cladding with influence of materials simulating the

Design of the Flow Plates for a Dual Cooled Fuel Assembly

International Nuclear Information System (INIS)

Kim, Jae Yong; Yoon, Kyung Ho; Lee, Young Ho; Lee, Kang Hee; Kim, Hyung Kyu

2009-01-01

In a dual cooled fuel assembly, the array and position of fuels are changed from those of a conventional PWR fuel assembly to achieve a power uprating. The flow plate provides flow holes to direct the heated coolant into/out of the fuel assembly and structural intensity to insure that the fuel rod is axially restrained within the spacer grids. So, flow plates of top/bottom end pieces (TEP/BEP) have to be modified into proper shape. Because the flow holes' area of a flow plate affects pressure drop, the flow holes' area must be larger than/equal to that of conventional flow plates. And design criterion of the TEP/BEP says that the flow plate should withstand a 22.241 kN axial load during handling lest a calculated stress intensity should exceed the Condition I allowable stress. In this paper, newly designed flow plates of a TEP/BEP are suggested and stress analysis is conducted to evaluate strength robustness of the flow plates for the dual cooled fuel assembly

CarbonNanoTubes (CNT) in bipolar plates for PEM fuel cell applications

Energy Technology Data Exchange (ETDEWEB)

Grundler, M.; Derieth, T.; Beckhaus, P.; Heinzel, A. [centre for fuel cell technology ZBT GmbH (Germany)

2010-07-01

Using standard mass production techniques for the fabrication of fuel cell components, such as bipolar plates, is a main issue for the commercialisation of PEM fuel cell systems. Bipolar plates contribute significantly to the cost structure of PEM stacks. In an upcoming fuel cell market a large number of bipolar plates with specific high-quality standards will be needed. At the Centre for Fuel Cell Technology (ZBT) together with the University of Duisburg-Essen fuel cell stacks based on injection moulded bipolar plates have been developed and demonstrated successfully [1]. This paper focuses on the interactions between carbon filling materials (graphite, carbon black and carbon nanotubes (CNT)) in compound based bipolar plates and especially the potential of CNTs, which were used in bipolar plates for the first time. The entire value added chain based on the feedstock, the compounding and injection moulding process, the component bipolar plate, up to the operation of a PEM single fuel cell stack with CNT-based bipolar plates is disclosed. (orig.)

Evaluation of plate type fuel options for small power reactors

International Nuclear Information System (INIS)

Andrzejewski, Claudio de Sa

2005-01-01

Plate type fuels are generally used in research reactor. The utilization of this kind of configuration improves significantly the overall performance fuel. The conception of new fuels for small power reactors based in plate-type configuration needs a complete review of the safety criteria originally used to conduce power and research reactor projects. In this work, a group of safety criteria is established for the utilization of plate-type fuels in small power reactors taking into consideration the characteristics of power and research reactors. The performance characteristics of fuel elements are strongly supported by its materials properties and the adopted configuration for its fissile particles. The present work makes an orientated bibliographic investigation searching the best material properties (structural materials and fuel compounds) related to the performance fuel. Looking for good parafermionic characteristics and manufacturing exequibility associated to existing facilities in national research centres, this work proposes several alternatives of plate type fuels, considering its utilization in small power reactors: dispersions of UO 2 in stainless steel, of UO 2 in zircaloy, and of U-Mo alloy in zircaloy, and monolithic plates of U-Mo cladded with zircaloy. Given the strong dependency of radiation damage with temperature increase, the safety criteria related to heat transfer were verified for all the alternatives, namely the DNBR; coolant temperature lower than saturation temperature; peak meat temperature to avoid swelling; peak fuel temperature to avoid meat-matrix reaction. It was found that all alternatives meet the safety criteria including the 0.5 mm monolithic U-Mo plate cladded with zircaloy. (author)

Postirradiation analysis of experimental uranium-silicide dispersion fuel plates

International Nuclear Information System (INIS)

Hofman, G.L.; Neimark, L.A.

1985-01-01

Low-enriched uranium silicide dispersion fuel plates were irradiated to maximum burnups of 96% of 235 U. Fuel plates containing 33 v/o U 3 Si and U 3 Si 2 behaved very well up to this burnup. Plates containing 33 v/o U 3 Si-Al pillowed between 90 and 96% burnup of the fissile atoms. More highly loaded U 3 Si-Al plates, up to 50 v/o were found to pillow at lower burnups. Plates containing 40 v/o U 3 Si showed an increase swelling rate around 85% burnup. 5 refs., 10 figs

PEM fuel cell bipolar plate material requirements for transportation applications

Energy Technology Data Exchange (ETDEWEB)

Borup, R.L.; Stroh, K.R.; Vanderborgh, N.E. [Los Alamos National Lab., NM (United States)] [and others

1996-04-01

Cost effective bipolar plates are currently under development to help make proton exchange membrane (PEM) fuel cells commercially viable. Bipolar plates separate individual cells of the fuel cell stack, and thus must supply strength, be electrically conductive, provide for thermal control of the fuel stack, be a non-porous materials separating hydrogen and oxygen feed streams, be corrosion resistant, provide gas distribution for the feed streams and meet fuel stack cost targets. Candidate materials include conductive polymers and metal plates with corrosion resistant coatings. Possible metals include aluminium, titanium, iron/stainless steel and nickel.

Calculation of mixed HEU-LEU cores for the HOR research reactor with the scale code system

International Nuclear Information System (INIS)

Leege, P.F.A. de; Gibcus, H.P.M.; Hoogenboom, J.E.; Vries, J.W. de

1997-01-01

The HOR reactor of Interfaculty Reactor Institute (IRI), Delft, The Netherlands, will be converted to use low enriched fuel (LEU) assemblies. As there are still many usable high enriched (HEU) fuel assemblies present, there will be a considerable reactor operation time with mixed cores with both HEU and LEU fuel assemblies. At IRI a comprehensive reactor physics code system and evaluated nuclear data is implemented for detailed core calculations. One of the backbones of the IRI code system is the well-known SCALE code system package. Full core calculations are performed with the diffusion theory code BOLD VENTURE, the nodal code SILWER, and the Monte Carlo code KENO Va. Results are displayed of a strategy from a HEU core to a mixed HEU-LEU core and eventually a LEU core. (author)

SEM Characterization of an Irradiated Monolithic U-10Mo Fuel Plate

International Nuclear Information System (INIS)

Keiser, D.D. Jr.; Jue, J.F.; Robinson, A.B.

2010-01-01

Results of scanning electron microscopy (SEM) characterization of irradiated U-7Mo dispersion fuel plates with differing amounts of matrix Si have been reported. However, to date, no results of SEM analysis of irradiated U-Mo monolithic fuel plates have been reported. This paper describes the first SEM characterization results for an irradiated monolithic U-10Mo fuel plate. Two samples from this fuel plate were characterized. One sample was produced from the low-flux side of the fuel plate, and another was produced at the high-flux side of the fuel plate. This characterization focused on the microstructural features present at the U-10Mo foil/cladding interface, particularly the interaction zone that had developed during fabrication and irradiation. In addition, the microstructure of the foil itself was investigated, along with the morphology of the observed fission gas bubbles. It was observed that a Si-rich interaction layer was present at the U-10Mo foil/cladding interface that exhibited relatively good irradiation behavior, and within the U-10Mo foil the microstructural features differed in some respects from what is typically seen in the U-Mo powders of an irradiated dispersion fuel.

Criticality Calculations of Fresh LEU and MOX Assemblies for Transport and Storage at the Balakovo Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Goluoglu, S.

2001-01-11

Transportation of low-enriched uranium (LEU) and mixed-oxide (MOX) assemblies to and within the VVER-1000-type Balakovo Nuclear Power Plant is investigated. Effective multiplication factors for fresh fuel assemblies on the railroad platform, fresh fuel assemblies in the fuel transportation vehicle, and fresh fuel assemblies in the spent fuel storage pool are calculated. If there is no absorber between the units, the configurations with all MOX assemblies result in higher effective multiplication factors than the configurations with all LEU assemblies when the system is dry. When the system is flooded, the configurations with all LEU assemblies result in higher effective multiplication factors. For normal operating conditions, effective multiplication factors for all configurations are below the presumed upper subcritical limit of 0.95. For an accident condition of a fully loaded fuel transportation vehicle that is flooded with low-density water (possibly from a fire suppression system), the presumed upper subcritical limit is exceeded by configurations containing LEU assemblies.

Parametric study of the deformation of dispersion fuel plates

International Nuclear Information System (INIS)

Vieira, Edeval; Leal Neto, Ricardo Mendes; Durazzo, Michelangelo

2011-01-01

The Nuclear and Energy Research Institute - IPEN-CNEN/SP produces routinely the nuclear fuel necessary for operating its research reactor, IEA-R1. This fuel consists of fuel plates containing U 3 Si 2 -Al composites as the meat, which are fabricated by rolling. The rolling process currently deployed was developed with base on information obtained from literature, which were used as premises for defining the current manufacturing procedures, according to a methodology with essentially empirical character. Despite the current rolling process to be perfectly stable and highly reproducible, it is not well characterized and therefore is not fully known. The objective of this work is to characterize the rolling process for producing fuel plates, presenting results of the evolution of all parameters of technological interest, after each rolling pass, obtaining information along the fuel plate deformation during the rolling process. (author)

Analysis of the critical and first full power operating cores for PARR using leu oxide fuel

International Nuclear Information System (INIS)

Khan, L.A.; Qazi, M.K.; Bokhari, I.H.; Fazal, R.

1989-10-01

This paper explains the analysis for determining the first full power operating core for PARR using LEU oxide fuel. The core configuration selected for this first full power operation contains about 6.13 kg of U-235 distributed in 19 standard and five control fuel elements. The neutron flux level is doubled when core is shifted from 5MW to 10 MW. Total nuclear power peaking factor of the core is 2.03. The analysis shows that the core can be operated safely at 5 MW with a flow rate of 520 meter cube per hour and at 10 MW with a flow rate of 900 meter cube per hour. (A.B.). 10 figs

Quality control of nuclear fuel plates using digital image processing techniques

International Nuclear Information System (INIS)

Salinas, Renato; Radd, Ulrich; Coronado, Harold; Olivares, Luis

2003-01-01

The Chilean Atomic Energy Commission (CCHEN) has developed the technology requires to manufacture low enriched uranium-235 nuclear fuel elements used in non-power reactor applications and in research. These fuel plates are assembled in two nuclear facilities located at La Reina (RECH-1) and Lo Aguirre where the present work was developed. Furthermore since high quality standards have been met, these facilities are able to export these nuclear fuel plates to foreign countries. Each MTR fuel elements consists of 16 low enriched uranium silicide (U 3 Si 2 ) fuel plates. A stringent quality assurance program requires among others, homogeneity measurements of uranium surface density values of these fuel plates, which are traditionally accomplished with optical densitometry methods. We have implemented and alternative technique which uses computer vision to determine uranium surface density values in these fuel plates. Both techniques are compared. Advantages of machine vision methods include considerable time saving and a complete quantitative evaluation of uranium densities as compared to the sparse technique involved in the optical densitometry method (Au)

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

SEM characterization of an irradiated monolithic U-10Mo fuel plate

International Nuclear Information System (INIS)

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

2010-01-01

Results of scanning electron microscopy (SEM) characterization of irradiated U-7Mo dispersion fuel plates with differing amounts of matrix Si have been reported. However, to date, no results of SEM analysis of irradiated U-Mo monolithic fuel plates have been reported. This paper describes the first SEM characterization results for an irradiated monolithic U-10Mo fuel plate. Two samples from this fuel plate were characterized. One sample was produced from the low-flux side of the fuel plate, and another was produced at the high-flux side of the fuel plate. This characterization focused on the microstructural features present at the U-10Mo foil/AA6061 cladding interface, particularly the interaction zone that had developed during fabrication and any continued development during irradiation. In addition, the microstructure of the foil itself was investigated, along with the morphology of the observed fission gas bubbles. It was observed that a Si-rich interaction layer was present at the U-10Mo foil/cladding interface that exhibited relatively good irradiation behavior, and within the U-10Mo foil the microstructural features differed in some respects from what is typically seen in the U-7Mo powders of an irradiated dispersion fuel. (author)

Operating experience, measurements, and analysis of the LEU whole core demonstration at the FNR

International Nuclear Information System (INIS)

Weha, D.K.; Drumm, C.R.; King, J.S.; Martin, W.R.; Lee, J.C.

1984-01-01

The 2-MW Ford Nuclear Reactor at the University of Michigan is serving as the demonstration reactor for the MTR-type low enrichment (LEU) fuel for the Reduced Enrichment for Research and Test Reactor program. Operational experience gained through six months of LEU core operation and seven months of mixed HEU-LEU core operation is presented. Subcadmium flux measurements performed with rhodium self-powered neutron detectors and iron wire activations are compared with calculations. Measured reactivity parameters are compared for HEU and LEU cores. Finally, the benchmark calculations for several HEU, LEU, and mixed HEU-LEU FNR cores and the International Atomic Energy Agency (IAEA) benchmark problem are presented. (author)

Quantitative determination of uranium distribution homogeneity in MTR fuel type plates

International Nuclear Information System (INIS)

Ferrufino, Felipe Bonito Jaldin

2011-01-01

IPEN/CNEN-SP produces the fuel to supply its nuclear research reactor IEA-R1. The fuel is assembled with fuel plates containing an U 3 Si 2 -Al composite meat. A good homogeneity in the uranium distribution inside the fuel plate meat is important from the standpoint of irradiation performance. Considering the lower power of reactor IEA-R1, the uranium distribution in the fuel plate has been evaluated only by visual inspection of radiographs. However, with the possibility of IPEN to manufacture the fuel for the new Brazilian Multipurpose Reactor (RMB), with higher power, it urges to develop a methodology to determine quantitatively the uranium distribution into the fuel. This paper presents a methodology based on X-ray attenuation, in order to quantify the uranium concentration distribution in the meat of the fuel plate by using optical densities in radiographs and comparison with standards. The results demonstrated the inapplicability of the method, considering the current specification for the fuel plates due to the high intrinsic error to the method. However, the study of the errors involved in the methodology, seeking to increase their accuracy and precision, can enable the application of the method to qualify the final product. (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

Irradiation testing of LEU fuels in the SILOE Reactor - Progress report

International Nuclear Information System (INIS)

Merchie, Francis; Baas, Claude; Martel, Patrick

1985-01-01

Irradiation of uranium-silicide fuels has continued in the SILOE reactor during the past year. Thickness vs. fission density data from four U 3 Si plates containing 5.5 and 6.0 g U/cm 3 have been analyzed, and the results are presented. The irradiation of a full 60 g U/cm 3 U 3 Si element has begun. In addition, four U 3 Si 2 plates containing 20 to 54 g U/cm 3 are now being irradiated. These irradiations and future plans are discussed in the paper. (author)

Fuel Thermo-physical Characterization Project. Fiscal Year 2014 Final Report

Energy Technology Data Exchange (ETDEWEB)

Burkes, Douglas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Andrew M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buck, Edgar C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Casella, Amanda J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Edwards, Matthew K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); MacFarlan, Paul J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pool, Karl N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Slonecker, Bruce D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Smith, Frances N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Steen, Franciska H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-03-15

The Office of Material Management and Minimization (M3) Reactor Conversion Fuel Thermo-Physical Characterization Project at Pacific Northwest National Laboratory (PNNL) was tasked with using PNNL facilities and processes to receive irradiated low enriched uranium–molybdenum (LEU-Mo) fuel plate samples and perform analysis in support of the M3 Reactor Conversion Program. This work is in support of the M3 Reactor Conversion Fuel Development Pillar that is managed by Idaho National Laboratory. The primary research scope was to determine the thermo-physical properties as a function of temperature and burnup. Work conducted in Fiscal Year (FY) 2014 complemented measurements performed in FY 2013 on four additional irradiated LEU-Mo fuel plate samples. Specifically, the work in FY 2014 investigated the influence of different processing methods on thermal property behavior, the absence of aluminum alloy cladding on thermal property behavior for additional model validation, and the influence of higher operating surface heat flux / more aggressive irradiation conditions on thermal property behavior. The model developed in FY 2013 and refined in FY 2014 to extract thermal properties of the U-Mo alloy from the measurements conducted on an integral fuel plate sample (i.e., U-Mo alloy with a thin Zr coating and clad in AA6061) continues to perform very well. Measurements conducted in FY 2014 on samples irradiated under similar conditions compare well to measurements performed in FY 2013. In general, there is no gross influence of fabrication method on thermal property behavior, although the difference in LEU-Mo foil microstructure does have a noticeable influence on recrystallization of grains during irradiation. Samples irradiated under more aggressive irradiation conditions, e.g., higher surface heat flux, revealed lower thermal conductivity when compared to samples irradiated at moderate surface heat fluxes, with the exception of one sample. This report documents thermal

HEU to LEU conversion experience at the UMass-Lowell research reactor

International Nuclear Information System (INIS)

White, John R.; Bobek, Leo M.

2005-01-01

The UMass-Lowell Research Reactor (UMLRR) operated safely with high-enriched uranium (HEU) fuel for over 25 years. Having reached the end of core lifetime and due to proliferation concerns, the reactor was recently converted to low-enriched uranium silicide (LEU) fuel. The actual process for converting the UMLRR from HEU to LEU fuel covered a period of over 15 years. The conversion effort - from the initial conceptual design studies in the late 1980s to the final offsite shipment of the spent HEU fuel in August 2004 - was a unique experience for the faculty and staff of a small university research reactor. This paper gives a historical view of the process and it highlights several key milestones along the road to successful completion of this project. (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)

Ni-based amorphous alloy-coating for bipolar plate of PEM fuel cell by electrochemical plating

International Nuclear Information System (INIS)

Yamaura, S; Kim, S C; Inoue, A

2013-01-01

In this study, the Ni-Cr-P amorphous alloy-coated bipolar plates were produced by electro-plating on the Cu base plates with a flow field. The power generation tests of a single fuel cell with those Ni-Cr-P bipolar plates were conducted at 353 K. It was found that the single fuel cell with those Ni-Cr-P bipolar plates showed excellent I-V performance as well as that with the carbon graphite bipolar plates. It was also found that the single cell with those Ni-Cr-P bipolar plates showed better I-V performance than that with the Ni-P amorphous alloy-coated bipolar plates. Furthermore, the long-time operation test was conducted for 440 h with those Ni-Cr-P bipolar plates at the constant current density of 200 mA·cm −2 . As a result, it was found that the cell voltage gradually decreased at the beginning of the measurement before 300 h and then the voltage was kept constant after 300 h.

Preliminary design study for a carbide LEU-nuclear thermal rocket

International Nuclear Information System (INIS)

Venneri, P.F.; Kim, Y.

2014-01-01

Nuclear space propulsion is a requirement for the successful exploration of the solar system. It offers the possibility of having both a high specific impulse and a relatively high thrust, allowing rapid transit times with a minimum usage of fuel. This paper proposes a nuclear thermal rocket design based on heritage NERVA rockets that makes use of Low Enriched Uranium (LEU) fuel. The Carbide LEU Nuclear Thermal Rocket (C-LEU-NTR) is designed to fulfill the rocket requirements as set forth in the NASA 2009 Mars Mission Design Reference Architecture 5.0, that is provide 25,000 lbf of thrust, operate at full power condition for at least two hours, and have a specific impulse close to 900 s. The neutronics analysis was done using MCNP5 with the ENDF/B-VII.1 neutron library. The thermal hydraulic calculations and size optimization were completed with a finite difference code being developed at the Center for Space Nuclear Research. (authors)

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)

Safety criterion for burnout of the plate-type fuel in pressurized conditions

International Nuclear Information System (INIS)

Komori, Y.; Kaminaga, M.; Sakurai, F.; Ando, H.; Sudo, Y.; Saito, M.; Futamura, Y.

1992-01-01

The reduced enrichment program for JMTR is now underway and the core conversion to LEU (Low Enrichment Uranium) is scheduled to be made in 1993. Consistent with the safety guide which have been recently developed for research and test reactors in Japan, the safety analysis for the JMTR LEU conversion was conducted. In the safety analysis, DNB (Departure from Nucleate Boiling) heat flux correlation for the JMTR downflow condition was reconsidered because recent studies on burnout show that DNB heat fluxes with thin rectangular channels under low flow rate and low pressure conditions are much lower than predicted values by conventional DNB correlations. Available DNB data, however, are very limited for the JMTR operation pressure range, so that DNB experiments were conducted simulating the JMTR fuel subchannel. Based mainly on the present experimental data, the DNB correlations scheme composed of three correlations was selected for the JMTR safety analysis. Errors of the correlations scheme with experimental data were evaluated in order to determine the allowable limit of the minimum DNB ratio for preventing fuel failure. (author)

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

Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

International Nuclear Information System (INIS)

Mo, S.C.

1991-01-01

The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed

Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

International Nuclear Information System (INIS)

Delafosse, Jacques.

1976-01-01

French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

Caramel, uranium oxide fuel plates for water cooled reactors

International Nuclear Information System (INIS)

Bussy, Pierre; Delafosse, Jacques; Lestiboudois, Guy; Cerles, J.-M.; Schwartz, J.-P.

1979-01-01

The fuel is composed of thin plates assembled parallel to each other to form bundles or assemblies. Each plate is composed of a pavement of uranium oxide pellets, insulated from each other by a zircaloy cladding. The 235 U enrichment does not exceed 8%. The range of uses for this fuel extends from electric power generating reactors to irradiation reactors for research work. A parametric study in test loops has made it possible to determine the operating limits of this thick fuel, without bursting. The resulting diagram gives the permissible power densities, with and without cycling for specific burn-ups beyond 50,000 MWd/t. The thinnest plates were also irradiated in total in the form of advance assemblies irradiated in the core of the OSIRIS pile prior to its transformation. This transformation and the operation of this reactor with a core of 'Caramel' elements is the main trial experiment of this fuel [fr

Making of fission 99Mo from LEU silicide(s): A radiochemists' view

International Nuclear Information System (INIS)

Kolar, Z.I.; Wolterbeek, H.Th.

2005-01-01

The present-day industrial scale production of 99 Mo is fission based and involves thermal-neutron irradiation in research reactors of highly enriched uranium (HEU, > 20 % 235 U) containing targets, followed by radiochemical processing of the irradiated targets resulting in the final product: a 99 Mo containing chemical compound of molybdenum. In 1978 a program (RERTR) was started to develop a substitute for HEU reactor fuel i.e. a low enriched uranium (LEU, 235 U) one. In the wake of that program studies were undertaken to convert HEU into LEU based 99 Mo production. Both new targets and radiochemical treatments leading to 99 Mo compounds were proposed. One of these targets is based on LEU silicide, U 3 Si 2 . Present paper aims at comparing LEU U 3 Si 2 and LEU U 3 Si with another LEU target i.e. target material and arriving at some preferences pertaining to 99 Mo production. (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

Post-pulse detail metallographic examinations of low-enriched uranium silicide plate-type miniature fuel

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki

1991-10-01

Pulse irradiation at Nuclear Safety Research Reactor (NSRR) was performed using low-enriched (19.89 w% 235 U) unirradiated silicide plate-type miniature fuel which had a density of 4.8 gU/cm 3 . Experimental aims are to understand the dimensional stability and to clarify the failure threshold of the silicide plate-type miniature fuel under power transient conditions through post-pulse detail metallographic examinations. A silicide plate-type miniature fuel was loaded into an irradiation capsule and irradiated by a single pulse. Deposited energies given in the experiments were 62, 77, 116 and 154 cal/g·fuel, which lead to corresponding peak fuel plate temperatures, 201 ± 28degC, 187 ± 10degC, 418 ± 74degC and 871 ± 74degC, respectively. Below 400degC, reliability and dimensional stability of the silicide plate fuel was sustained, and the silicide plate fuel was intact. Up to 540degC, wall-through intergranular crackings occurred in the Al-3%Mg alloy cladding. With the increase of the temperature, the melting of the aluminum cladding followed by recrystallization, the denudation of fuel core and the plate-through intergranular cracking were observed. With the increase of the temperature beyond 400degC, the bowing of fuel plate became significant. Above the temperature of 640degC molten aluminum partially reacted with the fuel core, partially flowed downward under the influence of surface tension and gravity, and partially formed agglomerations. Judging from these experimental observations, the fuel-plate above 400degC tends to reduce its dimensional stability. Despite of the apparent silicide fuel-plate failure, neither generation of pressure pulse nor that of mechanical energy occurred at all. (J.P.N.)

Supply of low enriched (LEU) and highly enriched uranium (HEU) for research reactors

International Nuclear Information System (INIS)

Mueller, H.

1997-01-01

Enriched uranium for research reactors in the form of LEU /= low enriched uranium at 19.75% U-235) and HEU (= highly enriched uranium at 90 to 93% U-235) was and is - due to its high U-235 enrichment - a political fuel other than enriched uranium for power reactors. The sufficient availability of LEU and HEU is a vital question for research reactors, especially in Europe, in order to perform their peaceful research reactor programs. In the past the USA were in the Western hemisphere sole supplier of LEU and HEU. Today the USA have de facto stopped the supply of LEU and HEU, for HEU mainly due to political reasons. This paper deals, among others, with the present availability of LEU and HEU for European research reactors and touches the following topics: - historical US supplies, - influence of the RERTR-program, - characteristics of LEU and HEU, - military HEU enters the civil market, -what is the supply situation for LEU and HEU today? - outlook for safe supplies of LEU and HEU. (author)

Mo-99 production on a LEU solution reactor

International Nuclear Information System (INIS)

Brown, R.W.; Thome, L.A.; Khvostionov, V.Y.

2005-01-01

A pilot homogenous reactor utilizing LEU has been developed by the Kurchatov Institute in Moscow along with their commercial partner TCI Medical. This solution reactor operates at levels up to 50 kilowatts and has successfully produced high quality Mo-99 and Sr-89. Radiochemical extraction of medical radionuclides from the reactor solution is performed by passing the solution across a series of inorganic sorbents. This reactor has commercial potential for medical radionuclide production using LEU UO 2 SO 4 fuel. Additional development work is needed to optimize multiple 50 kilowatt cores while at the same time, optimizing production efficiency and capital expenditure. (author)

Electromagnetic Acoustic Test of the Artificial Defects for a Plate-type Nuclear Fuel

International Nuclear Information System (INIS)

Jung, Hyun Kyu; Kim, Dong Min; Lee, Yoon Sang; Cheong, Yong Moo

2011-01-01

Most research and test reactors use the nuclear fuel plates which are consisted of a fuel meat in aluminum alloy. Last year, KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of the plate-type fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done under immersion condition, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined is a non-ferromagnetic material such as aluminum with a good acousto-elastic property, which requires an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an Electromagnetic Acoustic Transducer (EMAT) technology for an automated inspection of a nuclear fuel without water

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Analysis of hydraulic instability of ANS involute fuel plates

International Nuclear Information System (INIS)

Sartory, W.K.

1991-11-01

Curved shell equations for the involute Advanced Neutron Source (ANS) fuel plates are coupled to two-dimensional hydraulic channel flow equations that include fluid friction. A complete set of fluid and plate boundary conditions is applied at the entrance and exit and along the sides of the plate and the channel. The coupled system is linearized and solved to assess the hydraulic instability of the plates

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.

Core instrumentation and pre-operational procedures for core conversion HEU to LEU

International Nuclear Information System (INIS)

1984-02-01

This report is intended for the reactor operator, to be used as a manual or checklist for general guidance on pre-startup activities that need to be addressed in preparation for conversion to Low Enriched Fuel (LEU). All nuclear, thermodynamic and safety calculations should have been performed prior to this stage of the core conversion process. During these calculations and certainly before ordering the new LEU fuel elements the reactor operator needs to very carefully consider additional important factors concerning the new fuel: fuel reliability, reliability of fuel fabricator, reprocessing contract or fuel element storage and disposal, economics of the new fuel cycle. At this stage, too, a preoperational experimental programme has to be developed and presented to the regulatory authorities for approval. This experimental programme could lead to additional requirements on: in-core instrumentation, out-of-core instrumentation or additional experimental devices. Detailed instructions on specific tests and measurements are not provided in this report since much information on the subject is available in the open literature

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Benchmarking the new JENDL-4.0 library on criticality experiments of a research reactor with oxide LEU (20 w/o) fuel, light water moderator and beryllium reflectors

International Nuclear Information System (INIS)

Liem, Peng Hong; Sembiring, Tagor Malem

2012-01-01

Highlights: ► Benchmark calculations of the new JENDL-4.0 library. ► Thermal research reactor with oxide LEU fuel, H 2 O moderator and Be reflector. ► JENDL-4.0 library shows better C/E values for criticality evaluations. - Abstract: Benchmark calculations of the new JENDL-4.0 library on the criticality experiments of a thermal research reactor with oxide low enriched uranium (LEU, 20 w/o) fuel, light water moderator and beryllium reflector (RSG GAS) have been conducted using a continuous energy Monte Carlo code, MVP-II. The JENDL-4.0 library shows better C/E values compared to the former library JENDL-3.3 and other world-widely used latest libraries (ENDF/B-VII.0 and JEFF-3.1).

Numerical simulation research on rolling process of monolithic nuclear fuel plate

International Nuclear Information System (INIS)

Wan Jibo; Kong Xiangzhe; Ding Shurong; Xu Hongbin; Huo Yongzhong

2015-01-01

For the strain-rate-dependent constitutive relation of zircaloy cladding in UMo monolithic nuclear fuel plates, the three-dimensional stress updating algorithm was derived out, and the corresponding VUMAT subroutine to define its constitutive relation was developed and validated; the finite element model was built to simulate the frame rolling process of UMo monolithic nuclear fuel plates; with the explicit dynamic finite element method, the evolution rules of the deformation and contact pressure during the rolling process within the composite slab were obtained and analyzed. The research results indicate that it is convenient and efficient to define the strain-rate- dependent constitutive relations of materials with the user-defined material subroutine VUMAT; the rolling-induced contact pressure between the fuel meat and the covers varies with time, and the maximum pressure exits at the symmetric plane along the plate width direction. This study supplies a foundation and a computation method for optimizing the processing parameters to manufacture UMo monolithic nuclear fuel plates. (authors)

Fuel plate stability experiments and analysis for the Advanced Neutron Source

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Luttrell, C.R.; Yahr, G.T.

1992-01-01

The planned Advanced Neutron Source (ANS) and several existing reactors use closely spaced arrays of involute shaped fuel-plates which are cooled by water flowing through the channels between the plates. There is concern that at certain coolant flow velocities adjacent plates may deflect and touch, with resulting failure of the plates. Experiments have been conducted at the Oak Ridge National Laboratory to examine this potential phenomenon. Results of the experiments and comparison with analytical predictions are reported in this paper. The tests were conducted using full scale epoxy plate models of the aluminum/uranium silicide ANS involute shaped fuel plates. Use of epoxy plates and model theory allowed lower flow velocities and pressures to explore the potential failure mechanism. Plate deflections and channel pressures as function of the flow velocity are examined. Comparisons with mathematical models are noted. 12 refs

Fuel plate stability experiments and analysis for the Advanced Neutron Source

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Luttrell, C.R.; Yahr, G.T.

1993-05-01

The planned reactor for the Advanced Neutron Source (ANS) will use closely spaced arrays of involute-shaped fuel plates that will be cooled by water flowing through the channels between the plates. There is concern that at certain coolant flow velocities, adjacent plates may deflect and touch, with resulting failure of the plates. Experiments have been conducted at the Oak Ridge National Laboratory to examine this potential phenomenon. Results of the experiments and comparison with analytical predictions are reported. The tests were conducted using full-scale epoxy plate models of the aluminum/uranium silicide ANS involute-shaped fuel plates. Use of epoxy plates and model theory allowed lower flow velocities and pressures to explore the potential failure mechanism. Plate deflections and channel pressures as functions of the flow velocity are examined. Comparisons with mathematical models are noted

Automated ultrasonic scanning of flat plate nuclear fuel

International Nuclear Information System (INIS)

Barna, B.A.

1979-01-01

One of the most challenging problems in Non-Destructive Testing lies in making the inspection as rapid, precise, cost effective and operator independent as possible. Only by optimizing these four factors can a technology take full advantage of the quality control possible with NDT. This paper describes a highly complex application of high frequency ultrasonics to image extremely small and difficult to detect flaws in a production line environment. The objects of interest are flat plate nuclear fuel used in the Advanced Test Reactor at the Idaho National Engineering Laboratory. The plates are fabricated by hot rolling a sandwich of alloyed uranium fuel and aluminum cladding. After rolling, the block is flattened to a long thin plate approximately 1.27 m (55 inches) long, 102 mm (4 inches) wide and 1.25 mm (0.050 inches) thick. The core, or fuel area is nominally 0.75 mm (0.030 inches) thick with 0.25 mm (0.010 inches) of aluminum bonded to both sides. As might be expected the fabrication is a sensitive process which can introduce several flaws detrimental to the reactor operation if they are undetected. Two of the characteristics that must be examined are the cladding thickness of the aluminum left over the fuel and the quality of bond between the cladding and the fuel. If either the cladding is too thin or the bonding inadequate thermal and/or corrosive activity can crack the protective cladding

Highly conductive composites for fuel cell flow field plates and bipolar plates

Science.gov (United States)

Jang, Bor Z; Zhamu, Aruna; Song, Lulu

2014-10-21

This invention provides a fuel cell flow field plate or bipolar plate having flow channels on faces of the plate, comprising an electrically conductive polymer composite. The composite is composed of (A) at least 50% by weight of a conductive filler, comprising at least 5% by weight reinforcement fibers, expanded graphite platelets, graphitic nano-fibers, and/or carbon nano-tubes; (B) polymer matrix material at 1 to 49.9% by weight; and (C) a polymer binder at 0.1 to 10% by weight; wherein the sum of the conductive filler weight %, polymer matrix weight % and polymer binder weight % equals 100% and the bulk electrical conductivity of the flow field or bipolar plate is at least 100 S/cm. The invention also provides a continuous process for cost-effective mass production of the conductive composite-based flow field or bipolar plate.

Study on Al-alloy or silicide LEU for DR3 in Denmark

Energy Technology Data Exchange (ETDEWEB)

Haack, Karsten [Riso National Laboratory, DK 4000 Roskilde (Germany)

1985-07-01

The 10 MW D{sub 2}0-moderated and -cooled research reactor DR3 has at present HEU fuel available for continued operation till early 19. This report presents the status of a feasibility study prepared for selection of the best suited candidate LEU fuel type for DR3 at a potential conversion in 1988. At the moment two alternatives are evaluated: UAl-alloy with modified geometry and U{sub 3}Si{sub 2} with unchanged geometry. A decision on the type selected for further investigation is expected late 1984. The investigation should comprise development, in- and out-of-pile--testing and licensing activities on the potential LEU option. (author)

Development of core technology for research reactors using plate type fuels

International Nuclear Information System (INIS)

Ha, Jae Joo; Lee, Doo Jeong; Park, Cheol

2009-12-01

Around 250 research reactors are under operation over the world. However, about 2/3 have been operated more than 30 years and demands for replacements are expected in the near future. The number of expected units is around 110, and around 55 units from 40 countries will be expected to be bid in the world market. In 2007, Netherlands started international bidding process to construct a new 80MW RR (named PALLAS) with the target of commercial operation in 2016, which will replace the existing HFR(45MW). KAERI consortium has been participated in that bid. Most of RRs use plate type fuels as a fuel assembly, Be and Graphite as a reflector. On the other hand, in Korea, the KAERI is operating the HANARO, which uses a rod type fuel assembly and heavy water as a reflector. Hence, core technologies for RRs using plate type fuels are in short. Therefore, core technologies should be secured for exporting a RR. In chapter 2, the conceptual design of PALLAS which use plate type fuels are described including core, cooling system and connected systems, layout of general components. Experimental verification tests for the plate type fuel and second shutdown system and the code verification for nuclear design are explained in Chapter 3 and 4, respectively

An Expert System to Analyze Homogeneity in Fuel Element Plates for Research Reactors

International Nuclear Information System (INIS)

Tolosa, S.C.; Marajofsky, A.

2004-01-01

In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up. This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to x-ray images. These images are generated when the x-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized x-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate

An expert system to analyze homogeneity in fuel element plates for research reactors

International Nuclear Information System (INIS)

Cativa Tolosa, Sebastian; Marajofsky, Adolfo

2004-01-01

In the manufacturing control of Fuel Element Plates for Research Reactors, one of the problems to be addressed is how to determine the U-density homogeneity in a fuel plate and how to obtain qualitative and quantitative information in order to establish acceptance or rejection criteria for such, as well as carrying out the quality follow-up.This paper is aimed at developing computing software which implements an Unsupervised Competitive Learning Neural Network for the acknowledgment of regions belonging to a digitalized gray scale image. This program is applied to X-ray images. These images are generated when the X-ray beams go through a fuel plate of approximately 60 cm x 8 cm x 0.1 cm thick. A Nuclear Fuel Element for Research Reactors usually consists of 18 to 22 of these plates, positioned in parallel, in an arrangement of 8 x 7 cm. Carrying out the inspection of the digitalized X-ray image, the neural network detects regions with different luminous densities corresponding to U-densities in the fuel plate. This is used in quality control to detect failures and verify acceptance criteria depending on the homogeneity of the plate. This modality of inspection is important as it allows the performance of non-destructive measurements and the automatic generation of the map of U-relative densities of the fuel plate. (author)

Alternative bipolar plates design and manufacturing for PEM fuel cell

International Nuclear Information System (INIS)

Lee Chang Chuan; Norhamidi Muhamad; Jaafar Sahari

2006-01-01

Bipolar plates is one of the important components in fuel cell stack, it comprise up to 80% of the stack volume. Traditionally, these plates have been fabricated from graphite, owing to its chemical nobility, and high electrical and thermal conductivity; but these plates are brittle and relatively thick. Therefore increasing the stack volume and size. Alternatives to graphite are carbon-carbon composite, carbon-polymer composite and metal (aluminum, stainless steel, titanium and nickel based alloy). The use of coated and uncoated metal bipolar plates has received attention recently due to the simplicity of plate manufacturing. The thin nature of the metal substrate allows for smaller stack design with reduced weight. Lightweight coated metals as alternative to graphite plate is being developed. Beside the traditional method of machining and slurry molding, metal foam for bipolar plates fabrication seems to be a good alternative. The plates will be produced with titanium powder by Powder Metallurgy method using space holders technique to produce the meal foam flow-field. This work intends to facilitate the materials and manufacturing process requirements to produce cost effective foamed bipolar plates for fuel cell

Comparison of MCNP and WIMS-AECL/RFSP calculations against critical heavy water experiments in ZED-2 with CANFLEX-LVRF and CANFLEX-LEU fuels

International Nuclear Information System (INIS)

Bromley, B. P.; Watts, D. G.; Pencer, J.; Zeller, M.; Dweiri, Y.

2009-01-01

This paper summarizes calculations of MCNP5 and WIMS-AECL/RFSP compared against measurements in coolant void substitution experiments in the ZED-2 critical facility with CANFLEX R-LEU/RU (Low Enriched Uranium, Recovered Uranium) reference fuels and CANFLEX-LVRF (Low Void Reactivity Fuel) test fuel, and H 2 O/air coolants. Both codes are tested for the prediction of the change in reactivity with complete voiding of all fuel channels, and that for a checkerboard voiding pattern. Understanding these phenomena is important for the ACR-1000 R reactor. Comparisons are also made for the prediction of the axial and radial neutron flux distributions, as measured by copper foil activation. The experimental data for these comparisons were obtained from critical mixed lattice / substitution experiments in AECL's ZED-2 critical facility using CANFLEX-LEU/RU and CANFLEX-LVRF fuel in a 24-cm square lattice pitch at 25 degrees C. Substitution analyses were performed to isolate the properties (buckling, bare critical lattice dimensions) of the CANFLEX-LVRF fuel. This data was then used to further test the lattice physics codes. These comparisons establish biases/uncertainties and errors in the calculation of k eff , coolant void reactivity, checkerboard coolant void reactivity, and flux distributions. Results show small to modest biases in void reactivity and very good agreement for flux distributions. The importance of boundary conditions and the modeling of un-moderated fuel in the critical experiments are demonstrated. This comparison study provides data that supports code validation and gives good confidence in the reactor physics tools used in the design and safety analysis of the ACR-1000 reactor. (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.

Composite Bipolar Plate for Unitized Fuel Cell/Electrolyzer Systems

Science.gov (United States)

Mittelsteadt, Cortney K.; Braff, William

2009-01-01

In a substantial improvement over present alkaline systems, an advanced hybrid bipolar plate for a unitized fuel cell/electrolyzer has been developed. This design, which operates on pure feed streams (H2/O2 and water, respectively) consists of a porous metallic foil filled with a polymer that has very high water transport properties. Combined with a second metallic plate, the pore-filled metallic plates form a bipolar plate with an empty cavity in the center.

Experimental investigation of critical velocity in a parallel plate research reactor fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Castro, Alfredo J.A.; Scuro, Nikolas L.; Andrade, Delvonei A., E-mail: ajcastro@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

The fuel elements of a MTR (Material Testing Reactor) type nuclear reactor are mostly composed of aluminum coated fuel plates containing the core of uranium silica (U{sub 3}Si{sub 2}) dispersed in an aluminum matrix. These plates have a thickness of the order of millimeters and are much longer in relation to their thickness. They are arranged in parallel in the assembly of the fuel element to form channels between them a few millimeters in thickness, through which there is a flow of the coolant. This configuration, combined with the need for a flow at high flow rates to ensure the cooling of the fuel element in operation, may create problems of mechanical failure of fuel plate due to the vibration induced by the flow in the channels. In the case of critical velocity excessive permanent deflections of the plates can cause blockage of the flow channel in the reactor core and lead to overheating in the plates. For this study an experimental bench capable of high volume flows and a test section that simulates a plate-like fuel element with three cooling channels were developed. The dimensions of the test section were based on the dimensions of the Fuel Element of the Brazilian Multipurpose Reactor (RMB), whose project is being coordinated by the National Commission of Nuclear Energy (CNEN). The experiments performed attained the objective of reaching Miller's critical velocity condition. The critical velocity was reached with 14.5 m/s leading to the consequent plastic deformation of the flow channel plates. (author)

Modeling of the heat transfer performance of plate-type dispersion nuclear fuel elements

Science.gov (United States)

Ding, Shurong; Huo, Yongzhong; Yan, XiaoQing

2009-08-01

Considering the mutual actions between fuel particles and the metal matrix, the three-dimensional finite element models are developed to simulate the heat transfer behaviors of dispersion nuclear fuel plates. The research results indicate that the temperatures of the fuel plate might rise more distinctly with considering the particle swelling and the degraded surface heat transfer coefficients with increasing burnup; the local heating phenomenon within the particles appears when their thermal conductivities are too low. With rise of the surface heat transfer coefficients, the temperatures within the fuel plate decrease; the temperatures of the fuel plate are sensitive to the variations of the heat transfer coefficients whose values are lower, but their effects are weakened and slight when the heat transfer coefficients increase and reach a certain extent. Increasing the heat generation rate leads to elevating the internal temperatures. The temperatures and the maximum temperature differences within the plate increase along with the particle volume fractions. The surface thermal flux goes up along with particle volume fractions and heat generation rates, but the effects of surface heat transfer coefficients are not evident.

Nuclear reactor fuel structure containing uranium alloy wires embedded in a metallic matrix plate

International Nuclear Information System (INIS)

Travelli, A.

1988-01-01

A nuclear fuel-containing plate structure for a nuclear reactor is described; such structure comprising a pair of malleable metallic non-fissionable matrix plates having confronting surfaces which are pressure bonded together and fully united to form a bonded surface, and elongated malleable wire-like fissionable fuel members separately confined and fully enclosed between the matrix plates along the interface to afford a high fuel density as well as structural integrity and effective retention of fission products. The plates have separate recesses formed in the confronting surfaces for closely receiving the wire-like fissionable fuel members. The wire-like fissionable fuel members are made of a maleable uranium alloy capable of being formed into elongated wire-like members and capable of withstanding pressure bonding. The wire-like fissionable fuel members are completely separated and isolated by fully united portions of the interface

Release of fission products from miniature fuel plates at elevated temperature

International Nuclear Information System (INIS)

Posey, J.C.

1982-01-01

Three miniature fuel plates were tested at progressively higher temperatures. A U 3 Si plated blistered and released fission gases at 500 0 C. Two U 3 O 8 filled plates blistered and released fission gases at 550 0 C

Unirradiated characteristics of U-Si alloys as dispersed-phase fuels

International Nuclear Information System (INIS)

Domagala, R.F.; Wiencek, T.C.

1987-06-01

To satisfy the power demands of many research reactors, a new LEU fuel with a high density and U content was needed. Any fuel must be compatible with Al and its alloys so that it may be fabricable as a dispersed-phase in Al alloy and Al matrix plate-type elements following, as nearly as possible, established commercial manufacturing techniques. U-Si and U-Si-Al alloys at or near the composition of U 3 Si were immediately attractive because of work documented by the Canadians. 8 refs., 2 figs

Neutron flux measurement in the central channel (XC-1) of TRIGA 14 MW LEU core

International Nuclear Information System (INIS)

BARBOS, D.; BUSUIOC, P.; ROTH, Cs.; PAUNOIU, C.

2008-01-01

The TRIGA 14 MW reactor, operated by Institute for Nuclear Research Pitesti, Romania, is a pool type reactor, and has a rectangular shape which holds fuel bundles and is surrounded with beryllium reflectors. Each fuel bundle is composed of 25 nuclear fuel rods. The TRIGA 14 MW reactor was commissioned 28 years ago with HEU fuel rods. The conversion was gradually achieved, starting in February 1992 and completed in March 2006. The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel, replaced by LEU fuel, accompanied by a large set of theoretical evaluation and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC in the XC-1 water 1 water-filled channel were performed using multi multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral neutron flux value for LEU core is greater of 13% than for the standard HEU core. Also thermal neutron flux value for converted LEU core is smaller by 0.38% than for the standard HEU core. These differences appear because the foil activation detectors have been irradiated using a pneumatic rabbit having a diameter of 32 mm, whereas foil irradiations in standard HEU core has been performed with a pneumatic rabbit having a diameter of 14 mm, and therefore the neutron spectra in LEU core is less thermalized and the weight of fast neutron is greater

Release of fission products from miniature fuel plates at elevated temperature

International Nuclear Information System (INIS)

Posey, John C.

1983-01-01

Three miniature fuel plates were tested at progressively higher temperatures. A U 3 Si filled plate blistered and released fission gases at 500 deg. C. Two U 3 O 8 filled plates blistered and released fission gases at 550 deg. C. (author)

TRIGA low enrichment fuel

International Nuclear Information System (INIS)

Gietzen, A.

1993-01-01

Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with General Atomic's standard commercial warranty

TRIGA low enrichment fuel

International Nuclear Information System (INIS)

Gietzen, A.

1993-01-01

Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium-zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

French LEU fuel for research reactor with emphasis on the Osiris experience of core conversion and reactor operation with the new fuel

International Nuclear Information System (INIS)

Cerles, J.-M.

1981-09-01

One of the various activities carried out in France concerned with the design, fabrication and development of nuclear fuels was the development by the CEA of a plate type fuel (Caramel fuel). A Caramel fuel element is in the form of a plate consisting of two tight covering zircaloy sheets in which the UO 2 platelets are confined themselves within the network of a zircaloy grid. The plane geometry provides an effective means of overcoming the drawback of poor uranium oxide conductivity, and makes it possible to combine high specific power with low fuel temperature. The chief advantages of this fuel are the following: it is a very low enriched fuel. It can be used in research reactors demanding high volumetric powers and neutron fluxes, with a required enrichment significantly lower than 20% 235 U. The difference between the densities of UO 2 matrix and U-Al, 10.3 and 1.6 g/cm respectively, leads to a higher uranium charge, making it possible to reduce the enrichment to between 3 and 10%. Owing to fuel dispersion, any loss of tightness only puts a small amount of fissile material in contact with the coolant, thus limiting any contamination of the primary circuit. Another safety factor is the operating temperature, which is considerably lower than the temperature at which fission gases are liberated

Parametric study of the deformation of U3Si2-Al dispersion fuel plates

International Nuclear Information System (INIS)

Vieira, Edeval

2011-01-01

The Nuclear and Energy Research Institute - IPEN-CNEN/SP produces routinely the nuclear fuel necessary for operating its research reactor, IEA-R1. This fuel consists of fuel plates containing U 3 Si 2 -Al composites as the meat, which are fabricated by rolling. The rolling process currently deployed was developed with base on information obtained from literature, which were used as premises for defining the current manufacturing procedures, according to a methodology with essentially empirical character. Despite the current rolling process to be perfectly stable and highly reproducible, it is not well characterized and therefore is not fully known. The objective of this work is to characterize the rolling process for producing fuel plates, specifically the evolution of dimensional parameters of the fuel plate as a function of its deformation in the rolling process. Results are presented in terms of the evolution of the thickness of the fuel meat and cladding of the fuel plate along the deformation, as well as the terminals defects, microstructure and porosity of the fuel meat. (author)

Analysis of Advanced Fuel Kernel Technology

International Nuclear Information System (INIS)

Oh, Seung Chul; Jeong, Kyung Chai; Kim, Yeon Ku; Kim, Young Min; Kim, Woong Ki; Lee, Young Woo; Cho, Moon Sung

2010-03-01

The reference fuel for prismatic reactor concepts is based on use of an LEU UCO TRISO fissile particle. This fuel form was selected in the early 1980s for large high-temperature gas-cooled reactor (HTGR) concepts using LEU, and the selection was reconfirmed for modular designs in the mid-1980s. Limited existing irradiation data on LEU UCO TRISO fuel indicate the need for a substantial improvement in performance with regard to in-pile gaseous fission product release. Existing accident testing data on LEU UCO TRISO fuel are extremely limited, but it is generally expected that performance would be similar to that of LEU UO 2 TRISO fuel if performance under irradiation were successfully improved. Initial HTGR fuel technology was based on carbide fuel forms. In the early 1980s, as HTGR technology was transitioning from high-enriched uranium (HEU) fuel to LEU fuel. An initial effort focused on LEU prismatic design for large HTGRs resulted in the selection of UCO kernels for the fissile particles and thorium oxide (ThO 2 ) for the fertile particles. The primary reason for selection of the UCO kernel over UO 2 was reduced CO pressure, allowing higher burnup for equivalent coating thicknesses and reduced potential for kernel migration, an important failure mechanism in earlier fuels. A subsequent assessment in the mid-1980s considering modular HTGR concepts again reached agreement on UCO for the fissile particle for a prismatic design. In the early 1990s, plant cost-reduction studies led to a decision to change the fertile material from thorium to natural uranium, primarily because of a lower long-term decay heat level for the natural uranium fissile particles. Ongoing economic optimization in combination with anticipated capabilities of the UCO particles resulted in peak fissile particle burnup projection of 26% FIMA in steam cycle and gas turbine concepts

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

Feasibility of Electromagnetic Acoustic Evaluation for Quality Test of a Plate-type Nuclear Fuel

International Nuclear Information System (INIS)

Jung, Hyun Kyu; Lee, Yoon Sang; Cheong, Yong Moo

2010-01-01

Most research and test reactors use the nuclear fuel plates which are consisted of a fuel core in aluminum alloy. Recently KAERI signed a deal with the Jordan Atomic Energy Commission to build the research reactor and have to supply the plate-type nuclear fuels. For the demands of world market, KAERI started the research and development of the plate-type fuel elements and endeavored to achieve a localization of fuel fabrication. For the inspection of plate-type fuel elements to be used in Research Reactors, an immersion pulse-echo ultrasonic technique was applied. This inspection was done with water, so a nuclear fuel was immersed to be prone to corrosion and needed to have time and cost due to an additional process. The sample that will be examined within this paper is a non-ferromagnetic material such as aluminum which has a good acousto-elastic property, for an effective inspection of a bond quality for a nuclear fuel under a manufacturing environment. The purpose of this study is to investigate the feasibility of an EMAT technology for an automated inspection of a nuclear fuel without water

Safety assessment of U–Mo fuel mini plates irradiated in HANARO reactor

International Nuclear Information System (INIS)

Jo, Daeseong; Kim, Haksung

2015-01-01

Highlights: • Neutronic and thermal-hydraulic analyses of U–Mo fuel irradiated in HANARO reactor. • A mock-up irradiation target was designed and tested to measure the flow rate. • During normal operation, boiling does not occur. • During limiting accidents, boiling occurs. However, fuel integrity is maintained. - Abstract: Neutronic and thermal hydraulic characteristics of U–Mo fuel mini plates irradiated in the HANARO reactor were analyzed for the safety assessment of these plates. A total of eight fuel plates were double-stacked; each stack contained three 8.0 gU/cc U–7Mo fuel plates and one 6.5 gU/cc U–7Mo fuel plate. The neutronic and thermal hydraulic analyses were carried out using the MCNP code and TMAP code, respectively. The core status used in the study was the equilibrium core, and four Control Absorber Rod (CAR) locations were considered: 350 mm, 450 mm, 550 mm, and 650 mm away from the bottom of the core. For the fuels in the lower stack, the maximum heat flux was found at the CAR located at 450 mm. For the fuels in the upper stack, the maximum heat flux was found at the CAR located at 650 mm. The axial power distributions for the upper and lower stacks were selected on the basis of thermal margin analyses. A mock-up irradiation target assembly was designed and tested at the out-of-pile test facility to measure the flow rate through the irradiation site, given that the maximum flow rate through the irradiation site at the HANARO reactor is limited to 12.7 kg/s. For conservative analyses, measurement and correlation uncertainties and engineering hot channel factors were considered. During normal operation, the minimum ONB temperature margins for the lower and upper stacks are 41.6 °C and 31.8 °C, respectively. This means that boiling does not occur. However, boiling occurs during the limiting accidents. Nevertheless, the fuel integrity is maintained since the minimum DNBR are 1.96 for the Reactivity Insertion Accident (RIA) and 2

Development of LEU targets for 99Mo production and their chemical processing status 1989

International Nuclear Information System (INIS)

Vandegrift, G.F.; Kwok, J.D.; Chamberlain, D.B.; Hoh, J.C.; Streets, E.W.; Vogler, S.; Thresh, H.R.; Domagala, R.F.; Wiencek, T.C.; Matos, J.E.

1991-01-01

Most of the world's supply of Tc-99m for medical purposes is currently produced from Mo-99 derived from the fissioning of high enriched uranium (HEU). Substitution of low enriched uranium (LEU) silicide fuel for the HEU alloy and aluminide fuels used in current target designs will allow equivalent Mo-99 yields with no change in target geometries. Substitution of uranium metal will also allow the substitution of LEU for HEU. Efforts performed in 1989 focused on (1) fabrication of a uranium metal target by Hot Isostatic Pressing uranium metal foil to zirconium, (2) experimental investigation of the dissolution step for U 3 Si 2 targets, allowing us to present a conceptual design for the dissolution process and equipment, and (3) investigation of the procedures used to reclaim irradiated uranium from Mo-production targets, allowing us to further analyze the waste and by-product problems associated with the substitution of LEU for HEU. (orig.)

Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

Energy Technology Data Exchange (ETDEWEB)

Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [Unidad de Actividad Combustibles Nucleares Comision Nacional de Energia Atomica (CNE4), Avda. del Libertador, 8250 C1429BNO Buenos Aires (Argentina)

2002-07-01

One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm{sup 3}. PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

Advances and highlights of the CNEA qualification program as high density fuel manufacturer for research reactors

International Nuclear Information System (INIS)

Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H.

2002-01-01

One of the main objectives of CNEA regarding the fuel for research reactors is the development and qualification of the manufacturing of LEU high-density fuels. The qualification programs for both types of fuels, Silicide fuel and U- x Mo fuel, are similar. They include the following activities: development and set up of the fissile compound manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of mini plates and plates, design and fabrication of fuel assembly prototypes for irradiation, post-irradiation examination and feedback for manufacturing improvements. This paper describes the different activities performed within each program during the last year and the main advances and achievements of the programs within this period. The main achievements may be summarized in the following activities: Continuation of the irradiation of the first silicide fuel element in the R A3. Completion of the manufacturing of the second silicide fuel element, licensing and beginning of its irradiation in the R A3. Development of the HMD Process to manufacture U-Mo powder (pUMA project). Set up of fuel plates manufacturing at industrial level using U-Mo powder. Preliminary studies and the design for the irradiation of mini plates, plates and full scale fuel elements with U-Mo and 7 g U/cm 3 . PIE destructive studies for the P-04 silicide fuel prototype (accurate burnup determination through chemical analysis, metallography and SEM of samples from the irradiated fuel plates). Improvement and development of new characterization techniques for high density fuel plates quality control including US testing and densitometric analysis of X-ray examinations. The results obtained in this period are encouraging and also allow to foresee a wider participation of CNEA in the international effort to qualify U-Mo as a new material for the manufacturing of research reactor fuels. (author)

Low-enriched fuel particle performance review

International Nuclear Information System (INIS)

Homan, F.; Nabielek, H.; Yang, L.

1978-08-01

The available data on low-enriched uranium (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance, and fuel particle chemical performance

Babcock and Wilcox plate fabrication experience with uranium silicide spherical fuel

International Nuclear Information System (INIS)

Todd, Lawrence E.; Pace, Brett W.

1996-01-01

This report is written to present the fuel fabrication experience of Babcock and Wilcox using atomized spherical uranium silicide powder. The intent is to demonstrate the ability to fabricate fuel plates using spherical powder and to provide useful information proceeding into the next phase of work using this type of fuel. The limited quantity of resources- spherical powder and time, did not allow for much process optimizing in this work scope. However, the information contained within provides optimism for the future of spherical uranium silicide fuel plate fabrication at Babcock and Wilcox.The success of assembling fuel elements with spherical powder will enable Babcock and Wilcox to reduce overall costs to its customers while still maintaining our reputation for providing high quality research and test reactor products. (author)

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

Connection between end plates and rods in a BWR fuel element

International Nuclear Information System (INIS)

Cali', G.P.

1975-01-01

The problem of the connection between the end plates and the rods of a BWR fuel element is analytically formulated. The behaviour of the springs coupling the rods with the upper plate is analyzed with particular detail since the deformation of these springs affects the forces at the interface of the fuel element structure components. A tool is given to design the springs according to some considerations regarding the mechanical strength of the interacting components as well as the influence of the possible geometrical unevennes of the system that can arise during the fuel element lifetime. (Cali', G.P.)

Low-enriched fuel particle performance review

International Nuclear Information System (INIS)

Homan, F.; Nabielek, H.; Yang, L.

1978-08-01

The available data on low-enriched (LEU) fuel particles were reviewed under the United States-Federal Republic of Germany Agreement. The most influential factors controlling the irradiation performance of LEU fuel particles were found to be plutonium transport, fission product transport, fuel particle mechanical performance and fuel particle chemical performande. (orig.) [de

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)

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)

The Role of Friction Stir Welding in Nuclear Fuel Plate Fabrication

International Nuclear Information System (INIS)

Burkes, D.; Medvedev, P.; Chapple, M.; Amritkar, A.; Wells, P.; Charit, I

2009-01-01

The friction bonding process combines desirable attributes of both friction stir welding and friction stir processing. The development of the process is spurred on by the need to fabricate thin, high density, reduced enrichment fuel plates for nuclear research reactors. The work seeks to convert research and test reactors currently operating on highly enriched uranium fuel to operate on low enriched uranium fuel without significant loss in reactor performance, safety characteristics, or significant increase in cost. In doing so, the threat of global nuclear material proliferation will be reduced. Feasibility studies performed on the process show that this is a viable option for mass production of plate-type nuclear fuel. Adapting the friction stir weld process for nuclear fuel fabrication has resulted in the development of several unique ideas and observations. Preliminary results of this adaptation and process model development are discussed

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

Remarks on the influence of enrichment reduction on fuel cycle costs

International Nuclear Information System (INIS)

Krull, W.

1985-01-01

The cost factors influencing the fuel cycle cost analysis for research reactors are discussed in detail with special emphasis on fuel element fabrication costs, burnup and reprocessing costs. Two different aspects for the conversion from HEU to LEU are considered: plus 14% U-235 weight per LEU fuel element and plus ca. 50 % U-235 weight per LEU fuel element. The cost factors and these conversion aspects were taken for calculating the changes in fuel cycle costs for the three different meat materials U 3 O 8 , U 3 Si 2 and U 3 Si. The results of these calculations can be summarized as following: - if in the HEU case the fuel loading and the burnup of a fuel element is low there will be some economic advantages in the LEU case; - if in the HEU case the fuel loading and the burnup of a fuel element is high there will be economic disadvantages in the LEU case. (author)

Examinations of the irradiation behaviour of U3Si2 test fuel plates with low enrichment

International Nuclear Information System (INIS)

Muellauer, J.

1989-01-01

Five low-enriched (19.7% 235 U), high-density (4.7 gU/cm/ 3 ) U 3 Si 2 -test fuel plates (miniplates) with different fine grain contents have been qualified under irradiation. During the course of irradiation up to burnup of 63% 235 U depletion, no released fractions of gaseous or solid fission products from the fuel plate to the rig coolant were detected. The measured swelling rate of the fuel zone (meat) is less than 0.45% ΔV/10 20 fissions/cm 3 the blister-threshold temperature of the fuel plates is above 520 0 C. The favourable irradiation behavior of the U 3 Si 2 fuel plates was not influenced by using higher amounts of fine grained particles (40% [de

Loss-of-Flow and Loss-of-Pressure Simulations of the BR2 Research Reactor with HEU and LEU Fuel

Energy Technology Data Exchange (ETDEWEB)

Licht, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Sikik, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The reactor core of BR2 is located inside a pressure vessel that contains 79 channels in a hyperboloid configuration. The core configuration is highly variable as each channel can contain a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Because of this variability, a representative core configuration, based on current reactor use, has been defined for the fuel conversion analyses. The code RELAP5/Mod 3.3 was used to perform the transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. The input model has been modernized relative to that historically used at BR2 taking into account the best modeling practices developed by Argonne National Laboratory (ANL) and BR2 engineers.

Development of long-life low enrichment fuel

International Nuclear Information System (INIS)

Gietzen, A.J.; West, G.B.

1978-01-01

With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on non-proliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U. S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of this year, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

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)

Irradiation Performance of U-Mo Alloy Based ‘Monolithic’ Plate-Type Fuel – Design Selection

Energy Technology Data Exchange (ETDEWEB)

A. B. Robinson; G. S. Chang; D. D. Keiser, Jr.; D. M. Wachs; D. L. Porter

2009-08-01

A down-selection process has been applied to the U-Mo fuel alloy based monolithic plate fuel design, supported by irradiation testing of small fuel plates containing various design parameters. The irradiation testing provided data on fuel performance issues such as swelling, fuel-cladding interaction (interdiffusion), blister formation at elevated temperatures, and fuel/cladding bond quality and effectiveness. U-10Mo (wt%) was selected as the fuel alloy of choice, accepting a somewhat lower uranium density for the benefits of phase stability. U-7Mo could be used, with a barrier, where the trade-off for uranium density is critical to nuclear performance. A zirconium foil barrier between fuel and cladding was chosen to provide a predictable, well-bonded, fuel-cladding interface, allowing little or no fuel-cladding interaction. The fuel plate testing conducted to inform this selection was based on the use of U-10Mo foils fabricated by hot co-rolling with a Zr foil. The foils were subsequently bonded to Al-6061 cladding by hot isostatic pressing or friction stir bonding.

EVALUATION OF U10MO FUEL PLATE IRRADIATION BEHAVIOR VIA NUMERICAL AND EXPERIMENTAL BENCHMARKING

Energy Technology Data Exchange (ETDEWEB)

Samuel J. Miller; Hakan Ozaltun

2012-11-01

This article analyzes dimensional changes due to irradiation of monolithic plate-type nuclear fuel and compares results with finite element analysis of the plates during fabrication and irradiation. Monolithic fuel plates tested in the Advanced Test Reactor (ATR) at Idaho National Lab (INL) are being used to benchmark proposed fuel performance for several high power research reactors. Post-irradiation metallographic images of plates sectioned at the midpoint were analyzed to determine dimensional changes of the fuel and the cladding response. A constitutive model of the fabrication process and irradiation behavior of the tested plates was developed using the general purpose commercial finite element analysis package, Abaqus. Using calculated burn-up profiles of irradiated plates to model the power distribution and including irradiation behaviors such as swelling and irradiation enhanced creep, model simulations allow analysis of plate parameters that are either impossible or infeasible in an experimental setting. The development and progression of fabrication induced stress concentrations at the plate edges was of primary interest, as these locations have a unique stress profile during irradiation. Additionally, comparison between 2D and 3D models was performed to optimize analysis methodology. In particular, the ability of 2D and 3D models account for out of plane stresses which result in 3-dimensional creep behavior that is a product of these components. Results show that assumptions made in 2D models for the out-of-plane stresses and strains cannot capture the 3-dimensional physics accurately and thus 2D approximations are not computationally accurate. Stress-strain fields are dependent on plate geometry and irradiation conditions, thus, if stress based criteria is used to predict plate behavior (as opposed to material impurities, fine micro-structural defects, or sharp power gradients), unique 3D finite element formulation for each plate is required.

Performance and Fabrication Status of TREAT LEU Conversion Conceptual Design Concepts

Energy Technology Data Exchange (ETDEWEB)

IJ van Rooyen; SR Morrell; AE Wright; E. P Luther; K Jamison; AL Crawford; HT III Hartman

2014-10-01

Resumption of transient testing at the TREAT facility was approved in February 2014 to meet U.S. Department of Energy (DOE) objectives. The National Nuclear Security Administration’s Global Threat Reduction Initiative Convert Program is evaluating conversion of TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU). This paper describes briefly the initial pre-conceptual designs screening decisions with more detailed discussions on current feasibility, qualification and fabrication approaches. Feasible fabrication will be shown for a LEU fuel element assembly that can meet TREAT design, performance, and safety requirements. The statement of feasibility recognizes that further development, analysis, and testing must be completed to refine the conceptual design. Engineering challenges such as cladding oxidation, high temperature material properties, and fuel block fabrication along with neutronics performance, will be highlighted. Preliminary engineering and supply chain evaluation provided confidence that the conceptual designs can be achieved.

Fuel cell plates with skewed process channels for uniform distribution of stack compression load

Science.gov (United States)

Granata, Jr., Samuel J.; Woodle, Boyd M.

1989-01-01

An electrochemical fuel cell includes an anode electrode, a cathode electrode, an electrolyte matrix sandwiched between electrodes, and a pair of plates above and below the electrodes. The plate above the electrodes has a lower surface with a first group of process gas flow channels formed thereon and the plate below the electrodes has an upper surface with a second group of process gas flow channels formed thereon. The channels of each group extend generally parallel to one another. The improvement comprises the process gas flow channels on the lower surface of the plate above the anode electrode and the process gas flow channels on the upper surface of the plate below the cathode electrode being skewed in opposite directions such that contact areas of the surfaces of the plates through the electrodes are formed in crisscross arrangements. Also, the plates have at least one groove in areas of the surfaces thereof where the channels are absent for holding process gas and increasing electrochemical activity of the fuel cell. The groove in each plate surface intersects with the process channels therein. Also, the opposite surfaces of a bipolar plate for a fuel cell contain first and second arrangements of process gas flow channels in the respective surfaces which are skewed the same amount in opposite directions relative to the longitudinal centerline of the plate.

Basic research on high-uranium density fuels for research and test reactors

International Nuclear Information System (INIS)

Ugajin, M.; Itoh, A.; Akabori, M.

1992-01-01

High-uranium density fuels, uranium silicides (U 3 Si 2 , U 3 Si) and U 6 Me-type uranium alloys (Me = Fe, Mn, Ni), were prepared and examined metallurgically as low-enriched uranium (LEU) fuels for research and test reactors. Miniature aluminum-dispersion plate-type fuel (miniplate) and aluminum-clad disk-type fuel specimens were fabricated and subjected to the neutron irradiation in JMTR (Japan Materials Testing Reactor). Fuel-aluminum compatibility tests were conducted to elucidate the extent of reaction and to identify reaction products. The relative stability of the fuels in an aluminum matrix was established at 350degC or above. Experiments were also performed to predict the chemical form of the solid fission-products in the uranium silicide (U 3 Si 2 ) simulating a high burnup anticipated for reactor service. (author)

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

Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

International Nuclear Information System (INIS)

Zhang Zhuohua; Yu Junchong; Peng Shinian

2014-01-01

According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

ANL progress in developing a target and process for converting CNEA Mo-99 production to LEU

International Nuclear Information System (INIS)

Vandegrift, G.F.; Gelis, A.; Aase, S.; Bakel, A.; Freiberg, E.; Conner, C.

2002-01-01

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 convert 99 Mo production at Argentine Commission Nacional de Energia Atomica (CNEA) from HEU to LEU targets. 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, (2) developing means to improve digestion efficiency, and (3) modifying ion-exchange processes used in the CNEA recovery and purification of 99 Mo to deal with the lower volumes generated from LEU-foil digestion. (author)

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.

The development of in-process inventory walk-through examination system in the process at borrowing inspection between LEU fuel fabrication plants

International Nuclear Information System (INIS)

Nakamura, Norihito; Namekawa, Masaru; Owada, Isao; Kikuchi, Masaru; Kodani, Yoshiki; Nozawa, Yukio

2005-01-01

Since the Nuclear Material Control Center (NMCC) was designed the safeguards inspection organization by Ministry of Education, Culture, Sports, Science and Technology (MEXT) in December 1999, the NMCC has been performing safeguards inspection for the Nuclear Facilities in Japan. The NMCC has carried out the safeguards inspections to LEU Fuel Fabrication Plants (FFPs) and the NMCC has improved the method of safeguards inspection as it has changed over to the integrated safeguards from the year of 2005. Concerning the Borrowing inspection between LEU FFPs, which is the precondition to change over to the integrated safeguards, it is needed to estimate the entire inventory in the facility within the limited time. Therefore, the NMCC has developed the system called IWES (In-process inventory Walk-through Examination System) to examine the inventory in process smoothly, quickly and correctly at borrowing inspection, check the entire inventory quantity and evaluate them. This report describes how IWES aiming at effective/efficient confirmation of in-process inventory has been developed and how it is applied to the borrowing inspection activities. (author)

Determination of elastic constants of fuels plates based on uranium by ultrasound testing

International Nuclear Information System (INIS)

Moreira Castro, Martin Ignacio

2015-01-01

Current nuclear reactors use as U-235 U-enriched compounds enriched with U-235, requiring U-alloys that increase the amount of atoms available for nuclear fission in a convenient way. This study was carried out on fuel plates manufactured in the Chilean Nuclear Energy Commission, whose cores are composed of a dispersed mixture Al-U_3Si_2 and Al-U_7Mo, with different densities of uranium, covered by a coating of Al6061. The objective was to characterize elastically and classify the fuel plates analyzed. Specifically, five Al-U_3Si_2 fuel plates with 1.7 gU/cm"3, eight A-U_3Si_2 with 3.4 gU/cm"3, five of A-l U_3Si_2 with 4.8 gU/cm"3 were successfully studied. The apparent elastic constants (Young and Shear modules, and Poisson coefficient) were determined in the area where the fuel is located (MEAT) by means of an ultrasound sampling technique, thus being able to characterize them and classify them according to their composition. The behavior of the elastic constants generally shows a tendency to decrease as the amount of U_3Si_2 particles dispersed in the MEAT zone of the fuel plates increases. In addition, the non-destructive test method used made it possible to detect several differences between the fuel plates analyzed, such as the amount of reduction in rolling, among others. Additionally, six experimental fuel miniplates were analyzed whose meat were formed by a dispersion of the Al-UMo type, specifically: two of Al-U_7Mo with 6.0 gU/cm"3, two of Al-U_7Mo with 7.0 gU/ cm"3 and two of Al-U_7Mo with 8.0 gU/cm"3. The response of the U-Mo fuel miniplates against this technique was not good, so several ideas were proposed to improve this situation

Documentation Experiences for Jamaican SLOWPOKE-2 Conversion from HEU to LEU

International Nuclear Information System (INIS)

Warner, T.-A.; Dennis, H.; Antoine, J.

2015-01-01

The Jamaican SLOWPOKE–2 (JM–1) is a 20 kW research reactor manufactured by Atomic Energy of Canada Limited and has been operating since March 1984, in the department of the International Centre for Environmental and Nuclear Sciences (ICENS), at the University of the West Indies, Mona Campus in Kingston, Jamaica. The pool type reactor has been primarily used for Neutron Activation Analysis in environmental, agricultural, geochemical, health-related studies and mineral exploration. The University, assisted by the IAEA under the GTRI/RERTR program, is currently in the process of converting from HEU to LEU. Extensive documentation on policies, general requirements, elements of the conversion quality assurance (QA) system and conversion QA administrative procedures is required for the conversion. The core conversion activities are being carried out in accordance with current international standards and regulatory guidelines of the newly established Jamaican Radiation Safety Authority (RSA) with agreement between the RSA and IAEA or DOE related to Nuclear Safety and Control. The documentation structure has taken into consideration nuclear safety and licensing, LEU fuel design and conversion analysis, LEU fuel procurement and fabrication, removal of HEU fuel and reactor maintenance and conversion and commissioning, with the conversion QA manual at the apex of the structure. To a large extent, the documentation format will adhere to that of the IAEA applicable regulatory standards and guidance documents. The major challenge of the conversion activities, it is envisioned, will come from the absence of any previous regulatory framework in Jamaica; however, a timeline for the process, which includes training and equipping of regulators, will guide operation. (author)

Atomics International fuel fabrication facility and low enrichment program [contributed by T.A. Moss, AI

International Nuclear Information System (INIS)

Moss, T.A.

1993-01-01

The AI facility is approximately 30,000 square feet in area and consists of four general areas. One area is devoted to the production of UAl x powder. It consists of a series of arc melting furnaces, crushing lines, glove boxes, and compacting presses. The second area is used for the rolling of fuel plates. The third area is used for the machining of the plates to final size and also the machining of the fuel elements. In the fourth area the fuel plates are swaged into assemblies, and all welding and inspection operations are performed. As part of the lower enrichment program we are scheduled to put a second UAl x powder line into operation and we have had to expand some of our storage area. Under the low enrichment program the AI fuel facility will be modified to accommodate a separate low enrichment Al x production line and compacting line. This facility modification should be done by the end of the fiscal year. We anticipate producing fuel with an enrichment slightly less than 20% We anticipate powder being available for plate production shortly after the facility is completed. Atomics International is scheduled to conduct plate LEU verification work using fully enriched material in the June-July time period, at which time we will investigate what level of uranium loadings we can go to using the current process. It is anticipated that 55 volume percent uranium compound in our fuel form can be achieved

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

Energy Technology Data Exchange (ETDEWEB)

Collette, R. [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); King, J., E-mail: kingjc@mines.edu [Colorado School of Mines, Nuclear Science and Engineering Program, 1500 Illinois St, Golden, CO 80401 (United States); Buesch, C. [Oregon State University, 1500 SW Jefferson St., Corvallis, OR 97331 (United States); Keiser, D.D.; Williams, W.; Miller, B.D.; Schulthess, J. [Nuclear Fuels and Materials Division, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States)

2016-07-15

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends when comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. The results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program. - Highlights: • Automated image processing is used to extract fission gas bubble data from irradiated U−Mo fuel samples. • Verification and validation tests are performed to ensure the algorithm's accuracy. • Fission bubble parameters are predictably difficult to compare across samples of varying compositions. • The 2-D results suggest the need for more homogenized fuel sampling in future studies. • The results also demonstrate the value of 3-D reconstruction techniques.

International interest in the BONAPARTE measurement bench. Post-irradiation examination of lower-enriched fuel plates

International Nuclear Information System (INIS)

2014-01-01

The Belgian Nuclear Research Center SCK-CEN has developed a measurement bench (BONAPARTE) for the non-destructive analysis on fuel plate and rod type fuel elements. BONAPARTE is a modular device that can be employed for many purposes. The article discusses the employment of the BONAPARTE device for the accurate full post-irradiation mapping of fuel plate swelling with degree of precision of just a few micrometers.

Qualification of high-density fuel manufacturing for research reactors at CNEA

Energy Technology Data Exchange (ETDEWEB)

Adelfang, P.; Alvarez, L.; Boero, N.; Calabrese, R.; De La Fuente, M.; Echenique, P.; Markiewicz, M.; Pasqualini, E.; Ruggirello, G.; Taboada, H. [CNEA, Buenos Aires (Argentina)

2001-07-01

CNEA, the National Atomic Energy Commission of Argentina, is at the present a qualified supplier of uranium oxide fuel for research reactors. A new objective in this field is to develop and qualify the manufacturing of LEU high-density fuel for this type of reactors. According with the international trend Silicide fuel and U-xMo fuel are included in our program as the most suitable options. The facilities to complete the qualification of high-density MTR fuels, like the manufacturing plant installations, the reactor, the pool side fuel examination station and the hot cells are fully operational and equipped to perform all the activities required within the program. The programs for both type of fuels include similar activities: development and set up of the fuel material manufacturing technology, set up of fuel plate manufacturing, fabrication and irradiation of miniplates, fabrication and irradiation of full scale fuel elements, post-irradiation examination and feedback for manufacturing improvements. For silicide fuels most of these steps have already been completed. For U-xMo fuel the activities also include the development of alternative ways to obtain U-xMo powder, feasibility studies for large-scale manufacturing and the economical assessment. Set up of U-xMo fuel plate manufacturing is also well advanced and the fabrication of the first full scale prototype is foreseen during this year. (author)

Pore growth in U-Mo/Al dispersion fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo, E-mail: yskim@anl.gov [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Jeong, G.Y.; Sohn, D.-S. [Ulsan National Institute of Science and Technology, 50 UNIST-gil, Eonyang-eup, Ulju-gun, Ulsan, 689-798 (Korea, Republic of); Jamison, L.M. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2016-09-15

U-Mo/Al dispersion fuel is currently under development in the DOE’s Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. The model includes three major topics: fission gas release from the U-Mo and the IL to the pores, stress evolution in the fuel meat, and the effect of amorphous IL growth. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data set from full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model. The model showed fair agreement with the measured data. The model suggested that the growth of the IL has a critical effect on pore growth, as both its material properties and energetics are favorable to pore formation. Therefore, one area of the current effort, focused on suppressing IL growth, appears to be on the right track to improve the performance of this fuel.

Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

International Nuclear Information System (INIS)

Bodey, Isaac T.; Curtis, Franklin G.; Arimilli, Rao V.; Ekici, Kivanc; Freels, James D.

2015-01-01

The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and have

Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

Energy Technology Data Exchange (ETDEWEB)

Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-11-01

The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and have

A review of microstructural analysis on U3Si2-Al plate-type fuel

International Nuclear Information System (INIS)

Ti Zhongxin; Guo Yibai

1995-12-01

The microstructure of U 3 Si 2 -Al plate-type fuel, that is the microstructure of fuel particles, compatibility of the fuel particles and Al matrix, fuel particles distribution, dogbone area morphology, clad and meat thickness, bone quality of clad/frame and clad/fuel core, and the effect of these factors on products quality were comprehensively investigated and analyzed by means of optical microscopy (OM), scanning electron microscopy (SEM), X-ray diffraction (XRD), energy dispersive X-ray spectrometry (EDX), image processing technique, etc.. The main results are as following: U-7.7%Si alloy contains two phases: primary U 3 Si 2 and small amount of USi (about 12%), free-uranium was not detected in fuel particles; the dogbone area is the key factor affecting fuel plate quality (1 ref., 16 figs., 4 tabs.)

Development method for measuring thickness of nuclei and coating of fuel plates

International Nuclear Information System (INIS)

Borges Junior, Reinaldo

2013-01-01

One of the most important components of a nuclear reactor is the Nuclear Fuel. Currently, the most advanced commercial fuel, whose applicability in Brazilian reactors has been developed by IPEN since 1985, is the silicide U 3 Si 2 . This is formed by fuel plates with nuclei dispersion (where the fissile material (U 3 Si 2 ) is homogeneously dispersed in a matrix of aluminum) coated aluminum. This fuel is produced in Brazil with developed technology, the result of the efforts made by the group of manufacturing nuclear fuel (CCN - Center of Nuclear Fuel) of IPEN. Considering the necessity of increasing the power of the IEA- R1 and Brazilian Multipurpose Reactor Building (RMB), for the production of radioisotopes - mainly for the area of medicine - there will be significant increase in the production of nuclear fuel at IPEN. Given this situation, if necessary, make the development of more modern and automated classification techniques. Aiming at this goal, this work developed a new computational method for measuring thickness of core and cladding of fuel plates, which are able to perform such measurements in less time and with more meaningful statistical data when compared with the current method of measurement. (author)

Flow channel shape optimum design for hydroformed metal bipolar plate in PEM fuel cell

Energy Technology Data Exchange (ETDEWEB)

Peng, Linfa; Lai, Xinmin; Liu, Dong' an; Hu, Peng [State Key Laboratory of Mechanical System and Vibration, Shanghai Jiao Tong University, Shanghai 200240 (China); Ni, Jun [Department of Mechanical Engineering and Applied Mechanics, University of Michigan, Ann Arbor, MI 48109 (United States)

2008-03-15

Bipolar plate is one of the most important and costliest components of polymer electrolyte membrane (PEM) fuel cells. Micro-hydroforming is a promising process to reduce the manufacturing cost of PEM fuel cell bipolar plates made of metal sheets. As for hydroformed bipolar plates, the main defect is the rupture because of the thinning of metal sheet during the forming process. The flow channel section decides whether high quality hydroformed bipolar plates can be successively achieved or not. Meanwhile, it is also the key factor that is related with the reaction efficiency of the fuel cell stacks. In order to obtain the optimum flow channel section design prior the experimental campaign, some key geometric dimensions (channel depth, channel width, rib width and transition radius) of flow channel section, which are related with both reaction efficiency and formability, are extracted and parameterized as the design variables. By design of experiments (DOE) methods and an adoptive simulated annealing (ASA) optimization method, an optimization model of flow channel section design for hydroformed metal bipolar plate is proposed. Optimization results show that the optimum dimension values for channel depth, channel width, rib width and transition radius are 0.5, 1.0, 1. 6 and 0.5 mm, respectively with the highest reaction efficiency (79%) and the acceptable formability (1.0). Consequently, their use would lead to improved fuel cell efficiency for low cost hydroformed metal bipolar plates. (author)

Steady-State Thermal-Hydraulics Analyses for the Conversion of the BR2 Reactor to LEU

Energy Technology Data Exchange (ETDEWEB)

Licht, J. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Bergeron, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Van den Branden, G. [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Kalcheva, S [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Sikik, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium); Koonen, E [Belgian Nuclear Research Center (SCK-CEN), Mol (Belgium)

2016-09-01

BR2 is a research reactor used for radioisotope production and materials testing. It’s a tank-in-pool type reactor cooled by light water and moderated by beryllium and light water. The reactor core consists of a beryllium moderator forming a matrix of 79 hexagonal prisms in a hyperboloid configuration; each having a central bore that can contain a variety of different components such as a fuel assembly, a control or regulating rod, an experimental device, or a beryllium or aluminum plug. Based on a series of tests, the BR2 operation is currently limited to a maximum allowable heat flux of 470 W/cm² to ensure fuel plate integrity during steady-state operation and after a loss-of-flow/loss-of-pressure accident. A feasibility study for the conversion of the BR2 reactor from highly-enriched uranium (HEU) to low-enriched uranium (LEU) fuel was previously performed to verify it can operate safely at the same maximum nominal steady-state heat flux. An assessment was also performed to quantify the heat fluxes at which the onset of flow instability and critical heat flux occur for each fuel type. This document updates and expands these results for the current representative core configuration (assuming a fresh beryllium matrix) by evaluating the onset of nucleate boiling (ONB), onset of fully developed nucleate boiling (FDNB), onset of flow instability (OFI) and critical heat flux (CHF).

Fabrication of carbon-polymer composite bipolar plates for polymer electrolyte membrane fuel cells by compression moulding

International Nuclear Information System (INIS)

Raza, M.A.; Ahmed, R.; Saleem, A.; Din, R.U.

2009-01-01

Fuel cells are considered as one of the most important technologies to address the future energy and environmental pollution problems. These are the most promising power sources for road transportation and portable devices. A fuel cell is an electrochemical device that converts chemical energy into electrical energy. A fuel cell stack consists of bipolar plates and membrane electrode assemblies (MEA). The bipolar plate is by weight, volume and cost one of the most significant components of a fuel cell stack. Major functions of bipolar plates are to separate oxidant and fuel gas, provide flow channels, conduct electricity and provide heat transfer. Bipolar plates can be made from various materials including graphite, metals, carbon / carbon and carbon/ polymer composites. Materials for carbon-polymer composites are relatively inexpensive, less corrosive, strong and channels can be formed by means of a moulding process. Carbon-polymer composites are of two type i.e; thermosetting and thermoplastic. For thermosetting composite a bulk molding compound (BMC) was prepared by adding graphite, vinyl ester resin, methyl ethyl ketone peroxide and cobalt naphthalate. The BMC was thoroughly mixed, poured into a die mould of a bipolar plate with channels and hot pressed at a specific temperature and pressure. A bipolar plate was formed according to the die mould. Design of the mould is also discussed. Conducting polymers were also added to BMC to increase the conductivity of bipolar plates. Particle size of the graphite has also a significant effect on the conductivity of the bipolar plates. Thermoplastic composites were also prepared using polypropylene and graphite.

TRIGA Research Reactor Conversion to LEU and Modernization of Safety Related Systems

Energy Technology Data Exchange (ETDEWEB)

Sanda, R. M. [Institute for Nuclear Research Piteşti (SCN-Piteşti), Piteşti (Romania)

2014-08-15

The USA and IAEA proposed an international programme to reduce the enrichment of uranium in research reactors by converting nuclear fuel containing HEU into fuel containing 20% enriched uranium. The Government of Romania joined the programme and actively supported political, scientific, technical and economic actions that led to the conversion of the active area of the 14 MW TRIGA reactor at the Institute for Nuclear Research in Piteşti in May 2006. This confirmed the continuity of the Romanian Government’s non-proliferation policy and their active support of international cooperation. Conversion of the Piteşti research reactor was made possible by completion of milestones in the Research Agreement for Reactor Conversion, a contract signed with the US Department of Energy and Argonne National Laboratory. This agreement provided scientific and technical support and the possibility of delivery of all HEU TRIGA fuel to the United States. Additionally, about 65% of the fresh LEU fuel needed to start the conversion was delivered in the period 1992–1994. Furthermore, conversion was promoted through IAEA Technical Cooperation project ROM/4/024 project funded primarily by the United States that supported technical and scientific efforts and the delivery of the remaining required LEU nuclear fuel to complete the conversion. Nuclear fuel to complete the conversion was made by the French company CERCA with a tripartite contract among the IAEA, CERCA and Romania. The contract was funded by the US Department of Energy with a voluntary contribution by the Romanian Government. The contract stipulated manufacturing and delivery of LEU fuel by CERCA with compliance measures for quality, delivery schedule and safety requirements set by IAEA standards and Romanian legislation. The project was supported by the ongoing technical cooperation, safeguards, legal and procurement assistance of the IAEA, in particular its Department of Nuclear Safety. For Romanian research, the

Development of U6Fe-Al dispersions for the use of LEU in research and test reactors

International Nuclear Information System (INIS)

Nazare, S.

1983-01-01

For some time now, efforts are being made to develop fuel dispersions that would permit the use of low (approx. 20% 235-U) enriched uranium (LEU) instead of the currently used highly (approx. 93% 235-U) enriched uranium (HEU) in research and test reactors. Since penalties in the performance of the reactor have to be avoided, the 235-U content in the dispersion has at least to be retained at current levels. On account of their high U-densities, the major development effort has been focussed on the uranium silicides (U 3 Si, U 3 Si(Al), and U 3 Si 2 -based dispersions). With silicides as dispersants, it is possible to fabricate fuel element plates with U-densities in the dispersion of about 6.0 gU/cm 3 . In comparison to the silicides, the U 6 Fe-phase offers several advantages namely: higher U-density (approx. 17.0 gU/cm 3 ); relative ease of formation compared to U 3 Si; possible advantages with regard to reprocessing of the spent fuel due to the absence of silicon. The studies outlined here were performed with a view to investigating the preparation, reaction behavior and dimensional stability after heat treatment of U 6 Fe-Al dispersions

Development of U6Fe-Al dispersions for the use of LEU in research and test reactors

International Nuclear Information System (INIS)

Nazare, S.

1983-01-01

For some time now, efforts are being made to develop fuel dispersions that would permit the use of low (∼ 20% 235-U) enriched uranium (LEU) instead of the currently used highly (∼ 93% 235-U) enriched uranium (HEU) in research and test reactors. Since penalties in the performance of the reactor have to be avoided, the 235-U content in the dispersion has at least to be retained at current levels. On account of their high U-densities, the major development effort has been focussed on the uranium silicides [U 3 Si, U 3 Si(Al), and U 3 Si 2 - based dispersions. With silicides as dispersants, it is possible to fabricate fuel element plates with U-densities in the dispersion of about 6.0 g U/cm 3 . In comparison to the silicides, the U 6 Fe-phase offers several advantages namely: - higher U-density (∼ 17.0 g U/cm 3 ); - relative ease of formation compared to U 3 Si; - possible advantages with regard to reprocessing of the spent fuel due to the absence of silicon. The studies outlined here were therefore performed with a view to investigating the preparation, reaction behaviour and dimensional stability after heat treatment of U 6 Fe-Al dispersions

Corrosion of metal bipolar plates for PEM fuel cells: A review

Energy Technology Data Exchange (ETDEWEB)

Antunes, Renato A. [Engenharia de Materiais, Universidade Federal do ABC (UFABC), 09210-170 Santo Andre, SP (Brazil); Oliveira, Mara Cristina L.; Ett, Gerhard; Ett, Volkmar [Electrocell Ind. Com. Equip. Elet. LTDA, Centro de Inovacao, Empreendedorismo e Tecnologia (CIETEC), 05508-000 Sao Paulo, SP (Brazil)

2010-04-15

PEM fuel cells are of prime interest in transportation applications due to their relatively high efficiency and low pollutant emissions. Bipolar plates are the key components of these devices as they account for significant fractions of their weight and cost. Metallic materials have advantages over graphite-based ones because of their higher mechanical strength and better electrical conductivity. However, corrosion resistance is a major concern that remains to be solved as metals may develop oxide layers that increase electrical resistivity, thus lowering the fuel cell efficiency. This paper aims to present the main results found in recent literature about the corrosion performance of metallic bipolar plates. (author)

Design of metallic bipolar plates for PEM fuel cells.

Science.gov (United States)

2012-01-01

This project focused on the design and production of metallic bipolar plates for use in PEM fuel cells. Different metals were explored : and stainless steel was found out to be best suited to our purpose. Following the selection of metal, it was calc...

Some tooling for manufacturing research reactor fuel plates

International Nuclear Information System (INIS)

Knight, R.W.

1999-01-01

This paper will discuss some of the tooling necessary to manufacture aluminum-based research reactor fuel plates. Most of this tooling is intended for use in a high-production facility. Some of the tools shown have manufactured more than 150,000 pieces. The only maintenance has been sharpening. With careful design, tools can be made to accommodate the manufacture of several different fuel elements, thus, reducing tooling costs and maintaining tools that the operators are trained to use. An important feature is to design the tools using materials with good lasting quality. Good tools can increase return on investment. (author)

Some Tooling for Manufacturing Research Reactor Fuel Plates

International Nuclear Information System (INIS)

Knight, R.W.

1999-01-01

This paper will discuss some of the tooling necessary to manufacture aluminum-based research reactor fuel plates. Most of this tooling is intended for use in a high-production facility. Some of the tools shown have manufactured more than 150,000 pieces. The only maintenance has been sharpening. With careful design, tools can be made to accommodate the manufacture of several different fuel elements, thus, reducing tooling costs and maintaining tools that the operators are trained to use. An important feature is to design the tools using materials with good lasting quality. Good tools can increase return on investment

End plate for e.g. solid oxide fuel cell stack, sets thermal expansion coefficient of material to predetermined value

DEFF Research Database (Denmark)

2011-01-01

.05-0.3 mm. USE - End plate for solid oxide fuel cell stack (claimed). Can also be used in polymer electrolyte fuel cell stack and direct methanol fuel cell stack. ADVANTAGE - The robustness of the end plate is improved. The structure of the end plate is simplified. The risk of delamination of the stack...

Thermal-hydraulic calculations for KUHFR with reduced enrichment uranium fuel

International Nuclear Information System (INIS)

Mishima, Kaichiro; Shibata, Toshikazu.

1982-01-01

This report provides the preliminary results of the thermal-hydraulic calculations to study the safety aspects in fueling the KUHFR with reduced enrichment uranium. The calculations were based on what was outlined in the Safety Analysis Report for the KUHFR and the guidebook for research reactor core conversion, IAEA-TECDOC-233, published by the International Atomic Energy Agency. No significant differences in the thermal-hydraulic operating conditions have been found between HEU and MEU fuels. However, in LEU cases, the combination of three factors - larger power peaking with LEU fuel, smaller thermal conductivity of U 3 O 8 -Al fuel with high uranium densities, and thicker fuel meat - resulted in higher maximum fuel and surface temperatures with the LEU oxide fuel. (author)

Thermally induced dispersion mechanisms for aluminum-based plate-type fuels under rapid transient energy deposition

International Nuclear Information System (INIS)

Georgevich, V.; Taleyarkham, R.P.; Navarro-Valenti, S.; Kim, S.H.

1995-01-01

A thermally induced dispersion model was developed to analyze for dispersive potential and determine onset of fuel plate dispersion for Al-based research and test reactor fuels. Effect of rapid energy deposition in a fuel plate was simulated. Several data types for Al-based fuels tested in the Nuclear Safety Research Reactor in Japan and in the Transient Reactor Test in Idaho were reviewed. Analyses of experiments show that onset of fuel dispersion is linked to a sharp rise in predicted strain rate, which futher coincides with onset of Al vaporization. Analysis also shows that Al oxidation and exothermal chemical reaction between the fuel and Al can significantly affect the energy deposition characteristics, and therefore dispersion onset connected with Al vaporization, and affect onset of vaporization

Development of low enriched uranium target plates by thermo-mechanical processing of UAl2–Al matrix for production of 99Mo in Pakistan

International Nuclear Information System (INIS)

Ali, Kanwar Liaqat; Khan, Akhlaque Ahmad; Mushtaq, Ahmad; Imtiaz, Farhan; Ziai, Maratab Ali; Gulzar, Amir; Farooq, Muhammad; Hussain, Nazar; Ahmed, Nisar; Pervez, Shahid; Zaidi, Jamshed Hussain

2013-01-01

Uranium aluminide predominated with UAl 2 phase was prepared by arc-melting procedures and comminuted to required particle size. UAl 2 and Al powders were blended and compacted to achieve LEU fuel density of 2.17 g/cm 3 . The picture-frame technique was used to clad the dispersions (UAl 2 –Al) with aluminum. A few target plates were fabricated by thermo-mechanical processing (hot rolling and annealing) of UAl 2 –Al matrix contained in roll billet of Al. The fabricated plates were characterized by destructive and some of non-destructive testing techniques and then annealed to achieve required phase of uranium aluminide for proper dissolution in basic media

Postirradiation Examination Of U3O8-AL Plate Type Dispersion Fuel Element

International Nuclear Information System (INIS)

Nasution-Hasbullah; Sugondo; Amin, D.L.; Siti-Amini

1996-01-01

Postirradiation examination of plate type spent fuel element RIE-01 has been carried out in order to observer its physical changes and performance under irradiation in the reactor. The irradiation has been time more than two years with a declared burnup of 51.04 %. The examination included visual and dimensional measurement, measurement of burn-up distribution, wipe test and metallographic analysis. The results showed that all fuel plates retained their integrity. The colour changes were occurred on most of the plates significant suggesting that it was generated from the oxide layer formation. From gamma-scanning examination it could be deducted that the highest burn-up distribution of the plate was at position of 30 cm from the bottom. A more homogeneous distribution was found in the middle plate of the bundle. The increased plate thickness, as revealed by dimensional measurements as in agreement with the burn-up distribution pattern. Despite the changes observed in could be concluded that all changes occurred were still within the allowable limits and therefore it can recommended that an increase of the burn-up level above 51,04 % is still quite possible

Core physics analysis in support of the FNR HEU-LEU demonstration experiment

Energy Technology Data Exchange (ETDEWEB)

Losey, David C; Brown, Forrest B; Martin, William R; Lee, John C [Department of Nuclear Engineering, University of Michigan (United States)

1983-08-01

A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

Core physics analysis in support of the FNR HEU-LEU demonstration experiment

International Nuclear Information System (INIS)

Losey, David C.; Brown, Forrest B.; Martin, William R.; Lee, John C.

1983-01-01

A core neutronics analysis has been undertaken to assess the impact of low-enrichment fuel on the performance and utilization of the FNR As part of this analytic effort a computer code system has been assembled which will be of general use in analyzing research reactors with MTR-type fuel. The code system has been extensively tested and verified in calculations for the present high enrichment core. The analysis presented here compares the high-and-low enrichment fuels in batch and equilibrium core configurations which model the actual FNR operating conditions. The two fuels are compared for cycle length, fuel burnup, and flux and power distributions, as well as for the reactivity effects which are important in assessing the impact of LEU fuel on reactor shutdown margin. (author)

Corrosion-resistant, electrically-conductive plate for use in a fuel cell stack

Science.gov (United States)

Carter, J David [Bolingbrook, IL; Mawdsley, Jennifer R [Woodridge, IL; Niyogi, Suhas [Woodridge, IL; Wang, Xiaoping [Naperville, IL; Cruse, Terry [Lisle, IL; Santos, Lilia [Lombard, IL

2010-04-20

A corrosion resistant, electrically-conductive, durable plate at least partially coated with an anchor coating and a corrosion resistant coating. The corrosion resistant coating made of at least a polymer and a plurality of corrosion resistant particles each having a surface area between about 1-20 m.sup.2/g and a diameter less than about 10 microns. Preferably, the plate is used as a bipolar plate in a proton exchange membrane (PEMFC) fuel cell stack.

Research reactor fuel - an update

International Nuclear Information System (INIS)

Finlay, M.R.; Ripley, M.I.

2003-01-01

In the two years since the last ANA conference there have been marked changes in the research reactor fuel scene. A new low-enriched uranium (LEU) fuel, 'monolithic' uranium molybdenum, has shown such promise in initial trials that it may be suitable to meet the objectives of the Joint Declaration signed by Presidents Bush and Putin to commit to converting all US and Russian research reactors to LEU by 2012. Development of more conventional aluminium dispersion UMo LEU fuel has continued in the meantime and is entering the final qualification stage of multiple full sized element irradiations. Despite this progress, the original 2005 timetable for UMo fuel qualification has slipped and research reactors, including the RRR, may not convert from silicide to UMo fuel before 2007. The operators of the Swedish R2 reactor have been forced to pursue the direct route of qualifying a UMo lead test assembly (LTA) in order to meet spent fuel disposal requirements of the Swedish law. The LTA has recently been fabricated and is expected to be loaded shortly into the R2 reactor. We present an update of our previous ANA paper and details of the qualification process for UMo fuel

Prediction for the flow distribution and the pressure drop of a plate type fuel assembly

International Nuclear Information System (INIS)

Park, Jong Hark; Jo, Dea Sung; Chae, Hee Taek; Lee, Byung Chul

2011-01-01

A plate type fuel assembly widely used in many research reactors does not allow the coolant to mix with neighboring fuel channels due to the completely separated flow channels. If there is a serious inequality of coolant distribution among channels, it can reduce thermal-hydraulic safety margin, as well as it can cause a deformation of fuel plates by the pressure difference between neighboring channels, thus the flow uniformity in the fuel assembly should be confirmed. When designing a primary cooling system (PCS), the pressure drop through a reactor core is a dominant value to determine the PCS pump size. The major portion of reactor core pressure drop is caused by the fuel assemblies. However it is not easy to get a reasonable estimation of pressure drop due to the geometric complexity of the fuel assembly and the thin gaps between fuel assemblies. The flow rate through the gap is important part to determine the total flow rate of PCS, so it should be estimated as reasonable as possible. It requires complex and difficult jobs to get useful data. In this study CFD analysis to predict the flow distribution and the pressure drop were conducted on the plate type fuel assembly, which results would be used to be preliminary data to determine the PCS flow rate and to improve the design of a fuel assembly

Corrosion and pyrophoricity of ZPPR fuel plates: Implications for basin storage

International Nuclear Information System (INIS)

Totemeier, T.C.; Hayes, S.L.; Pahl, R.G.; Crawford, D.C.

1997-01-01

This paper presents the results of recent experimentation and analysis of the pyrophoric behavior of corroded Zero Power Physics Reactor (ZPPR) HEU fuel plates and the implications of these results for the handling, drying, and passivation of uranium metal fuels stored in water basins. The ZPPR plates were originally clad in 1980; crevice corrosion of the uranium metal in a dry storage environment has occurred due to the use of porous cladding end plugs. The extensive corrosion has resulted in bulging and, in some cases, breaching of the cladding over a 15 year storage period. Processing of the plates has been initiated to recover the highly enriched uranium metal and remove the storage vulnerability identified with the corroded plates, which have been shown to contain significant quantities of the pyrophoric compound uranium hydride (UH 3 ). Experiments were undertaken to determine effective passivation techniques for the corrosion product; analysis and modeling was performed to determine whether heat generated by rapid hydride re-oxidation could ignite the underlying metal plates. The results of the initial passivation experiment showed that simple exposure of the hydride-containing corrosion product to an Ar-3 vol.% O 2 environment was insufficient to fully passivate the hydride--flare-up of the product occurred during subsequent vigorous handling in air. A second experiment demonstrated that corrosion product was fully stable following grinding of the product to a fine powder in the Ar-3 vol.% O 2 atmosphere. Numerical modeling of a corroded plate indicated that ignition of the plate due to the heat from hydride re-oxidation was likely if hydride fractions in the corrosion product exceeded 30%

Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

Energy Technology Data Exchange (ETDEWEB)

Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

Tensile Test of Welding Joint Parts for a Plate-type Fuel Assembly

International Nuclear Information System (INIS)

Yoon, K. H.; Kim, J. Y.; Kim, H. J.; Yim, J. S.

2013-01-01

The tensile tests were performed using an INSTRON 4505 (universal tensile) testing machine. These welding joints are composed of two parts for the soundness of the fuel assembly; one is the side plate with a fixing bar and the other is a side plate with an end fitting. These two joint parts are fabricated by TIG welding method. The tensile tests of the welding joints of a plate-type FA are executed by a tensile test. The fixture configurations for the specimen are very important to obtain the strict test results. The maximum strength has an approximately linear correlation with the unit bonding length of the welding joints. In spite of these results, the maximum strengths of the welding joints are satisfied according to the minimum requirement. These tensile tests of the joint parts for a plate-type fuel assembly (FA) have to be executed to evaluate the structural strength. For the tensile test, the joint parts of a FA used in the test are made of aluminum alloy (Al6061-T6)

Comparison of MCNP calculations against measurements in moderator temperature experiments with CANFLEX-LEU in ZED-2

International Nuclear Information System (INIS)

Watts, D.G.; Adams, F.P.; Zeller, M.B.; Bromley, B.P.

2008-01-01

This paper summarizes sample calculations of MCNP5 compared against measurements of moderator temperature coefficient experiments in the ZED-2 critical facility with CANFLEX-LEU fuel. MCNP5 is tested for key parameters associated with various reactor physics phenomena of interest for CANDU/ACR-1000) reactors, including reactivity changes with coolant density, moderator density, and moderator temperature, and also normalized flux distributions. The experimental data for these comparisons were obtained from critical experiments in AECL's ZED-2 critical facility using CANFLEX-LEU fuel in a 24-cm square lattice pitch. These comparisons establish biases/uncertainties in the calculation of k-eff, coolant void reactivity, and moderator temperature coefficient of reactivity. Results show very little bias in the moderator temperature coefficient of reactivity, and very good agreement in the calculation of normalized flux distributions. (author)

New high density MTR fuel. The CEA-CERCA-COGEMA development program

International Nuclear Information System (INIS)

Languille, A.; Durand, J.P.; Gay, A.

1999-01-01

The development of a new generation of LEU, high in density and with reprocessing capacities MTR fuel, is a key issue to provide reactor operators with a smooth operation which is necessary for a long term development of Nuclear Energy. In the RRFM'98 meeting, a joint contribution of CEA, CERCA and COGEMA presented a technical classification of the potential candidates uranium alloys. In this paper this MTR working group presents the development program of a new high density fuel. This program is composed of three main steps: Basic Data analysis and collection, Plate Tests (Irradiation and Post Irradiation Examinations) and Lead Test Assemblies (Irradiation and Post Irradiation Examinations). The goal to be reached is to make this new fuel available before the end of the present US return policy. (author)

Thermal Characteristic Of AIMg2 Cladding And Fuel Plates Of U3Si2-Al With Various Uranium Loading

International Nuclear Information System (INIS)

Aslina, Br. G.; Suparjo; Aggraini, D.; Hasbullah, N.

1998-01-01

Thermal characteristic analyzed in this paper included linear expansion value, coefficient expansion, and enthalpy of cladding material fuel core and fuel plate of U 3 Si 2 -AI. Before analyzing, the fresh cladding of AIMg2 (without treatment) and the rolled AIMg2 were annealed at temperature of 425 o C for 1 hour, and the fuel plates of U 3 Si 2 -AI was prepared for various uranium loading of 0.9 - 3.6 - 4.2 - 4.8 and 5.2 g/cm 3 . Linear expansion nominal value and expansion coefficient were analyzed by using Dilatometer whereas enthalpy determination used Differential Thermal Analysis (DTA). The linear expansion and expansion coefficient analysis was performed to study the dimension cladding and of fuel plates during their stay in the reactor core, whereas determination of enthalpy was carried out to estimate the energy absorbed and released by fuel meat of U 3 Si 2 -AI to the cooling water through AlMg2 as a cladding. The result showed that the linear expansion and expansion coefficient of fresh AIMg2 cladding, rolled AIMg2 and fuel plates of U 3 Si 2 -AI are increased with the increase of temperature as well as the increase of uranium loading. The enthalpy measure showed that the enthalpy of fresh AIMg2 is smaller than that of rolled AIMg2 but melting temperature of fresh AIMg2 is greater than that of rolled AIMg2. The enthalpy of fuel plates and meat of U 3 Si 2 -AI is less than that of plates of U 3 Si 2 -AI. The enthalpy of fuel platers and meat of U 3 Si 2 -AI decrease with the increase of uranium loading. It is concluded that the fuel meat more reactive than fuel plates of U 3 Si 2 -AI

Study of the residual porosity in fuel plate cores based on U3O8 - Al dispersions

International Nuclear Information System (INIS)

Durazzo, M.

2005-01-01

The residual porosity in the meat of nuclear dispersion fuel plates, the fabrication voids, explains the corrosion behaviour of the meats when exposed to the water used as coolant and moderator of MTR type research reactors. The fabrication voids also explain variations in irradiation performance of many fuel dispersion for nuclear reactors. To obtain improved corrosion and irradiation performance, we must understand the fabrication factors that control the amount of void volume in fuel plate meats. The purpose of this study was to investigate the void content of aluminum-base dispersion-type U 3 O 8 -Al fuel plates depending on the characteristics of the starting fuel dispersion used to produce the fuel meat, which is fabricated by pressing. The void content depends on the U 3 O 8 concentration. For a particular U 3 O 8 content, the rolling process establishes a constant void concentration, which is called equilibrium porosity. The equilibrium quantity of voids is insensitive to the initial density of the fuel compact. (author)

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

Energy Technology Data Exchange (ETDEWEB)

Montierth, Leland M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-07-19

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

Criticality safety evaluation for the Advanced Test Reactor enhanced low enriched uranium fuel elements

International Nuclear Information System (INIS)

Montierth, Leland M.

2016-01-01

The Global Threat Reduction Initiative (GTRI) convert program is developing a high uranium density fuel based on a low enriched uranium (LEU) uranium-molybdenum alloy. Testing of prototypic GTRI fuel elements is necessary to demonstrate integrated fuel performance behavior and scale-up of fabrication techniques. GTRI Enhanced LEU Fuel (ELF) elements based on the ATR-Standard Size elements (all plates fueled) are to be fabricated for testing in the Advanced Test Reactor (ATR). While a specific ELF element design will eventually be provided for detailed analyses and in-core testing, this criticality safety evaluation (CSE) is intended to evaluate a hypothetical ELF element design for criticality safety purposes. Existing criticality analyses have analyzed Standard (HEU) ATR elements from which controls have been derived. This CSE documents analysis that determines the reactivity of the hypothetical ELF fuel elements relative to HEU ATR elements and whether the existing HEU ATR element controls bound the ELF element. The initial calculations presented in this CSE analyzed the original ELF design, now referred to as Mod 0.1. In addition, as part of a fuel meat thickness optimization effort for reactor performance, other designs have been evaluated. As of early 2014 the most current conceptual designs are Mk1A and Mk1B, that were previously referred to as conceptual designs Mod 0.10 and Mod 0.11, respectively. Revision 1 evaluates the reactivity of the ATR HEU Mark IV elements for a comparison with the Mark VII elements.

Assessment of the effectiveness of the LEU Reform Rule and its implementation

International Nuclear Information System (INIS)

Moran, B.W.; Nations, J.O.; Hammond, G.A.

1993-11-01

The US Nuclear Regulatory Commission (NRC) amended its material control and accounting (MC ampersand A) requirements in 1985 for licensees possessing and using special nuclear material (SNM) of low strategic significance in quantities larger than one effective kilogram (kg). The goal of the Low-Enriched Uranium (LEU) Reform Rule (i.e., 10CFR 74.31) was to establish MC ampersand A requirements for the LEU licensees at a level consistent with the safeguards risk associated with the relatively low strategic importance of such material. The amended requirements were written in a performance-oriented manner, rather than a prescriptive one, in an effort to allow the licensees the opportunity to choose the most cost-effective means of satisfying the requirements. The LEU Reform Rule was implemented in January 1988 and the fuel cycle facilities have had sufficient experience in implementing the rule to allow a meaningful review of its effectiveness. This document provides technical analysis and recommendations to assist the NRC in making a determination if the rule is achieving its intended purpose, and if not, to make the necessary changes to accomplish this

Core conversion from rod to plate type fuel elements in research reactors

International Nuclear Information System (INIS)

Khattab, M.S.; Mina, A.R.

1997-01-01

Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

International light water nuclear fuel fabrication supply. Are fabrication services assured?

International Nuclear Information System (INIS)

Rothwell, Geoffrey

2010-01-01

This paper examines the cost structure of fabricating light water reactor (LWR) fuel with low-enriched uranium (LEU, with less than 5% enrichment). The LWR-LEU fuel industry is decades old, and (except for the high entry cost of designing and licensing a fuel fabrication facility and its fuel), labor and additional fabrication lines can be added at Nth-of-a-Kind cost to the maximum capacity allowed by a site license. The industry appears to be competitive: nuclear fuel fabrication capacity is assured with many competitors and reasonable prices. However, nuclear fuel assurance has become an important issue for nations now to considering new nuclear power plants. To provide this assurance many proposals equate 'nuclear fuel banks' (which would require fuel for specific reactors) with 'LEU banks' (where LEU could be blended into nuclear fuel with the proper enrichment) with local fuel fabrication. The policy issues (which are presented, but not answered in this paper) become (1) whether the construction of new nuclear fuel fabrication facilities in new nuclear power nations could lead to the proliferation of nuclear weapons, and (2) whether nuclear fuel quality can be guaranteed under current industry arrangements, given that fuel failure at one reactor can lead to forced shutdowns at many others. (author)

Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Ilas, Germina [ORNL; Primm, Trent [ORNL

2011-05-01

An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

Energy Technology Data Exchange (ETDEWEB)

Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

2006-01-01

The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

Low enrichment fuel conversion for Iowa State University

International Nuclear Information System (INIS)

Rohach, A.F.; Hendrickson, R.A.

1991-08-01

Work during the reported period was centered primarily in preparation for receiving the LEU fuel and the shipping of the HEU fuel. This included development of procedures and tools for the disassembly process. During the period we held many practice sessions applying these tools and practices to a dummy fuel assembly. The LEU fuel was received on April 10, 1991 and the reactor was shut down on May 3, 1991 for refueling. The twelve HEU fuel assemblies in the UTR-10 reactor core were removed and disassembled during the week of May 6--9, 1991. The disassembly process went smoothly with only a few minor problems. Also during this reporting period several experimental measurements and preventative maintenance tasks were accomplished. Finally procedures and practices have been developed for the new LEU fuel loading and critical experiments which are to be completed during the late summer of 1991

Design of experiment study of the parameters that affect performance of three flow plate configurations of a proton exchange membrane fuel cell

International Nuclear Information System (INIS)

Carton, J.G.; Olabi, A.G.

2010-01-01

Low temperature hydrogen fuel cells are electrochemical devices which offer a promising alternative to traditional power sources. Fuel cells produce electricity with a reaction of the fuel (hydrogen) and air. Fuel cells have the advantage of being clean; only producing water and heat as by products. The efficiency of a fuel cell varies depending on the type; SOFC with CHP for example, can have a system efficiency of up to 65%. What the Authors present here is a comparison between three different configurations of flow plates of a proton exchange membrane fuel cell, the manufacturer's serpentine flow plate and two new configurations; the maze flow plate and the parallel flow plate. A study of the input parameters affecting output responses of voltage, current, power and efficiency of a fuel cell is performed through experimentation. The results were taken from direct readings of the fuel cell and from polarisation curves produced. This information was then analysed through a design of experiment to investigate the effects of the changing parameters on different configurations of the fuel cell's flow plates. The results indicate that, in relation to current and voltage response of the polarisation curve and the corresponding graphs produced from the DOE, the serpentine flow plate design is a much more effective design than the maze or parallel flow plate design. It was noted that the parallel flow plate performed reasonably well at higher pressures but over all statically the serpentine flow plate performed better.

A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

International Nuclear Information System (INIS)

Sohn, Hee Dong

2012-02-01

In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

A study on the effect of stainless steel plate position on neutron multiplication factor in spent fuel storage racks

Energy Technology Data Exchange (ETDEWEB)

Sohn, Hee Dong

2012-02-15

In spent fuel storage racks, which are just composed of stainless steel plates without neutron absorbing materials, neutron multiplication factors are investigated as the variation of the water gap that exists between the fuel assembly and the stainless steel plates. The stainless steel plate has a low moderating power compared with water because it has a lower elastic scattering cross section, as well as far less change of lethargy in an elastic collision than water. Thus, if stainless steel plates are installed around the fuel assembly instead of water, it is hard for neutrons to be thermalized properly. Therefore, the neutron multiplication factor can be decreased because the thermal neutron fluence and the total neutron production rate in fuel rods are decreased. A stainless steel plate has also has a thermal neutron absorption cross section. Thus, it can absorb thermal neutrons around the fuel assembly. The dominant factor which can cause a decrease in the neutron multiplication factor is the interruption of neutron moderation by stainless steel plates. Therefore, the neutron multiplication factor should always be kept at its lowest point, if stainless steel plates are installed on the specific position where interruptions of the neutron moderation occur most often, allowing for thermal neutrons to be absorbed. The stainless steel plate position is 7 mm away from the outermost surface of the fuel assembly with a pitch of 280mm. The specific position appearing the lowest neutron multiplication factor as the pitch variation from 260mm to 290mm with 10mm interval is also investigated. The lowest neutron multiplication factor also occurs 7mm or 8mm away from the outermost surface of the fuel assembly

Circular arc fuel plate stability experiments and analyses for the advanced neutron source

International Nuclear Information System (INIS)

Swinson, W.F.; Battiste, R.L.; Yahr, G.T.

1995-08-01

The thin fuel plates planned for the Advanced Neutron Source are to be cooled by forcing heavy water at high velocity, 25 m/s, through thin cooling channels on each side of each plate. Because the potential for structural failure of the plates is a design concern, considerable effort has been expended in assessing this potential. As part of this effort, experimental flow tests and analyses to evaluate the structural response of circular arc plates have been conducted, and the results are given in this report

Non-destructive evaluation methods to improve quality control in low enrichment MTR fuel plate production

International Nuclear Information System (INIS)

Milne, J.M.; Lidington, B.; Hawker, B.M.

1991-01-01

This paper summarises some preliminary non-destructive measurements made recently at the Harwell Laboratory on a prototype low enrichment MTR fuel plate. The measurements were intended to indicate the potential of two different techniques for improving quality control in plate production. Pulse Video Thermography (PVT) is being considered as an alternative to ultrasound transmission measurements for the detection and sizing of lack of thermal bonding between the fuel and the clad layers, either to verify the indications from the established ultrasonic methods before destroying the plate or as a replacement method of inspection. High frequency pulse-echo ultrasonics is being considered for providing maps of clad layer thickness on each side of the plate. The measurements have indicated the potential for both methods, but more work is required, using a test plate containing controlled defects, to establish their capability. (orig.)

The velocity measurement by LDV at the simulated plate fuel assembly

International Nuclear Information System (INIS)

Tae Sung Ha

2001-01-01

For a more accurate safety analysis for McMaster Nuclear Reactor (MNR), local velocity measurements in a mock-up of the 18-plate fuel assembly are conducted over the range of M=2.0kg/s to 5.0kg/s (u=0.59m/s to 1.48m/s). To enable the measurement of the mass flow distribution through the channels by Laser Doppler Velocimeter(LDV), the curved fuel plate assembly is modified to flat fuel plates. The experimental result shows that the velocity profile is fairly symmetric for the 1st channel to the 17th subchannel at its center. The velocity in the peripheral area is slightly decreased while that directly above the circular pipe is correspondingly increased due to the effect of blockage by the exit endfitting. The mass flow rate fraction is fairly well distributed from the 1st to the 9th channels; at the outmost channels (1st and 3rd subchannels) the flow is approximately 95-97% of the average channel flow and at the central channels (4th and 8th subchannels) the flow is about 102-105% of the average channel mass flow rate. It is shown that the measured mass flow distribution is consistent with the results of the numerical calculation except 1st and 17th channels. (author)

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

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.

1980-12-01

A common plate-type fuel for research and test reactors is U/sub 3/O/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 nonpeaceful 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.

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

Energy Technology Data Exchange (ETDEWEB)

Dunn, Floyd E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States). Nuclear Reactor Lab.; Wilson, Erik [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-01

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.

Determination of Dancoff correction thermal utilization and thermal disadvantage factors of HEU and LEU cores of an MNSR

International Nuclear Information System (INIS)

Ofori, Y. T.

2013-07-01

Ghana Research Reactor-1 (GHARR-1), an MNSR (Miniature Neutron Source Reactor) is to be converted from HEU (Highly Enriched Uranium) to LEU (Low Enriched Uranium) fuel, along with all current MNSRs in various other countries. The purpose of the conversion is to minimize the use of HEU for non-proliferation of high-grade nuclear fuel. In this research work, a comparative study has been performed for the determination of the Dancoff, thermal utilization and thermal disadvantage factors of highly enriched uranium (HEU) and potential low enriched uranium (LEU) cores of GHARR-1. A one group transport theory and collision probability based methodologies was used to develop mathematical formulations for thermal utilization factor and thermal disadvantage factor assuming isotropic scattering. This methodology was implemented in a FORTRAN 95 based computer program THERMCALC, which uses Bessell and BesselK as subroutines developed to calculate the modified Bessel functions I n and K n respectively using the polynomial approximation method. Furthermore, a Dancoff correction factor of 0.1519 thermal utilization factor of 0.9767 and a thermal disadvantage factor of 1.894 were obtained for the 90.2% highly enriched Uranium core of GHARR-1. The results compare favorably with literature. Thus THERMCALC can be used as a reliable tool for the calculation of Dancoff, thermal utilization and disadvantage factors of MNSR cores. Other potential LEU cores; UO 2 (with different fuel meat densities and enrichments) and U 3 Si 2 have also been analysed. UO 2 with 12.6% of Uranium-235 was chosen as the most potential LEU core for the GHARR-1. (au)

A neutronics feasibility study for the LEU conversion of Poland's Maria research reactor

International Nuclear Information System (INIS)

Bretscher, M. M.

1998-01-01

The MARIA reactor is a high-flux multipurpose research reactor which is water-cooled and moderated with both beryllium and water. Standard HEU (80% 235 U)fuel assemblies consist of six concentric fuel tubes of a U-Al alloy clad in aluminum. Although the inventory of HEU (80%) fuel is nearly exhausted, a supply of highly-loaded 36%-enriched fuel assemblies is available at the reactor site. Neutronic equilibrium studies have been made to determine the relative performance of fuels with enrichments of 80%, 36% and 19.7%. These studies indicate that LEU (19.7%) densities of about 2.5 gU/cm 3 and 3.8 gU/cm 3 are required to match the performance of the MARIA reactor with 80%-enriched and with 36%-enriched fuels, respectively

Recent status and future aspect of plate type fuel element technology with high uranium density at NUKEM

International Nuclear Information System (INIS)

Hrovat, M.F.; Hassel, H.-W.

1983-01-01

According to the present state of development full size test fuel elements with UAl x , U 3 O 8 , and U 3 Si 2 fuel were fabricated at Nukem in production scale. The maximum uranium densities amount to 1.8 g/cc for UAI x , 2.9 g/cc for U 3 O 8 , and 4.76 g/cc for U 3 Si 2 . The irradiation performance of these fuel elements is good: Up to the end of September 1982 the following burnups were achieved: 73% with UA1 x , 60% with U 3 O 8 , 39% with U 3 Si 2 ; no defects could be detected. For an economical fuel element production with reduced 235-U enrichment chemical uranium recycling methods were developed allowing immediate scrap recovery at minimum waste generation. In addition test plates with UAl x and U 3 O 8 fuel were successfully irradiated in the ORR up to a burnup of 75 %. The relatively high uranium meat densities of these test plates amount to 2.2 g/cc for UAI x , and 3.14 g/cc for U 3 O 8 fuel. Apart from plates with standard geometry also plates with increased meat thickness were inserted. (author)

Temperature feedback of TRIGA MARK-II fuel

Science.gov (United States)

Usang, M. D.; Minhat, M. S.; Rabir, M. H.; M. Rawi M., Z.

2016-01-01

We study the amount of temperature feedback on reactivity for the three types of TRIGA fuel i.. ST8, ST12 and LEU fuel, are used in the TRIGA MARK II reactor in Malaysia Nuclear Agency. We employ WIMSD-5B for the calculation of kin f for a single TRIGA fuel surrounded by water. Typical calculations of TRIGA fuel reactivity are usually limited to ST8 fuel, but in this paper our investigation extends to ST12 and LEU fuel. We look at the kin f of our model at various fuel temperatures and calculate the amount reactivity removed. In one instance, the water temperature is kept at room temperature of 300K to simulate sudden reactivity increase from startup. In another instance, we simulate the sudden temperature increase during normal operation where the water temperature is approximately 320K while observing the kin f at various fuel temperatures. For accidents, two cases are simulated. The first case is for water temperature at 370K and the other is without any water. We observe that the higher Uranium content fuel such as the ST12 and LEU have much smaller contribution to the reactivity in comparison to the often studied ST8 fuel. In fact the negative reactivity coefficient for LEU fuel at high temperature in water is only slightly larger to the negative reactivity coefficient for ST8 fuel in void. The performance of ST8 fuel in terms of negative reactivity coefficient is cut almost by half when it is in void. These results are essential in the safety evaluation of the reactor and should be carefully considered when choices of fuel for core reconfiguration are made.

Development of the uranium recovery process from rejected fuel plates in the fabrication of MTR type nuclear fuel

International Nuclear Information System (INIS)

Fleming Rubio, Peter Alex

2010-01-01

The current work was made in Conversion laboratory belonging to Chilean Nuclear Energy Commission, CCHEN. This is constituted by the development of three hydrometallurgical processes, belonging to the recovery of uranium from fuel plates based on uranium silicide (U_3Si_2) process, for nuclear research reactors MTR (Material Testing Reactor) type, those that come from the Fuel Elements Manufacture Plant, PEC. In the manufacturing process some of these plates are subjected to destructive tests by quality requirement or others are rejected for non-compliance with technical specifications, such as: lack of homogenization of the dispersion of uraniferous compound in the meat, as well as the appearance of the defects, such as blisters, so-called "dog bone", "fish tail", "remote islands", among others. Because the uranium used is enriched in 19.75% U_2_3_5 isotope, which explains the high value in the market, it must be recovered for reuse, returning to the production line of fuel elements. The uranium silicide, contained in the plates, is dispersed in an aluminum matrix and covered with plates and frames of ASTM 6061 Aluminum, as a sandwich coating, commonly referred to as 'meat' (sandwich meat). As aluminum is the main impurity, the process begins with this metal dissolution, present in meat and plates, by NaOH reaction, followed by a vacuum filtration, washing and drying, obtaining a powder of uranium silicide, with a small impurities percentage. Then, the crude uranium silicide reacts with a solution of hydrofluoric acid, dissolving the silicon and simultaneously precipitating UF_4 by reaction with HNO_3, obtaining an impure UO_2(NO_3)_2 solution. The experimental work was developed and implemented at laboratory scale for the three stages pertaining to the uranium recovery process, determining for each one the optimum operation conditions: temperature, molarity or concentration, reagent excess, among others (author)

Continuing investigations for technology assessment of 99Mo production from LEU [low enriched uranium] targets

International Nuclear Information System (INIS)

Vandegrift, G.F.; Kwok, J.D.; Marshall, S.L.; Vissers, D.R.; Matos, J.E.

1987-01-01

Currently much of the world's supply of 99m Tc for medical purposes is produced from 99 Mo derived from the fissioning of high enriched uranium (HEU). This paper presents the results of our continuing studies on the effects of substituting low enriched uranium (LEU) for HEU in targets for the production of fission product 99 Mo. Improvements in the electrodeposition of thin films of uranium metal continue to increase the appeal for the substitution of LEU metal for HEU oxide films in cylindrical targets. The process is effective for targets fabricated from stainless steel or zircaloy. Included is a cost estimate for setting up the necessary equipment to electrodeposit uranium metal on cylindrical targets. Further investigations on the effect of LEU substitution on processing of these targets are also reported. Substitution of uranium silicides for the uranium-aluminium alloy or uranium aluminide dispersed fuel used in current target designs will allow the substitution of LEU for HEU in these targets with equivalent 99 Mo-yield per target and no change in target geometries. However, this substitution will require modifications in current processing steps due to 1) the insolubility of uranium silicides in alkaline solutions and 2) the presence of significant quantities of silicate in solution. Results to date suggest that substitution of LEU for HEU can be achieved. (Author)

Tensile mechanical properties of U3Si2-Al fuel plate

International Nuclear Information System (INIS)

Xu Yong; Hu Huawei; Zhuang Hongquan; Wang Xishu

2003-01-01

The fuel plate made of fuel meat, with the U 3 Si 2 -Al dispersion fuel center, and 6061 Al alloy cladding, is a new kind of fuel used in research reactors. The mechanical property data of the fuel meat is the basic data in the design of fuel group, but the mechanical property of this fuel meat has not been studied all over the world till now. In this paper, the mechanical properties of U 3 Si 2 -Al fuel meats of different sizes used in research reactors are investigated and analyzed, and at the same time the carrying capacity of tensile in different directions are also compared. In order to get more knowledge about the mechanical properties of the fuel meat, the tensile experiment has been carried out repeatedly. Considering the lower ratio of elongation and the brittleness, the microscope has been used to examine the zone of fracture after tensile test. (authors)

High density dispersion fuel

International Nuclear Information System (INIS)

Hofman, G.L.

1996-01-01

A fuel development campaign that results in an aluminum plate-type fuel of unlimited LEU burnup capability with an uranium loading of 9 grams per cm 3 of meat should be considered an unqualified success. The current worldwide approved and accepted highest loading is 4.8 g cm -3 with U 3 Si 2 as fuel. High-density uranium compounds offer no real density advantage over U 3 Si 2 and have less desirable fabrication and performance characteristics as well. Of the higher-density compounds, U 3 Si has approximately a 30% higher uranium density but the density of the U 6 X compounds would yield the factor 1.5 needed to achieve 9 g cm -3 uranium loading. Unfortunately, irradiation tests proved these peritectic compounds have poor swelling behavior. It is for this reason that the authors are turning to uranium alloys. The reason pure uranium was not seriously considered as a dispersion fuel is mainly due to its high rate of growth and swelling at low temperatures. This problem was solved at least for relatively low burnup application in non-dispersion fuel elements with small additions of Si, Fe, and Al. This so called adjusted uranium has nearly the same density as pure α-uranium and it seems prudent to reconsider this alloy as a dispersant. Further modifications of uranium metal to achieve higher burnup swelling stability involve stabilization of the cubic γ phase at low temperatures where normally α phase exists. Several low neutron capture cross section elements such as Zr, Nb, Ti and Mo accomplish this in various degrees. The challenge is to produce a suitable form of fuel powder and develop a plate fabrication procedure, as well as obtain high burnup capability through irradiation testing

Technical report: technical development on the silicide plate-type fuel experiment at nuclear safety research reactor

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Soyama, Kazuhiko; Ichikawa, Hiroki

1991-08-01

According to a reduction of fuel enrichment from 45 w/o 235 U to 20 w/o, an aluminide plate-type fuel used currently in the domestic research and material testing reactors will be replaced by a silicide plate-type one. One of the major concern arisen from this alternation is to understand the fuel behavior under simulated reactivity initiated accident (RIA) conditions, this is strongly necessary from the safety and licensing point of view. The in-core RIA experiments are, therefore, carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute (JAERI). The silicide plate-type fuel consisted of the ternary alloy of U-Al-Si as a meat with uranium density up to 4.8 g/cm 3 having thickness by 0.51 mm and the binary alloy of Al-3%Mg as a cladding by thickness of 0.38 mm. Comparison of the physical properties of this metallic plate fuel with the UO 2 -zircaloy fuel rod used conventionally in commercial light water reactors shows that the heat conductivity of the former is of the order of about 13 times greater than the latter, however the melting temperature is only one-half (1570degC). Prior to in-core RIA experiments, there were some difficulties lay in our technical path. This report summarized the technical achievements obtained through our four years work. (J.P.N.)

Fabrication and testing of U–7Mo monolithic plate fuel with Zircaloy cladding

Energy Technology Data Exchange (ETDEWEB)

Pasqualini, E.E. [Laboratorio de Nanotecnología Nuclear, Centro Atómico Constituyentes, Comisión Nacional de Energía Atómica, Av. General Paz 1499, B1650KNA, San Martín, Prov. Buenos Aires (Argentina); Robinson, A.B. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Porter, D.L., E-mail: Douglas.Porter@inl.gov [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Wachs, D.M. [Idaho National Laboratory, P. O. Box 1625, Idaho Falls, ID, 83415-6188 (United States); Finlay, M.R. [Australian Nuclear Science and Technology Organisation, PMB 1, Menai, NSW, 2234 (Australia)

2016-10-15

Nuclear fuel designs are being developed to replace highly enriched fuel used in research and test reactors with fuels of low enrichment. In the most challenging cases, U–(7–10 wt%)Mo monolithic plate fuels are proposed. One of the considered designs includes aluminum-alloy cladding, which provides some challenges in fabrication and fuel/cladding interaction during service. Zircaloy cladding, specifically Zry–4, was investigated as an alternative cladding, and development of a fabrication method was performed by researchers with the Comisión Nacionalde Energia Atómica (CNEA) in Argentina, resulting in test fuel plates (Zry–4 clad U–7Mo) which were subsequently tested in the Advanced Test Reactor in Idaho. Because Zry–4 and U–(7–10)Mo have similar high-temperature mechanical properties, fabrication was simplified in that the fuel foil and cladding could be co-rolled and bonded. The challenge was to prevent a thermal-expansion mismatch, which could destroy the fuel/cladding bond before complete bonding was achieved; the solution was to prevent the composites from cooling significantly during or between roll passes. The final product performed very well in-reactor, showing good bonding, very little fuel/cladding interaction—either from fabrication or in-reactor testing—and little swelling, especially no detectable heterogeneous bubble formation at the fuel/cladding interface tested to a fission density of up to 2.7E+21 (average) fissions/cm{sup 3}, 3.8E+21 (peak).

Experience on wet storage spent fuel sipping at IEA-R1 Brazilian research reactor

International Nuclear Information System (INIS)

Perrotta, J.A.; Terremoto, L.A.A.; Zeituni, C.A.