Abstract: Recent research has appointed the need to replace the classic fuel concept, used in light water reactors. Uranium dioxide has a weak point due to the low thermal conductivity, that produce high temperatures on the fuel. The ceramic composite fuel formed of uranium dioxide (UO2), with the addition of beryllium oxide (BeO), presents high thermal conductivity compared with UO2. The oxidation of zirconium generates hydrogen gas that can create a detonation condition. One of the preferred options are the ferritic alloys formed of iron-chromium and aluminum (FeCrAl), that should avoid the hydrogen release due to oxidation. In general, the FeCrAl alloys containing 10–20Cr, 3–5Al, and 0–0.12Y in weight percent. The FeCrAl alloys should exhibit a slow oxidation kinetics due to chemical composition. Resistance to oxidation in the presence of steam is improved as a function of the content of chromium and aluminum. In this way, the thermal and mechanical properties of the UO2-BeO-10%vol, composite fuel were coupled with FeCrAl alloys and added to the fuel codes. In this work, we examine the fuel rod behavior of UO2-10%vol-BeO/FeCrAl, including a simulated transient of reactivity. The fuels behavior shown reduced temperature with UO2-BeO/Zr, UO2-BeO/FeCrAl also were compared with UO2/Zr system. The case reactivity initiated accident analyzed, reproducing the fuel rod called VA-1 using UO2/Zr alloys and compared with UO2-BeO/FeCrAl.

Palavras-Chave: aluminium alloys; beryllium oxides; chromium alloys; composite materials; computerized simulation; coupling; f codes; fuel rods; iron alloys; mechanical properties; nuclear fuels; thermodynamic properties; transients; uranium dioxide; zirconium

GOMES, DANIEL S.; ABE, ALFREDO Y.; MUNIZ, RAFAEL O.R.; GIOVEDI, CLAUDIA. Analysis of UO2-BEO fuel under transient using fuel performance code. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017. Disponível em: http://repositorio.ipen.br/handle/123456789/28184. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The main challenge in the nuclear area since the Fukushima Daiichi accident is to develop fuel materials to be applied in nuclear reactors aiming to increase the safety under normal operation as well as transient and accident conditions. These efforts are concentrated in the Advanced Technology Fuel (ATF) program that has as main scopes to study cladding materials to replace the zirconium-based alloys, and fuel materials presenting higher thermal conductivity compared to the conventional uranium dioxide fuel pellet. In this sense, iron-based alloys, which were used with a good performance as cladding material in the first Pressurized Water Reactors (PWR), have becoming a good option. The assessment of the behavior of different materials previously to perform irradiation tests, which are time consuming, can be performed using fuel performance codes, but for this, the conventional fuel performance codes must be modified to implement the properties of the materials that are being studied. This paper presents the results obtained using a modified version of the well-known TRANSURANUS code, obtained from the implementation of the stainless steel 348 properties as cladding material. The simulations were performed using data available in the open literature related to a PWR irradiation experiment. The results obtained using the modified version of the code were compared to those obtained using the original code version for zircaloy-4. The performance of both cladding materials was evaluated by means of the comparison of parameters such as gap thickness, fuel centerline temperature, internal pressure, and cladding stress and strain.

Palavras-Chave: cladding; comparative evaluations; computerized simulation; fuel rods; iron alloys; nuclear fuels; performance; pwr type reactors; stainless steel-348; steady-state conditions; t codes; zircaloy 4

GIOVEDI, CLAUDIA; MELO, CAIO; ABE, ALFREDO Y.; SILVA, ANTONIO T.; MARTINS, MARCELO R. Fuel performance of iron-based alloy cladding using modified TRANSURANUS code. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 4943-4949. Disponível em: http://repositorio.ipen.br/handle/123456789/30812. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

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

Palavras-Chave: aluminium; cladding; dispersion nuclear fuels; foils; fuel elements; fuel plates; irradiation; molybdenum alloys; physical properties; swelling; uranium alloys; zirconium

GOMES, DANIEL S.; SILVA, ANTONIO T.; ABE, ALFREDO Y.; MUNIZ, RAFAEL O.R.; GIOVEDI, CLAUDIA. High density fuels using dispersion and monolithic fuel. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017. Disponível em: http://repositorio.ipen.br/handle/123456789/28189. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The standard fuel system based on UO2–zirconium alloy has been utilized on nearly 90% of worldwide nuclear power light water reactors. After the Fukushima Daiichi accident, alternative cladding materials to zirconium-based alloys are being investigated in the framework of accident tolerance fuel (ATF) program. One of the concepts of ATF is related to cladding materials that could delay the onset of high temperature oxidation, as well as ballooning and burst, in order to improve reactor safety systems, and consequently increase the coping time for the reactor operators in accident condition, especially under Loss-of-Coolant Accident (LOCA) scenario. The ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium-based alloys based on its outstanding resistance to oxidation under superheated steam environment due to the development of alumina oxide on the alloy surface in case of LOCA; moreover, FeCrAl alloys present quite well performance under normal operation conditions due to the thin oxide rich in chromium that acts as a protective layer. The assessment and performance of new fuel systems rely on experimental irradiation program and fuel performance code simulation, therefore the aim of this work is to contribute to the computational modeling capabilities in the framework of the ATF concept. The well-known TRANSURANUS fuel performance code that is used by safety authorities, industries, laboratories, research centers and universities was modified in order to support FeCrAl alloy as cladding material. The modification of the TRANSURANUS code was based on existing data (material properties) from open literature and as verification process was performed considering LOCA accident scenario.

Palavras-Chave: accident-tolerant nuclear fuels; aluminium alloys; chromium alloys; cladding; comparative evaluations; fuel rods; iron alloys; loss of coolant; performance; t codes; zirconium alloys

ABE, ALFREDO Y.; MELO, CAIO; GIOVEDI, CLAUDIA; SILVA, ANTONIO T. Modification of TRANSURANUS fuel performance code in the ATF framework. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 21-25, 2019, Santos, SP. Proceedings... Rio de Janeiro: Associação Brasileira de Energia Nuclear, 2019. p. 5036-5045. Disponível em: http://repositorio.ipen.br/handle/123456789/30814. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item

Abstract: The research on accident-tolerant fuels (ATFs) increased after the Fukushima event. This benefitted risk management in nuclear operations. In this investigation, the physical properties of the materials being developed for the ATF program were compared with those of the standard UO2–Zr fuel system. The research efforts in innovative fuel design include rigorous characterization of thermal, mechanical, and chemical assessment, with the objectives of making the burnup cycle longer, increasing power density, and improving safety performance. Fuels must reach a high uranium density—above that supported by UO2—and possess coating that exhibits better oxidation resistance than Zircaloy. The uranium density and thermal conductivity of ATFs, such as U3Si2, UN, and UC, is higher than that of UO2; their combination with advanced cladding provides possible fuel–cladding options. An ideal combination of fuel and cladding must increase fuel performance in loss-of-coolant scenarios. The disadvantages of U3Si2, UN, and UC are their swelling rates, which are higher than that of UO2. The thermal conductivities of ATFs are approximately four times higher than that of UO2. To prevent the generation of hydrogen due to oxidation of zirconium-based alloys in contact with steam, cladding options, such as ferritic alloys, were studied. It was verified that FeCrAl alloys and SiC provide better response under severe conditions because of their thermophysical properties. The findings of this study indicate that U3Si2 and the FeCrAl fuel cladding concept should replace UO2–Zr as the fuel system of choice.

Palavras-Chave: accident-tolerant nuclear fuels; aluminium alloys; chromium alloys; cladding; comparative evaluations; computerized simulation; f codes; fuel rods; iron alloys; loss of coolant; steady-state conditions; swelling; thermal conductivity; thermal expansion; transients; uranium carbides; uranium nitrides; uranium silicides; zircaloy

GOMES, DANIEL S.; SILVA, ANTONIO T.; ABE, ALFREDO Y.; MUNIZ, RAFAEL O.R.; GIOVEDI, CLAUDIA. Simulation of accident-tolerant U3Si2 fuel using FRAPCON code. In: INTERNATIONAL NUCLEAR ATLANTIC CONFERENCE, October 22-27, 2017, Belo Horizonte, MG. Proceedings... Rio de Janeiro, RJ: Associação Brasileira de Energia Nuclear, 2017. Disponível em: http://repositorio.ipen.br/handle/123456789/28188. Acesso em: $DATA.

Esta referência é gerada automaticamente de acordo com as normas do estilo IPEN/SP (ABNT NBR 6023) e recomenda-se uma verificação final e ajustes caso necessário.

Como referenciar este item