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Shin, Sang Hun; Kim, Jun Hwan; Kim, June-Hyung; Ryu, Woo Seog; Park, Sang Gyu; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of]

2015-10-15

Nowadays, in Korea, advanced cladding such as FC92 is developed and its transient behaviors are required for the safety analysis of SFR. Design and safety analyses of sodium-cooled fast reactor [SFR] require understanding fuel pin responses to a wide range of off-normal events. In a loss-of-flow [LOF] or transient over-power [TOP], the temperature of the cladding is rapidly increased above its steady-state service temperature. Transient tests have been performed in sections of fuel pin cladding and a large data base has been established for austenitic stainless steel such as 20% cold-worked 316 SS and ferritic/martensitic steels such as HT9. This paper summarizes the technical status of transient testing facilities and their results. Previous researches showed the transient behaviors of HT9 cladding. For the safety analyses in SFR in Korea, simulated transient tests with newly developed FC92 as well as HT9 cladding are being carried out.

Construction of in-situ creep strain test facility for the SFR fuel cladding Energy Technology Data Exchange [ETDEWEB] Park, Sang Gyu; Heo, Hyeong Min; Kim, Jun Hwan; Kim, Sung Ho [KAERI, Daejeon [Korea, Republic of] 2016-05-15 In this study, in-situ laser inspection creep test machine was developed for the measuring the creep strain of SFR fuel cladding materials. Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances to a void swelling. HT9 steel [12CrMoVW] is initially developed as a material for power plants in Europe in the 1960. This steel has experienced to expose up to 200dpa in FFTE and EBR-II. Ferritic-Martensitic steel's maximum creep strength in existence is 180Mpa for 106 hour 600 .deg., but HT9 steel is 60Mpa. Because SFR is difficult to secure in developing and applying materials, HT9 steel has accumulated validated data and is suitable for SFR component. And also, because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels, such as HT9 and FC92[12Cr-2W] are preferable to utilize in the fuel cladding of an SFR in KAERI. The pressurized thermal creep test of HT9 and FC92 claddings are being conducted in KAERI, but the change of creep strain in cladding is not easy to measure during the creep test due to its pressurized and closed conditions. In this paper, in-situ laser inspection pressurized creep test machine developed for SFR fuel cladding specimens is described. Moreover, the creep strain rate of HT9 at 650 .deg. C was examined from the in-situ laser inspection pressurized creep test machine.
Boron-bearing Influences of 9Cr-0.5Mo-2W-V-Nb Ferritic/Martensitic Steels for a SFR Fuel Cladding International Nuclear Information System [INIS] Baek, Jong-Hyuk; Han, Chang-Hee; Kim, Woo-Gon; Kim, Sung-Ho; Lee, Chan-Bock 2008-01-01 Currently the principal materials in a SFR [sodium-cooled fast reactor] of Gen-IV nuclear system are considering stainless steels [e.g. austenitic steels and ferritic/martensitic steels] for pressure boundary and structural applications in the primary circuit [cladding, duct, cold and hot leg piping, and pressure vessel]. There are sound technical justifications for these material selections, and the adoption of these stainless steels for a wide range of nuclear and non-nuclear applications has generated much industrial technology and experience. However, there are strong incentives to develop advanced materials, especially cladding, for the Gen-IV SFR. The Gen-IV SFR is to have a considerable increase in safety and be economically competitive when compared with the conventional water reactors. To accomplish these objectives, the development of the fuel cladding material should be set forth as a premise because its integrity is directly related to those of the reactor system as well as the fuel in the Gen-IV SFR. Since last year, a R and D program was launched to develop the improved ferritic/martensitic steel for the Gen-IV SFR fuel cladding. Categories of materials considered in the program included 8 - 12% Cr ferritic/ martensitic steels. A strong recommendation was made for the development of a high strength steel equivalent to or superior to ASTM Gr.92 steel to offset the difficulties encountered with commercial available steels of the 8 - 12% Cr group. That is, since fuel cladding in the Gen-IV SFR would operate under higher temperatures than 600 .deg. C, contacting with liquid sodium, and be irradiated by neutrons to as high as 200dpa, the cladding should thus sustain both superior irradiation and temperature stabilities during an operational life. The newly developed advanced steel should overcome the severe drawback; mechanical properties, especially creep, are deteriorated at a higher temperature over 600 .deg. C. In this study, as one of the composition
Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor [SFR] International Nuclear Information System [INIS] Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock 2013-01-01 To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor [SFR], preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube
Development of Cr Electroplated Cladding Tube for preventing Fuel-Cladding Chemical Interaction [FCCI] Energy Technology Data Exchange [ETDEWEB] Kim, Jun Hwan; Woo, Je Woong; Kim, Sung Ho; Cheon, Jin Sik; Lee, Byung Oon; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2015-05-15 Metal fuel has been selected as a candidate fuel in the SFR because of its superior thermal conductivity as well as enhanced proliferation resistance in connection with the pyroprocessing. However, metal fuel suffers eutectic reaction [Fuel Cladding Chemical Interaction, FCCI] with the fuel cladding made of stainless steel at reactor operating temperature so that cladding thickness gradually reduces to endanger reactor safety. In order to mitigate FCCI, barrier concept has been proposed between the fuel and the cladding in designing fuel rod. Regarding this, KAERI has initiated barrier cladding development to prevent interdiffusion process as well as enhance the SFR fuel performance. Previous study revealed that Cr electroplating has been selected as one of the most promising options because of its technical and economic viability. This paper describes the development status of the Cr electroplating technology for the usage of fuel rod in SFR. This paper summarizes the status of Cr electroplating technology to prevent FCCI in metal fuel rod. It has been selected for the ease of practical application at the tube inner surface. Technical scoping, performance evaluation and optimization have been carried out. Application to the tube inner surface and in-pile test were conducted which revealed as effective.
Weld Joint Design for SFR Metallic Fuel Element Closures Energy Technology Data Exchange [ETDEWEB] Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Yoon, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2014-05-15 The sodium-cooled fast reactor [SFR] system is among the six systems selected for Gen-IV promising systems and expected to become available for commercial introduction around 2030. In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the joint designs for endplug welding were investigated. For the irradiation test of SFR metallic fuel element, the TIG welding technique was adopted and the welding joint design was developed based on the welding conditions and parameters established. In order to make SFR metallic fuel elements, the weld joint design was developed based on the TIG welding technique.
Endplug Welding Techniques developed for SFR Metallic Fuel Elements Energy Technology Data Exchange [ETDEWEB] Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2013-10-15 In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.
Endplug Welding Techniques developed for SFR Metallic Fuel Elements International Nuclear Information System [INIS] Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan 2013-01-01 In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established
Challenges in mechanical modeling of SFR fuel rod transient behavior Energy Technology Data Exchange [ETDEWEB] Feria, F.; Herranz, L. E. 2013-07-01 Modeling of SFR fuel rod mechanical behavior under transient conditions entails the development of a creep law to predict cladding viscoplastic strain. In this regard, this work is focused on defining a proper clad creep law structure as the basis to set a suitable model under SFR off-normal conditions as transient overpower and loss of fluid. To do so, a review of in-codes clad creep models has been done by using SAS-SFR, SCANAIR and ASTEC. The proposed creep model has been structured in two parts: viscoplastic behaviour before the failure [primary and secondary creep] and the failure due to viscoplastic collapse [tertiary creep]. In order to model the first part, Norton creep law has been proposed as a conservative option. An irradiation hardening factor should be included for best estimate calculations. The recommendation for the second part is to apply a failure criterion based on strain limit or rupture time, which allows achieving conservative results.
Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method International Nuclear Information System [INIS] Shin, Andong; Jeong, Hyedong; Suh, Namduk; Kim, Hyochan; Yang, Yongsik 2014-01-01 Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute [KAERI] and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety [KINS] has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even
Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method Energy Technology Data Exchange [ETDEWEB] Shin, Andong; Jeong, Hyedong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon [Korea, Republic of]; Kim, Hyochan; Yang, Yongsik [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2014-10-15 Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute [KAERI] and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety [KINS] has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even
Development of Melting Crucible Materials of Metallic Fuel Slug for SFR International Nuclear Information System [INIS] Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M. 2010-01-01 The fabrication process of metallic fuel for SFR[sodium fast reactor] of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA[minor actinide] elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized
Preliminary Analysis of the Bundle-Duct Interaction for the fuel of SFR Energy Technology Data Exchange [ETDEWEB] Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2008-10-15 BDI [Bundle-Duct Interaction] occurs in the fuel of SFR [Sodium-cooled Fast Reactor] due to the radial expansion and bowing of a fuel pin bundle. Under the BDI condition, excess cladding strain and hot spots would occur. Therefore, BDI, which is the dominant deformation mechanisms in a fuel pin bundle, should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE and BMBOO, have been developed to evaluate the BDI behavior. The bundle duct interaction model is also being developed for SFR in Korea. This model is based on ANSYS. In this paper, the fuel pin configuration model for the BDI calculation was established. The preliminary analysis of the bundle-duct interaction was performed to evaluate the fuel design concept.
Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor International Nuclear Information System [INIS] Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock 2007-09-01 The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor [SFR]. From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents [74%] were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number [9] of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean
Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor Energy Technology Data Exchange [ETDEWEB] Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock 2007-09-15 The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor [SFR]. From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents [74%] were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number [9] of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.
Preliminary Analysis of the Fuel Bundle Stiffness by ANSYS for SFR Energy Technology Data Exchange [ETDEWEB] Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2008-05-15 In SFR [Sodium-cooled Fast Reactor] the temperature of the fuel pin is higher than that of the hexagonal duct, so the thermal expansion rate of the fuel bundle is higher than that of the duct. The neutron fluence and the fuel pin pressure are also increased according to the burnup. So the radial expansion and bowing of a fuel pin bundle would occur, and then fuel bundle would interact with a duct. This phenomenon is called bundle-to-duct interaction [BDI]. Under the BDI condition, excess cladding strain and hot spots would occur. Therefore BDI as well as the core mechanics should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE, SHADOW, and MARSE, have been developed to evaluate the BDI behavior. The ANSYS based model is also being developed to analysis the bundle duct interaction for SFR in Korea. In this paper, the fuel pin/bundle model for analyzing the bending deflection and oval deformation was described. The preliminary analysis of the fuel bundle stiffness was performed by the developed model.
Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data Energy Technology Data Exchange [ETDEWEB] Yacout, A. M. [Argonne National Lab. [ANL], Argonne, IL [United States]. Nuclear Engineering Division; Billone, M. C. [Argonne National Lab. [ANL], Argonne, IL [United States]. Nuclear Engineering Division 2016-09-16 The US sodium cooled fast reactor [SFR] metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor [IFR] program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment [e.g., previous U-Fissium alloy fuel]. The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of data were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor [ALMR] program.
Metallic Reactor Fuel Fabrication for SFR Energy Technology Data Exchange [ETDEWEB] Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2015-05-15 The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of [1] fuel slug casting, [2] loading and fabrication of the fuel rods, and [3] fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.
Design of FCI Experiments to Understand Fuel Out-Pin Phenomena in the SFR Energy Technology Data Exchange [ETDEWEB] Heo, Hyo; Park, Seong Dae [Ulsan National Institute of Science and Technology, Ulsan [Korea, Republic of]; Jerng, Dong Wook; Bang, In Cheol [Chungang Univ., Seoul [Korea, Republic of] 2014-05-15 It is important to guarantee a passive nuclear safety regarding enhanced negative reactivity by fragmenting the molten fuel. In the SFR, it has a strong point that the negative reactivity is immediately introduced when the metal fuel is melted by the UTOP or ULOP accident. These characteristics of the metal fuel can prevent from progressing in severe accidents such as core disruptive accidents [CDA]. As key phenomena in the accidents, fuel-coolant interaction [FCI] phenomena have been studied over the last few decades. Especially, several previous researches focused on instability and fragmentation of a core melt jet in water. However, the studies showed too limited phenomena to fully understand. In the domestic SFR technology development, researches for severe accidents tend to lag behind ones of other countries. Or, South Korea has a very basic level of the research such as literature survey. Recently, the SAS4A code, which was developed at Argonne National Laboratory [ANL] for thermal-hydraulic and neutronic analyses of power and flow transients in liquid-metal-cooled nuclear reactors [LMRs], is still under development to consider for a metal fuel. The other countries carried out basic experiments for molten fuel and coolant interactions. However, in a high temperature condition, methods for analysis of structural interaction between molten fuel and fuel cladding are very limited. The ultimate objective of the study is to evaluate the possibility of recriticality accident induced by fuel-coolant interaction in the SFR adopting metal fuel. It is a key point to analyze the molten-fuel behavior based on the experimental results which show fuel-coolant interaction with the simulant materials. It is necessary to establish the test facility, to build database, and to develop physical models to understand the FCI phenomena in the SFR; molten fuel-coolant interaction as soon as the molten fuel is ejected to the sodium coolant channel and molten fuel-coolant interaction

Structure of fuel performance audit code for SFR metal fuel Energy Technology Data Exchange [ETDEWEB] Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon [Korea, Republic of]; Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon [Korea, Republic of] 2012-10-15 A Sodium Cooled Fast Reactor [SFR] is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety [KINS], the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.
Fabrication of preliminary fuel rods for SFR International Nuclear Information System [INIS] Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan 2012-01-01 Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I [EBR-I] and the Experimental Breeder Reactor-II [EBR-II] in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor [DFR] in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor [IFR] program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR [sodium-cooled fast reactor] related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC[quality control] of the fuel rods was performed
Status of SFR Metal Fuel Development International Nuclear Information System [INIS] Lee, Chan Bock; Lee, Byoung Oon; Kim, Ki Hwan; Kim, Sung Ho 2013-01-01 Conclusion: • Metal fuel recycling in SFR: - Enhanced utilization of uranium resource; - Efficient transmutation of minor actinides; - Inherent passive reactor safety; - Proliferation resistance with pyro-electrochemical fuel recycling. • Demonstration of technical feasibility of recycling TRU metal fuel by 2020: - Remote fuel fabrication; - Irradiation performance up to high burnup
Economic Analysis of Pyro-SFR Fuel Cycle International Nuclear Information System [INIS] Gao, Fanxing; Park, Byungheung; Kwon, Eunha; Ko, Wonil 2010-01-01 In this study, based on the material flow the economics of Pyro-SFR has been estimated. These are mainly two methodologies to perform nuclear fuel cycle cost study which is based on the material flow calculations. One is equilibrium model and the other is dynamic model. Equilibrium model focus on the batch study with the assumptions that the whole system is in a steady state and mass flow as well as the electricity production all through the fuel cycle is in equilibrium state, which calculates the electricity production within a certain period and associated material flow with reference to unit cost in order to obtain the cost of electricity. Dynamic model takes the time factor into consideration to simulate the actual cases. Compared with the dynamic analysis model, the outcome of equilibrium model is more theoretical comparisons, especially with regard to the large uncertainty of the development of the pyro-technology evaluated. In this study equilibrium model was built to calculate the material flow on a batch basis. With the unit cost being determined, the cost of each step of fuel cycle could be obtained, so does the FMC. Due to the unavoidable uncertainty with certain unit costs, evaluated cost range and uncertainty study are applied. Sensitivity analysis has also been performed to obtain the breakeven uranium price for Pyro-SFR against PWR-O T. Economics of Pyro-SFR fuel cycle scenario has been calculated and compared by employing equilibrium model. The LFCC were obtained, Pyro-SFR 7.68 mills/kWh. The Uranium price is the dominant driver of LFCC. Pyro-techniques also weight considerably in Pyro-SFR scenario. On consideration of the current unavoidable uncertainties introduced by certain cost data, cost range and triangle techniques were used to perform the uncertainty study which indicates that the gap between Pyro-SFR and PWR-O T fuel cycle scenario is relatively small
KALIMER-600-clad Core Fuel Assembly Calculation using MATRA-LMR [V2.0] Code International Nuclear Information System [INIS] Kim, Young Gyun; Kim, Young Il 2006-12-01 Since the sodium boiling point is very high, maximum cladding and pin temperatures are used for design limit condition in sodium cooled liquid metal reactor. It is necessary to predict accurately the temperature distribution in the core and in the subassemblies to increase the sodium coolant efficiency. Based on the MATRA code, which is developed for PWR analysis, MATRA-LMR has been developed for SFR. The major modifications are: the sodium properties table is implemented as subprogram in the code, Heat transfer coefficients are changed for SFR, te pressure drop correlations are changed for more accurate calculations, which are Novendstern, Chiu-Rohsenow-Todreas, and Cheng-Todreas correlations. This This report describes briefly code structure and equations of MATRA-LMR [Version 2.0], explains input data preparation and shows some calculation results for the KALIMER-600-clad core fuel assembly for which has been performed the conceptual design of the core in the year 2006
Compatibility Behavior of the Ferritic-Martensitic Steel Cladding under the Liquid Sodium Environment Energy Technology Data Exchange [ETDEWEB] Kim, Jun Hwan; Baek, Jong Hyuk; Kim, Sung Ho; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2012-05-15 Fuel cladding is a component which confines uranium fuel to transport energy into the coolant as well as protect radioactive species from releasing outside. Sodium-cooled Fast Reactor [SFR] has been considered as one of the most probable next generation reactors in Korea because it can maximize uranium resource as well as reduce the amount of PWR spent fuel in conjunction with pyroprocessing. Sodium has been selected as the coolant of the SFR because of its superior fast neutron efficiency as well as thermal conductivity, which enables high power core design. However, it is reported that the fuel cladding materials like austenitic and ferritic stainless steel react sodium coolant so that the loss of the thickness, intergranular attack, and carburization or decarburization process may happen to induce the change of the mechanical property of the cladding. This study aimed to evaluate material property of the cladding material under the liquid sodium environment. Ferritic-martensitic steel [FMS] coupon and cladding tube were exposed at the flowing sodium then the microstructural and mechanical property were evaluated. mechanical property of the cladding was evaluated using the ring tension test
Core Design Concept and Core Structural Material Development for a Prototype SFR International Nuclear Information System [INIS] Chang, Jinwook 2013-01-01 • Core design Concept: – Initial core is Uranium metal fueled core, then it will evolve into TRU core; – Tight pressure drop constraint lowers power density; – Trade-off studies with relaxed pressure drop constraint [~0.4MPa] are on-going; – Major feature will be finalized this year. • KAERI is developing advanced cladding for high burnup fuel in Ptototype SFR: – Advanced cladding materials are now developing, which shows superior high temperature mechanical property to the conventional material; – Processing technologies related to tube making process are now developed to enhance high temperature mechanical propertyl – Preliminary HT9 cladding tube was manufactured and out-of pile mechanical properties were evaluated. Advanced cladding tube is now being developed and being prepared for irradiation test
Korean SFR development program and technical activities for improving economical competitiveness International Nuclear Information System [INIS] Yoo, Jaewoon 2013-01-01 Future Plan: • Construction cost evaluation of PGSFR and commercial SFR; – Component based capital cost evaluation of PGSFR is undergoing and will be completed by the first half of 2014; – Component cost is only based on the experience from that of LWR; • Cost Benefit Analysis of Future Nuclear Energy Mix; – With revised National Energy Plan [as of 2013]; – Near-term: Benefit from LWR spent fuel recycling: - In Korean law, Share of Expense for spent fuel disposal is reserved as 0.4M$ per a LWR spent fuel assembly [as of 2003]; – Long-term: Competitive power plant to LWR with self sustainable feature; • Revision of commercial SFR conceptual design; – Less constraint in material [fuel, cladding] irradiation experience; – More innovative features as long-term goal
Fuel assembly and fuel cladding tube International Nuclear Information System [INIS] Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro. 1996-01-01 A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. [I.N.]
Radioactive Waste Generation in Pyro-SFR Nuclear Fuel Cycle International Nuclear Information System [INIS] Gao, Fanxing; Park, Byung Heung; Ko, Won Il 2011-01-01 Which nuclear fuel cycle option to deploy is of great importance in the sustainability of nuclear power. SFR fuel cycle employing pyroprocessing [named as Pyro- SFR Cycle] is one promising fuel cycle option in the near future. Radioactive waste generation is a key criterion in nuclear fuel cycle system analysis, which considerably affects the future development of nuclear power. High population with small territory is one special characteristic of ROK, which makes the waste management pretty important. In this study, particularly the amount of waste generation with regard to the promising advanced fuel cycle option was evaluated, because the difficulty of deploying an underground repository for HLW disposal requires a longer time especially in ROK
Fuel-cladding chemical interaction International Nuclear Information System [INIS] Gueneau, C.; Piron, J.P.; Dumas, J.C.; Bouineau, V.; Iglesias, F.C.; Lewis, B.J. 2015-01-01 The chemistry of the nuclear fuel is very complex. Its chemical composition changes with time due to the formation of fission products and depends on the temperature level history within the fuel pellet and the clad during operation. Firstly, in thermal reactors, zircaloy oxidation from reaction with UO 2 fuel under high-temperature conditions will be addressed. Then other fuel-cladding interaction phenomena occurring in fast reactors will be described. Large thermal gradients existing between the centre and the periphery of the pellet induce the radial redistribution of the fuel constituents. The fuel pellet can react with the clad by different corrosion processes which can involve actinide and/or fission product transport via gas, liquid or/and solid phases. All these phenomena are briefly described in the case of different kinds of fuels [oxide, carbide, nitride, metallic] to be used in fast reactors. The way these phenomena are taken into account in fuel performance codes is presented. [authors]
BWR fuel clad behaviour following LOCA International Nuclear Information System [INIS] Chaudhry, S.M.; Vyas, K.N.; Dinesh Babu, R. 1996-01-01 Flow and pressure through the fuel coolant channel reduce rapidly following a loss of coolant accident. Due to stored energy and decay heat, fuel and cladding temperatures rise rapidly. Increase in clad temperature causes deterioration of mechanical properties of clad material. This coupled with increase of pressure inside the cladding due to accumulation of fission gases and de-pressurization of coolant causes the cladding to balloon. This phenomenon is important as it can reduce or completely block the flow passages in a fuel assembly causing reduction of emergency coolant flow. Behaviour of a BWR clad is analyzed in a design basis LOCA. Fuel and clad temperatures following a LOCA are calculated. Fission gas release and pressure is estimated using well established models. An elasto-plastic analysis of clad tube is carried out to determine plastic strains and corresponding deformations using finite-element technique. Analysis of neighbouring pins gives an estimate of flow areas available for emergency coolant flow. [author]. 7 refs, 6 figs, 3 tabs
Chemical compatibility between cladding alloys and advanced fuels International Nuclear Information System [INIS] Fee, D.C.; Johnson, C.E. 1975-05-01 The National Advanced Fuels Program requires chemical, mechanical, and thermophysical properties data for cladding alloys. The compatibility behavior of cladding alloys with advanced fuels is critically reviewed. in carbide fuel pins, the principal compatibility problem is cladding carburization, diffusion of carbon into the cladding matrix accompanied by carbide precipitation. Carburization changes the mechanical properties of the cladding alloy. The extent of carburization increases in sodium [versus gas] bonded fuels. The depth of carburization increases with increasing sesquicarbide [M 2 C 3 ] content of the fuel. In nitride fuel pins, the principal compatibility problem is cladding nitriding, diffusion of nitrogen into the cladding matrix accompanied by nitride precipitation. Nitriding changes the mechanical properties of the cladding alloy. In both carbide and nitride fuel pins, fission products do not migrate appreciably to the cladding and do not appear to contribute to cladding attack. 77 references. [U.S.]
Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel Energy Technology Data Exchange [ETDEWEB] Boltax, A [Westinghouse Electric Corporation, Advanced Reactor Division, Madison, PA [United States]; Biancheria, A 1977-04-01 Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction [FCMI] effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. [author]
Fuel-cladding mechanical interaction effects in fast reactor mixed oxide fuel International Nuclear Information System [INIS] Boltax, A.; Biancheria, A. 1977-01-01 Thermal and fast reactor irradiation experiments on mixed oxide fuel pins under steady-state and power change conditions reveal evidence for significant fuel-cladding mechanical interaction [FCMI] effects. Analytical studies with the LIFE-III fuel performance code indicate that high cladding stresses can be produced by general and local FCMI effects. Also, evidence is presented to show that local cladding strains can be caused by the accumulation of cesium at the fuel-cladding interface. Although it is apparent that steady-state FCMI effects have not given rise to cladding breaches in current fast reactors, it is anticipated that FCMI may become more important in the future because of interest in: higher fuel burnups; increased power ramp rates; load follow operation; and low swelling cladding alloys. [author]
Nuclear-powered pacemaker fuel cladding study International Nuclear Information System [INIS] Shoup, R.L. 1976-07-01 The fabrication of fuel capsules with refractory metal and alloy clads used in nuclear-powered cardiac pacemakers precludes the expedient dissolution of the clad in inorganic acid solutions. An experiment to measure penetration rates of acids on commonly used fuel pellet clads indicated that it is not impossible, but that it would be very difficult to dissolve the multiple cladding. This work was performed because of a suggestion that a 238 PuO 2 -powered pacemaker could be transformed into a terrorism weapon
Fuel cladding tube and fuel rod for BWR type reactor International Nuclear Information System [INIS] Urata, Megumu; Mitani, Shinji. 1995-01-01 A fuel cladding tube has grooves fabricated, on the surface thereof, with a predetermined difference between crest and bottom [depth of the groove] in the circumferential direction. The cross sectional shape thereof is sinusoidal. The distribution of the grain size of iron crud particles in coolants is within a range about from 2μm to 12μm. If the surface roughness of the fuel cladding tube [depth of the groove] is determined greater than 1.6μm and less than 12.5, iron cruds in coolants can be positively deposited on the surface of the fuel cladding tube. In addition, once deposited iron cruds can be prevented from peeling from the surface of the fuel cladding tube. With such procedures, iron cruds deposited and radioactivated on the fuel cladding tube can be prevented from peeling, to prevent and reduce the increase of radiation dose on the surface of the pipelines without providing any additional device. [I.N.]
Nuclear fuel cladding material International Nuclear Information System [INIS] Nakahigashi, Shigeo. 1982-01-01 Purpose: To largely improve the durability and the safety of fuel cladding material. Constitution: Diffusion preventive layers, e.g., aluminum or the like are covered on both sides of a zirconium alloy base layer of thin material, and corrosion resistant layers, e.g., copper or the like are covered thereon. This thin plate material is intimately wound in a circularly tubular shape in a plurality of layers to form a fuel cladding tube. With such construction, corrosion of the tube due to fuel and impurity can be prevented by the corrosion resistant layers, and the diffusion of the corrosion resistant material to the zirconium alloy can be prevented by the diffusion preventive layers. Since a plurality of layers are cladded, even if the corrosion resistant layers are damaged or cracked due to stress corrosion, only one layer is damaged or cracked, but the other layers are not affected. [Sekiya, K.]
Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts International Nuclear Information System [INIS] Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji 2010-01-01 The Korea Atomic Energy Research Institute [KAERI] has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study
Mechanisms of fuel-cladding chemical interaction: US interpretation International Nuclear Information System [INIS] Adamson, M.G. 1977-01-01 Proposed mechanisms of fuel-cladding chemical interaction [FCCI] in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components [as tellurides] in liquid Cs-Te. [author]

Mechanisms of fuel-cladding chemical interaction: US interpretation Energy Technology Data Exchange [ETDEWEB] Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA [United States] 1977-04-01 Proposed mechanisms of fuel-cladding chemical interaction [FCCI] in LMFBR fuel pins are reviewed and examined in terms of in-pile and out-of-pile data. From this examination several factors are identified which may govern the occurrence of localized deep intergranular penetrations of Type-316SS cladding. Using a plausible mechanistic hypothesis for FCCI, first steps have been taken towards developing a quantitative, physically-meaningful, mathematical method of predicting cladding wastage in operating fuel pins. Both kinetic and thermodynamic aspects of FCCI are considered in the development of this prediction method, together with a fuel chemistry model that describes the evolution of thermochemical conditions at the fuel-cladding gap. On the basis of results from recent fuel pin and laboratory tests a thermal transport mechanism has been proposed to explain the thermal gradient-induced migration of Fe, Cr, and Ni from cladding into the fuel. This mechanism involves chemical transport of the metallic cladding components [as tellurides] in liquid Cs-Te. [author]
Analysis of fuel cladding chemical interaction in mixed oxide fuel pins International Nuclear Information System [INIS] Weber, J.W.; Dutt, D.S. 1976-01-01 An analysis is presented of the observed interaction between mixed oxide 75 wt percent UO 2 --25 wt percent PuO 2 fuel and 316--20 percent CW stainless steel cladding in LMFBR type fuel pins irradiated in EBR-II. A description is given of the test pins and their operating conditions together with, metallographic observations and measurements of the fuel/cladding reaction, and a correlation equation is developed relating depth of cladding attack to temperature and burnup. Some recent data on cladding reaction in fuel pins with low initial O/M in the fuel are given and compared with the correlation equation curves
X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel Energy Technology Data Exchange [ETDEWEB] Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon [Korea, Republic of] 2016-10-15 In KINS [Korea Institute of Nuclear Safety], to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable.
X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel International Nuclear Information System [INIS] Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk 2016-01-01 In KINS [Korea Institute of Nuclear Safety], to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable
Characterization and modelling of the thermodynamic behavior of SFR fuel under irradiation International Nuclear Information System [INIS] Pham-Thi, Tam-Ngoc 2014-01-01 For a burn-up higher than 7 at%, the volatile FP like Cs, I and Te or metallic [Mo] are partially released from the fuel pellet in order to form a layer of compounds between the outer surface of the fuel and the inner surface of the stainless cladding. This layer is called the JOG, french acronym for Joint-Oxyde-Gaine. My subject is focused on two topics: the thermodynamic study of the [Cs-I-Te-Mo-O] system and the migration of those FP towards the gap to form the JOG. The thermodynamic study was the first step of my work. On the basis of critical literature survey, the following systems have been optimized by the CALPHAD method: Cs-Te, Cs-I and Cs-Mo-O. In parallel, an experimental study is undertaken in order to validate our CALPHAD modelling of the Cs-Te system. In a second step, the thermodynamic data coming from the CALPHAD modelling have been introduced into the database that we use with the thermochemical computation code ANGE [CEA code derived from the SOLGASMIX software] in order to calculate the chemical composition of the irradiated fuel versus burn-up and temperature. In a third and last step, the thermochemical computation code ANGE [Advanced Numeric Gibbs Energy minimizer] has been coupled with the fuel performance code GERMINAL V2, which simulates the thermo-mechanical behavior of SFR fuel. [author] [fr
Method of processing spent fuel cladding tubes International Nuclear Information System [INIS] Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi. 1986-01-01 Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. [Takahashi, M.]
Pressurized water reactor fuel performance problems connected with fuel cladding corrosion processes International Nuclear Information System [INIS] Dobrevski, I.; Zaharieva, N. 2008-01-01 Generally, Pressurized Water Reactor [WWER, PWR] Fuel Element Performance is connected with fuel cladding corrosion and crud deposition processes. By transient to extended fuel cycles in nuclear power reactors, aiming to achieve higher burnup and better fuel utilization, the role of these processes increases significantly. This evolution modifies the chemical and electrochemical conditions in the reactor primary system, including change of fuel claddings' environment. The higher duty cores are always attended with increased boiling [sub-cooled nucleate boiling] mainly on the feed fuel assemblies. This boiling process on fuel cladding surfaces can cause different consequences on fuel element cladding's environment characteristics. In the case of boiling at the cladding surfaces without or with some cover of corrosion product deposition, the behavior of gases dissolved in water phase is strongly influenced by the vapor generation. The increase of vapor partial pressure will reduce the partial pressures of dissolved gases and will cause their stripping out. By these circumstances the concentrations of dissolved gases in cladding wall water layer can dramatically decrease, including also the case by which all dissolved gases to be stripped out. On the other hand it is known that the hydrogen is added to primary coolant in order to avoid the production of oxidants by radiolysis of water. It is clear that if boiling strips out dissolved hydrogen, the creation of oxidizing conditions at the cladding surfaces will be favored. In this case the local production of oxidants will be a result from local processes of water radiolysis, by which not only both oxygen [O 2 ] and hydrogen [H 2 ] but also hydrogen peroxide [H 2 O 2 ] will be produced. While these hydrogen and oxygen will be stripped out preferentially by boiling, the bigger part of hydrogen peroxide will remain in wall water phase and will act as the most important factor for creation of oxidizing conditions in fuel
Elastic plastic analysis of fuel element assemblies - hexagonal claddings and fuel rods International Nuclear Information System [INIS] Mamoun, M.M.; Wu, T.S.; Chopra, P.S.; Rardin, D.C. 1979-01-01 Analytical studies have been conducted to investigate the structural, thermal, and mechanical behavior of fuel rods, claddings and fuel element assemblies of several designs for a conceptual Safety Test Facility [STF]. One of the design objectives was to seek a geometrical configuration for a clad by maximizing the volume fraction of fuel and minimizing the resultant stresses set-up in the clad. The results of studies conducted on various geometrical configurations showed that the latter design objective can be achieved by selecting a clad of an hexagonal geometry. The analytical studies necessitated developing solutions for determining the stresses, strains, and displacements experienced by fuel rods and an hexagonal cladding subjected to thermal fuel-bowing loads acting on its internal surface, the external pressure of the coolant, and elevated temperatures. This paper presents some of the initially formulated analytical methods and results. It should be emphasized that the geometrical configuration considered in this paper may not necessarily be similar to that of the final design. Several variables have been taken into consideration including cladding thickness, the dimensions of the fuel rod, the temperature of the fuel and cladding, the external pressure of the cooling fluid, and the mechanical strength properties of fuel and cladding. A finite-element computer program, STRAW Code, has also been employed to generate several numerical results which have been compared with those predicted by employing the initially formulated solutions. The theoretically predicted results are in good agreement with those of the STRAW Code. [orig.]
Inspection system for Zircaloy clad fuel rods International Nuclear Information System [INIS] Yancey, M.E.; Porter, E.H.; Hansen, H.R. 1975-10-01 A description is presented of the design, development, and performance of a remote scanning system for nondestructive examination of fuel rods. Characteristics that are examined include microcracking of fuel rod cladding, fuel-cladding interaction, cladding thickness, fuel rod diameter variation, and fuel rod bowing. Microcracking of both the inner and outer fuel rod surfaces and variations in wall thickness are detected by using a pulsed eddy current technique developed by Argonne National Laboratory [ANL]. Fuel rod diameter variation and fuel rod bowing are detected by using two linear variable differential transformers [LVDTs] and a signal conditioning system. The system's mechanical features include variable scanning speeds, a precision indexing system, and a servomechanism to maintain proper probe alignment. Initial results indicate that the system is a very useful mechanism for characterizing irradiated fuel rods
Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design International Nuclear Information System [INIS] Roake, W.E. 1977-01-01 Fuel-cladding-chemical-interaction [FCCI] is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals
Implications and control of fuel-cladding chemical interaction for LMFBR fuel pin design Energy Technology Data Exchange [ETDEWEB] Roake, W E [Westinghouse-Hanford Co., Richland, WA [United States] 1977-04-01 Fuel-cladding-chemical-interaction [FCCI] is typically incorporated into the design of an LMFBR fuel pin as a wastage allowance. Several interrelated factors are considered during the evolution of an LMFBR fuel pin design. Those which are indirectly affected by FCCI include: allowable pin power, fuel restructuring, fission gas migration and release from the fuel, fuel cracking, fuel swelling, in-reactor cladding creep, cladding swelling, and the cladding mechanical strain. Chemical activity of oxygen is the most readily controlled factor in FCCI. Two methods are being investigated: control of total oxygen inventory by limiting fuel O/M, and control of oxygen activity with buffer metals.
The fuel-cladding interfacial friction coefficient in water-cooled reactor fuel rods International Nuclear Information System [INIS] Smith, E. 1979-01-01 A central problem in the development of cladding failure criteria and of effective operational, design or material remedies is to know whether the cladding stress is enhanced significantly near cladding ridges, pellet chips or fuel pellet cracks; the latter may also be coincident with cladding ridges at pellet-pellet interfaces. As regards the fuel pellet crack source of cladding stress concentration, the magnitude of the uranium dioxide-Zircaloy interfacial friction coefficient μ governs the magnitude and distribution of the enhanced cladding stress. Considerable discussion, particularly at a Post-Conference Seminar associated with the SMIRT 4 Conference, has focussed on the value of μ, the author taking the view that it is unlikely to be large [< 0.5]. The reasoning behind this view is as follows. A fuel pellet should fracture during a power ramp when the tensile hoop stress within the pellet exceeds the fuel's fracture stress. Since the preferred position for a fuel pellet crack to form is at the fuel-cladding interface midway between existing fuel cracks, where the interfacial shear stress changes sign, the pellet segment size after a power ramp provides a limit to the magnitude of the interfacial shear stresses and consequently to the value of μ. With this argument as a basis, the author's early work used the Gittus fuel rod model, in which there is a symmetric distribution of fuel pellet cracks and symmetric interfacial slippage, to show that μ < 0.5 if it is assumed that the average hoop stress within the cladding attains yield levels. It was therefore suggested that a high interfacial friction coefficient is unlikely to be operative during a power ramp; this result was used to support the view that interfacial friction effects do not play a dominant role in stress corrosion crack formation within the cladding. [orig.]
Review and evaluation of cladding attack of LMFBR fuel International Nuclear Information System [INIS] Koizumi, M.; Nagai, S.; Furuya, H.; Muto, T. 1977-01-01 The behavior of cladding inner wall corrosion during irradiation was evaluated in terms of fuel density, fuel form, O/M ratio, plutonium concentration, cladding composition, cladding pretreatment, cladding inner diameter, burnup and cladding inner wall temperature. Factors which influence the corrosion are O/M ratio [oxygen to metal ratio], burn up, cladding inner diameter and cladding inner wall temperature. Maximum cladding inner wall corrosion depth was formulated as a function of O/M ratio, burn up and cladding inner wall temperature
Cladding properties under simulated fuel pin transients International Nuclear Information System [INIS] Hunter, C.W.; Johnson, G.D. 1975-01-01 A description is given of the HEDL fuel pin testing program utilizing a recently developed Fuel Cladding Transient Tester [FCTT] to generate the requisite mechanical property information on irradiated and unirradiated fast reactor fuel cladding under temperature ramp conditions. The test procedure is described, and data are presented
Fuel cladding mechanical interaction during power ramps International Nuclear Information System [INIS] Guerin, Y. 1985-01-01 Mechanical interaction between fuel and cladding may occur as a consequence of two types of phenomenon: i] fuel swelling especially at levels of caesium accumulation, and ii] thermal differential expansion during power changes. Slow overpower ramps which may occur during incidental events are of course one of the circumstances responsible for this second type of fuel cladding mechanical interaction [FCMI]. Experiments and analysis of this problem that have been done at C.E.A. allow to determine the main parameters which will fix the level of stress and the risk of damage induced by the fuel in the cladding during overpower transients
Fuel-cladding chemical interaction in mixed-oxide fuels International Nuclear Information System [INIS] Lawrence, L.A.; Weber, J.W.; Devary, J.L. 1978-10-01 The character and extent of fuel-cladding chemical interaction [FCCI] was established for UO 2 -25 wt% PuO 2 clad with 20% cold worked Type 316 stainless steel irradiated at high cladding temperatures to peak burnups greater than 8 atom %. The data base consists of 153 data sets from fuel pins irradiated in EBR-II with peak burnups to 9.5 atom %, local cladding inner surface temperatures to 725 0 C, and exposure times to 415 equivalent full power days. As-fabricated oxygen-to-metal ratios [O/M] ranged from 1.938 to 1.984 with the bulk of the data in the range 1.96 to 1.98. HEDL P-15 pins provided data at low heat rates, approx. 200 W/cm, and P-23 series pins provided data at higher heat rates, approx. 400 W/cm. A design practice for breeder reactors is to consider an initial reduction of 50 microns in cladding thickness to compensate for possible FCCI. This approach was considered to be a conservative approximation in the absence of a comprehensive design correlation for extent of interaction. This work provides to the designer a statistically based correlation for depth of FCCI which reflects the influences of the major fuel and operating parameters on FCCI
Experimental assessment of fuel-cladding interactions Energy Technology Data Exchange [ETDEWEB] Wood, Elizabeth Sooby [Los Alamos National Lab. [LANL], Los Alamos, NM [United States] 2017-06-29 A range of fuel concepts designed to better tolerate accident scenarios and reactor transients are currently undergoing fundamental development at national laboratories as well as university and industrial partners. Pellet-clad mechanical and chemical interaction can be expected to affect fuel failure rates experienced during steady state operation, as well as dramatically impact the response of the fuel form under loss of coolant and other accident scenarios. The importance of this aspect of fuel design prompted research initiated by AFC in FY14 to begin exploratory efforts to characterize this phenomenon for candidate fuelcladding systems of immediate interest. Continued efforts in FY15 and FY17 aimed to better understand and simulate initial pellet-clad interaction with little-to-no pressure on the pellet-clad interface. Reported here are the results from 1000 h heat treatments at 400, 500, and 600°C of diffusion couples pairing UN with a FeCrAl alloy, SiC, and Zr-based cladding candidate sealed in evacuated quartz ampoules. No gross reactions were observed, though trace elemental contaminants were identified.
Sodium-cooled fast reactor [SFR] fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient Energy Technology Data Exchange [ETDEWEB] Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr 2014-12-15 Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor [SFR] fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.
Pellet-clad interaction in water reactor fuels Energy Technology Data Exchange [ETDEWEB] NONE 2004-07-01 The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. [A.L.B.]
Pellet-clad interaction in water reactor fuels International Nuclear Information System [INIS] 2004-01-01 The aim of this seminar is was to draw up a comprehensive picture of the pellet clad interaction and its impact on the fuel rod. This document is a detailed abstract of the papers presented during the following five sessions: industrial goals, fuel material behaviour in PCI situation, cladding behaviour relevant to PCI, in pile rod behaviour and modelling of the mechanical interaction between pellet and cladding. [A.L.B.]

UK experience on fuel and cladding interaction in oxide fuels Energy Technology Data Exchange [ETDEWEB] Batey, W [Dounreay Experimental Reactor Establishment, Thurso, Caithness [United Kingdom]; Findlay, J R [AERE, Harwell, Didcot, Oxon [United Kingdom] 1977-04-01 The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor [DFR]. Supporting information has been obtained from irradiation programmes in Materials Testing Reactors [MTR]. Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed.
UK experience on fuel and cladding interaction in oxide fuels International Nuclear Information System [INIS] Batey, W.; Findlay, J.R. 1977-01-01 The occurrence of fuel cladding interactions in fast reactor fuels has been observed in UK irradiations over a period of years. Chemical incompatibility between fuel and clad represents a potential source of failure and has, on this account, been studied using a variety of techniques. The principal fuel of interest to the UK for fast reactor application is mixed uranium plutonium oxide clad in stainless steel and it is in this field that the majority of work has been concentrated. Some consideration has been given to carbide fuels, because of their application as an advanced fuel. This experience is described in the accompanying paper. Several complementary initiatives have been followed to investigate the interactions in oxide fuel. The principal source of experimental information is from the experimental fuel irradiation programme in the Dounreay Fast Reactor [DFR]. Supporting information has been obtained from irradiation programmes in Materials Testing Reactors [MTR]. Conditions approaching those in a fast reactor are obtained and the effects of specific variables have been examined in specifically designed experiments. Out-of-reactor experiments have been used to determine the limits of fuel and cladding compatibility and also to give indications of corrosion The observations from all experiments have been examined in the light of thermo-dynamic predictions of fuel behaviour to assess the relative significance of various observations and operating conditions. An experimental programme to control and limit the interactions in oxide fuel is being followed
Fuel compliance model for pellet-cladding mechanical interaction International Nuclear Information System [INIS] Shah, V.N.; Carlson, E.R. 1985-01-01 This paper describes two aspects of fuel pellet deformation that play significant roles in determining maximum cladding hoop strains during pellet-cladding mechanical interaction: compliance of fragmented fuel pellets and influence of the pellet end-face design on the transmission of axial compressive force in the fuel stack. The latter aspect affects cladding ridge formation and explains several related observations that cannot be explained by the hourglassing model. An empirical model, called the fuel compliance model and representing the above aspects of fuel deformation, has been developed using the results from two Halden experiments and incorporated into the FRAP-T6 fuel performance code
High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel Energy Technology Data Exchange [ETDEWEB] Perez, Emmanuel [Idaho National Lab. [INL], Idaho Falls, ID [United States]; Keiser, Jr., Dennis D. [Idaho National Lab. [INL], Idaho Falls, ID [United States]; Forsmann, Bryan [Boise State Univ., ID [United States]; Janney, Dawn E. [Idaho National Lab. [INL], Idaho Falls, ID [United States]; Henley, Jody [Idaho National Lab. [INL], Idaho Falls, ID [United States]; Woolstenhulme, Eric C. [Idaho National Lab. [INL], Idaho Falls, ID [United States] 2016-02-01 High-temperature fuel-cladding chemical interactions [FCCI] between TRIGA [Training, Research, Isotopes, General Atomics] fuel elements and the 304 stainless steel [304SS] are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium [U] particles dispersed in a zirconium-hydride [Zr-H] matrix. In reactor, the fuel is encased in 304-stainless-steel [304SS] or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide [Er-O] additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state [prior to exposure to high temperature], characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy [SEM] and transmission electron microscopy [TEM] with sample preparation via focused ion beam in situ-liftout-technique.
High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel International Nuclear Information System [INIS] Perez, Emmanuel; Keiser Jr, Dennis D.; Forsmann, Bryan; Janney, Dawn E.; Henley, Jody; Woolstenhulme, Eric C. 2016-01-01 High-temperature fuel-cladding chemical interactions [FCCI] between TRIGA [Training, Research, Isotopes, General Atomics] fuel elements and the 304 stainless steel [304SS] are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium [U] particles dispersed in a zirconium-hydride [Zr-H] matrix. In reactor, the fuel is encased in 304-stainless-steel [304SS] or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide [Er-O] additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state [prior to exposure to high temperature], characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy [SEM] and transmission electron microscopy [TEM] with sample preparation via focused ion beam in situ-liftout-technique.
Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting International Nuclear Information System [INIS] Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B. 2013-01-01 The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II [EBR-II] driver and the Fast Flux Test Facility [FFTF] fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs
Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting Energy Technology Data Exchange [ETDEWEB] Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B. [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2013-10-15 The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II [EBR-II] driver and the Fast Flux Test Facility [FFTF] fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs.
Effects of cold worked and fully annealed claddings on fuel failure behaviour International Nuclear Information System [INIS] Saito, Shinzo; Hoshino, Hiroaki; Shiozawa, Shusaku; Yanagihara, Satoshi 1979-12-01 Described are the results of six differently heat-treated Zircaloy clad fuel rod tests in NSRR experiments. The purpose of the test is to examine the extent of simulating irradiated claddings in mechanical properties by as-cold worked ones and also the effect of fully annealing on the fuel failure bahaviour in a reactivity initiated accident [RIA] condition. As-cold worked cladding does not properly simulated the embrittlement of the irradiated one in a RIA condition, because the cladding is fully annealed before the fuel failure even in the short transient. Therefore, the fuel behaviour such as fuel failure threshold energy, failure mechanism, cladding deformation and cladding oxidation of the fully annealed cladding fuel, as well as that of the as-cold worked cladding fuel, are not much different from that of the standard stress-relieved cladding fuel. [author]
The ballooning of fuel cladding tubes: theory and experiment International Nuclear Information System [INIS] Shewfelt, R.S.W. 1988-01-01 Under some conditions, fuel clad ballooning can result in considerable strain before rupture. If ballooning were to occur during a loss-of-coolant accident [LOCA], the resulting substantial blockage of the sub-channel would restrict emergency core cooling. However, circumferential temperature gradients that would occur during a LOCA may significantly limit the average strain at failure. Understandably, the factors that control ballooning and rupture of fuel clad are required for the analysis of a LOCA. Considerable international effort has been spent on studying the deformation of Zircaloy fuel cladding under conditions that would occur during a LOCA. This effort has established a reasonable understanding of the factors that control the ballooning, failure time, and average failure strain of fuel cladding. In this paper, both the experimental and theoretical studies of the fuel clad ballooning are reviewed. [author]
State-of-the-technology review of fuel-cladding interaction International Nuclear Information System [INIS] Bailey, W.J.; Wilson, C.L.; MacGowan, L.J.; Pankaskie, P.J. 1977-12-01 A literature survey and a summarization of postulated fuel-cladding-interaction mechanisms and associated supportive data are reported. The results of that activity are described in the report and include comments on experience with power-ramped fuel, fuel-cladding mechanical interaction, stress-corrosion cracking and fission-product embrittlement, potential remedial actions, fuel-cladding-interaction mechanistic considerations, other ongoing programs, and related patents of interest. An assessment of the candidate fuel concepts to be evaluated as part of this program is provided
Advanced LWR Nuclear Fuel Cladding Development International Nuclear Information System [INIS] Bragg-Sitton, S.; Griffith, G. 2012-01-01 The Advanced Light Water Reactor [LWR] Nuclear Fuel Development Research and Development [R and D] Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental enhancements are required in the areas of nuclear fuel composition, cladding integrity, and fuel/cladding interaction to allow improved fuel economy via power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an 'accident tolerant' fuel system that would offer improved coping time under accident scenarios. In a staged development approach, the LWRS program will engage stakeholders throughout the development process to ensure commercial viability of the investigated technologies. Applying minimum performance criteria, several of the top-ranked materials and fabrication concepts will undergo a rigorous series of mechanical, thermal and chemical characterization tests to better define their properties and operating potential in a relatively low-cost, nonnuclear test series. A reduced number of options will be recommended for test rodlet fabrication and in-pile nuclear testing under steady-state, transient and accident conditions. [author]
Computer analysis of elongation of the WWER fuel rod claddings International Nuclear Information System [INIS] Scheglov, A.; Proselkov, V. 2008-01-01 In this paper description of mechanisms influencing changes of the WWER fuel cladding length and axial forces influencing fuel and cladding are presented. It is shown that shortening of the fuel claddings in case of high burnup can be explained by the change of the fuel and cladding reference state caused by reduction of the fuel rod power level - during reactor outages. It is noted that the presented calculated data are to be reviewed and interpreted as the preliminary results; further work is needed for their confirmation. [authors]
Performance of refractory alloy-clad fuel pins International Nuclear Information System [INIS] Dutt, D.S.; Cox, C.M.; Millhollen, M.K. 1984-12-01 This paper discusses objectives and basic design of two fuel-cladding tests being conducted in support of SP-100 technology development. Two of the current space nuclear power concepts use conventional pin type designs, where a coolant removes the heat from the core and transports it to an out-of-core energy conversion system. An extensive irradiation testing program was conducted in the 1950's and 1960's to develop fuel pins for space nuclear reactors. The program emphasized refractory metal clad uranium nitride [UN], uranium carbide [UC], uranium oxide [UO 2 ], and metal matrix fuels [UCZr and BeO-UO 2 ]. Based on this earlier work, studies presented here show that UN and UO 2 fuels in conjunction with several refractory metal cladding materials demonstrated high potential for meeting space reactor requirements and that UC could serve as an alternative but higher risk fuel
Accident tolerant fuel cladding development: Promise, status, and challenges Science.gov [United States] Terrani, Kurt A. 2018-04-01 The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.
Development Status of Accident Tolerant Fuel Cladding for LWRs Energy Technology Data Exchange [ETDEWEB] Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2016-10-15 Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels [ATFs] can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.
Establishment of Experimental Apparatus and Mechanical Test for SFR Metallic Fuel International Nuclear Information System [INIS] Kim, Sun Ki; Lee, Chong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Yoon Myung; Lee, Chan Bock 2010-12-01 U-Zr binary alloys and U-Zr-Ce ternary alloys as SFR surrogate metallic fuels were fabricated by a casting process. Tensile tests were performed to evaluate the mechanical properties of the fuels. As a results, the mechanical properties such as yield strength, ultimate tensile strength, and elongation were measured. In this report, these experimental results are presented
Fuel clad chemical interactions in fast reactor MOX fuels Energy Technology Data Exchange [ETDEWEB] Viswanathan, R., E-mail: rvis@igcar.gov.in 2014-01-15 Clad corrosion being one of the factors limiting the life of a mixed-oxide fast reactor fuel element pin at high burn-up, some aspects known about the key elements [oxygen, cesium, tellurium, iodine] in the clad-attack are discussed and many Fuel–Clad-Chemical-Interaction [FCCI] models available in the literature are also discussed. Based on its relatively superior predictive ability, the HEDL [Hanford Engineering Development Laboratory] relation is recommended: d/μm = [{0.507 ⋅ [B/[at.% fission]] ⋅ [T/K-705] ⋅ [[O/M]_i-1.935]} + 20.5] for [O/M]{sub i} ⩽ 1.98. A new model is proposed for [O/M]{sub i} ⩾ 1.98: d/μm = [B/[at.% fission]] ⋅ [T/K-800]{sup 0.5} ⋅ [[O/M]{sub i}-1.94] ⋅ [P/[W cm{sup −1}]]{sup 0.5}. Here, d is the maximum depth of clad attack, B is the burn-up, T is the clad inner surface temperature, [O/M]{sub i} is the initial oxygen-to-[uranium + plutonium] ratio, and P is the linear power rating. For fuels with [n[Pu]/n[M = U + Pu]] > 0.25, multiplication factors f are recommended to consider the potential increase in the depth of clad-attack.
Secondary hydriding of defected zircaloy-clad fuel rods International Nuclear Information System [INIS] Olander, D.R.; Vaknin, S. 1993-01-01 The phenomenon of secondary hydriding in LWR fuel rods is critically reviewed. The current understanding of the process is summarized with emphasis on the sources of hydrogen in the rod provided by chemical reaction of water [steam] introduced via a primary defect in the cladding. As often noted in the literature, the role of hydrogen peroxide produced by steam radiolysis is to provide sources of hydrogen by cladding and fuel oxidation that are absent without fission-fragment irradiation of the gas. Quantitative description of the evolution of the chemical state inside the fuel rod is achieved by combining the chemical kinetics of the reactions between the gas and the fuel and cladding with the transport by diffusion of components of the gas in the gap. The chemistry-gas transport model provides the framework into which therate constants of the reactions between the gases in the gap and the fuel and cladding are incorporated. The output of the model calculation is the H 2 0/H 2 ratio in the gas and the degree of claddingand fuel oxidation as functions of distance from the primary defect. This output, when combined with a criterion for the onset of massive hydriding of the cladding, can provide a prediction of the time and location of a potential secondary hydriding failure. The chemistry-gas transport model is the starting point for mechanical and H-in-Zr migration analyses intended to determine the nature of the cladding failure caused by the development of the massive hydride on the inner wall
Stainless steel clad for light water reactor fuels. Final report International Nuclear Information System [INIS] Rivera, J.E.; Meyer, J.E. 1980-07-01 Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps
Evaluation of Spent Fuel Recycling Scenario using Pyro-SFR related System International Nuclear Information System [INIS] Lee, Yong Kyo; Kim, Sang Ji; Kim, Young Jin 2014-01-01 It is needed to validate whether the recycling scenario connecting pyro-processing and sodium-cooled fast reactor[SFR] is promising or not. The latest technologies of pyro-processing are applied to SFR and the recycling scenario is evaluated through the SFR's performance analysis. The analyzed SFR is KALIMER-600 TRU burner which purpose is to transmute transuranics [TRU]. National policy of CANDU SF management has not been decided yet. However, the stored quantity of this SF is large enough not to be neglected. So this study includes additionally the recycling scenario of CANDU SF. Adopting the mass ratio of TRU and RE recovered in pyro-processing is 4 to 1 on PWR SF recycling, the sodium void reactivity is higher than design basis of metal fuel. So the current pyro-processing technology is may not be acceptable. If pyro-processing technology of CANDU SF is assumed to be the same as PWR's case, CANDU recycling scenario is acceptable. Transmutation performance is worse than PWR's, while the sodium void reactivity is within design limit

Evolutionary developments of advanced PWR nuclear fuels and cladding materials International Nuclear Information System [INIS] Kim, Kyu-Tae 2013-01-01 Highlights: • PWR fuel and cladding materials development processes are provided. • Evolution of PWR advanced fuel in U.S.A. and in Korea is described. • Cutting-edge design features against grid-to-rod fretting and debris are explained. • High performance data of advanced grids, debris filters and claddings are given. -- Abstract: The evolutionary developments of advanced PWR fuels and cladding materials are explained with outstanding design features of nuclear fuel assembly components and zirconium-base cladding materials. The advanced PWR fuel and cladding materials development processes are also provided along with verification tests, which can be used as guidelines for newcomers planning to develop an advanced fuel for the first time. The up-to-date advanced fuels with the advanced cladding materials may provide a high level of economic utilization and reliable performance even under current and upcoming aggressive operating conditions. To be specific, nuclear fuel vendors may achieve high fuel burnup capability of between 45,000 and 65,000 MWD/MTU batch average, overpower thermal margin of as much as 15% and longer cycle length up to 24 months on the one hand and fuel failure rates of around 10 −6 on the other hand. However, there is still a need for better understanding of grid-to-rod fretting wear mechanisms leading to major PWR fuel defects in the world and subsequently a driving force for developing innovative spacer grid designs with zero fretting wear-induced fuel failure
FEA stress analysis considering cavity formation of metallic fuel pin under transient state Energy Technology Data Exchange [ETDEWEB] Jung, Hyun-Woo; Oh, Young-Ryun; Kim, Yun-Jae [Korea University, Seoul [Korea, Republic of] 2016-05-15 The aim of this research is to study the stress state of the fuel and the cladding under transient state using the commercial finite element analysis software, ABAQUS v6.13. It is checked out that the gap distance between the fuel and the cladding is a major factor determining FCMI stress. In this regard, initial boundary condition of the fuel pin such as the initial gap distance should be set carefully when the stress analysis of the fuel pin under transient state is conducted. In case of simulating cavity formation, it is confirmed that the new cavity simulation model that elements in cavity region lose their stiffness is valid. There is a great deal of research into SFR, which is one of GEN IV reactors. When it comes to the accidents of SFR, there are two cases of accident process. One of them is In-pin process that molten fuel is discharged into upper plenum. The other is Ex-pin process that the molten fuel is discharged into coolant because of breakage of cladding.
Fuel cladding behavior under rapid loading conditions Science.gov [United States] Yueh, K.; Karlsson, J.; Stjärnsäter, J.; Schrire, D.; Ledergerber, G.; Munoz-Reja, C.; Hallstadius, L. 2016-02-01 A modified burst test [MBT] was used in an extensive test program to characterize fuel cladding failure behavior under rapid loading conditions. The MBT differs from a normal burst test with the use of a driver tube to simulate the expansion of a fuel pellet, thereby producing a partial strain driven deformation condition similar to that of a fuel pellet expansion in a reactivity insertion accident [RIA]. A piston/cylinder assembly was used to pressurize the driver tube. By controlling the speed and distance the piston travels the loading rate and degree of sample deformation could be controlled. The use of a driver tube with a machined gauge section localizes deformation and allows for continuous monitoring of the test sample diameter change at the location of maximum hoop strain, during each test. Cladding samples from five irradiated fuel rods were tested between 296 and 553 K and loading rates from 1.5 to 3.5/s. The test rods included variations of Zircaloy-2 with different liners and ZIRLO, ranging in burn-up from 41 to 74 GWd/MTU. The test results show cladding ductility is strongly temperature and loading rate dependent. Zircaloy-2 cladding ductility degradation due to operational hydrogen pickup started to recover at approximately 358 K for test condition used in the study. This recovery temperature is strongly loading rate dependent. At 373 K, ductility recovery was small for loading rates less than 8 ms equivalent RIA pulse width, but longer than 8 ms the ductility recovery increased exponentially with increasing pulse width, consistent with literature observations of loading rate dependent brittle-to-ductile [BTD] transition temperature. The cladding ductility was also observed to be strongly loading rate/pulse width dependent for BWR cladding below the BTD temperature and Pressurized Water Reactor [PWR] cladding at both 296 and 553 K.
Resistance welding of ODS cladding fuel a nuclear reactor of the fourth generation International Nuclear Information System [INIS] Corpace, F. 2011-01-01 ODS steels [Oxide Dispersion Strengthened] are candidate materials for fuel cladding in Sodium Fast Reactors [SFR], one of the studied concepts for the fourth generation of nuclear power plants. These materials possess good mechanical properties at high temperatures due to a dispersion of nano-meter-sized oxides into the matrix. Previous studies have shown that melting can induce a decrease in mechanical properties at high temperatures due to modifications of the nano-meter-sized oxide dispersion. Therefore the fusion welding techniques are not recommended and the solid state bonding has to be evaluated. This study is focused on resistance upset welding. Welding experiments and numerical simulations of the process are coupled in this thesis. All laboratory tests [experimental and numerical] are built using the experimental design method to evaluate the effects of the process parameters on the welding and on the weld. A 20Cr ODS steel is used for the experimental protocol. The first part is dedicated to the study of the influence of the process parameters on the welding process. The numerical simulations show that the welding steps can be divided in three stages. First, the contact temperature between the faying surfaces increases. The process is then driven in the second stage by the pieces geometry and especially the current constriction due to the thinness of the clad compared to the massive plug. Therefore, the heat generation is mainly located in the clad part out of the electrode leading to its collapse which is the third stage of the welding step. The evaluation of the process parameters influence on the physical phenomena [thermal, mechanical...] occurring during the welding step, allows adjusting them in order to influence thermal and mechanical solicitations undergone by the pieces during the welding process. The second part consists in studying the influence of physical phenomena on the welds. In the process parameter range, some welds exhibit compactness
POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION Vojtěch Caha 2016-12-01 Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux [CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations [e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.
Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification Energy Technology Data Exchange [ETDEWEB] Benoit, Timothy [Argonne National Lab. [ANL], Argonne, IL [United States]. Nuclear Engineering Division; Hlotke, John Daniel [Argonne National Lab. [ANL], Argonne, IL [United States]. Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. [ANL], Argonne, IL [United States]. Nuclear Engineering Division 2017-07-05 This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 [NQA-1] criteria and identifies and describes the quality assurance process[es] by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generated during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.
Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor International Nuclear Information System [INIS] Stauff, N. 2011-01-01 Compared with earlier plant designs [Phenix, Super-Phenix, EFR], Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide [U-Pu]O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide [U-Pu]C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW[e] break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel
Cladding failure margins for metallic fuel in the integral fast reactor International Nuclear Information System [INIS] Bauer, T.H.; Fenske, G.R.; Kramer, J.M. 1987-01-01 The reference fuel for Integral Fast Reactor [IFR] is a ternary U-Pu-Zr alloy with a low swelling austenitic or ferritic stainless steel cladding. It is known that low melting point eutectics may form in such metallic fuel-cladding systems which could contribute to cladding failure under accident conditions. This paper will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel
Advanced ceramic cladding for water reactor fuel International Nuclear Information System [INIS] Feinroth, H. 2000-01-01 Under the US Department of Energy's Nuclear Energy Research Initiatives [NERI] program, continuous fiber ceramic composites [CFCCs] are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined
Prevention of nuclear fuel cladding materials corrosion International Nuclear Information System [INIS] Yang, K.R.; Yang, J.C.; Lee, I.C.; Kang, H.D.; Cho, S.W.; Whang, C.K. 1983-01-01 The only way which could be performed by the operator of nuclear power plant to minimizing the degradation of nuclear fuel cladding material is to control the water quality of primary coolant as specified standard conditions which dose not attack the cladding material. If the water quality of reactor coolant does not meet far from the specification, the failure will occure not only cladding material itself but construction material of primary system which contact with the coolant. The corrosion product of system material are circulate through the whole primary system with the coolant and activated by the neutron near the reactor core. The activated corrosion products and fission products which released from fuel rod to the coolant, so called crud, will repeate deposition and redeposition continuously on the fuel rod and construction material surface. As a result we should consider heat transfer problem. In this study following activities were performed; 1. The crud sample was taken from the spent fuel rod surface of Kori unit one and analized for radioactive element and non radioactive chemical species. 2. The failure mode of nuclear fuel cladding material was estimated by the investigation of releasing type of fission products from the fuel rod to the reactor coolant using the iodine isotopes concentration of reactor coolants. 3. A study was carried out on the sipping test results of spent fuel and a discussion was made on the water quality control records through the past three cycle operation period of Kori unit one plant. [Author]
Corrosion behaviour of zircaloy 4 fuel rod cladding in EDF power plants Energy Technology Data Exchange [ETDEWEB] Romary, H; Deydier, D [EDF, Direction de l` Equipment SEPTEN, Villeurbanne [France] 1997-02-01 Since the beginning of the French nuclear program, a surveillance of fuel has been carried out in order to evaluate the fuel behaviour under irradiation. Until now, nuclear fuels provided by suppliers have met EDF requirements concerning fuel behaviour and reliability. But, the need to minimize the costs and to increase the flexibility of the power plants led EDF to the definition of new targets: optimization of the core management and fuel cycle economy. The fuel behaviour experience shows that some of these new requirements cannot be fully fulfilled by the present standard fuel due to some technological limits. Particularly, burnup enhancement is limited by the oxidation and the hydriding of the Zircaloy 4 fuel rod cladding. Also, fuel suppliers and EDF need to have a better knowledge of the Zy-4 cladding behaviour in order to define the existing margins and the limiting factors. For this reason, in-reactor fuel characterization programs have been set up by fuel suppliers and EDF for a few years. This paper presents the main results and conclusions of EDF experience on Zy-4 in-reactor corrosion behaviour. Data obtained from oxide layer or zirconia thickness measurements show that corrosion performance of Zy-4 fuel rod cladding, as irradiated until now in EDF reactors, is satisfactory but not sufficient to meet the future needs. The fuel suppliers propose in order to improve the corrosion resistance of fuel rod cladding, low tin Zy-4 cladding and then optimized Zy-4 cladding. Irradiation of these claddings are ongoing. The available corrosion data show the better in-reactor corrosion resistance of optimized Zy-4 fuel rod cladding compared to the standard Zy-4 cladding. The scheduled fuel surveillance program will confirm if the optimized Zy-4 fuel rod cladding will meet the requirements for the future high burnup and high flexibility fuel. [author]. 10 refs, 19 figs, 4 tabs.
Examination of Zircaloy-clad spent fuel after extended pool storage International Nuclear Information System [INIS] Bradley, E.R.; Bailey, W.J.; Johnson, A.B. Jr.; Lowry, L.M. 1981-09-01 This report presents the results from metallurgical examinations of Zircaloy-clad fuel rods from two bundles [0551 and 0074] of Shippingport PWR Core 1 blanket fuel after extended water storage. Both bundles were exposed to water in the reactor from late 1957 until discharge. The estimated average burnups were 346 GJ/kgU [4000 MWd/MTU] for bundle 0551 and 1550 GJ/kgU [18,000 MWd/MTU] for bundle 0074. Fuel rods from bundle 0551 were stored in deionized water for nearly 21 yr prior to examination in 1980, representing the world's oldest pool-stored Zircaloy-clad fuel. Bundle 0074 has been stored in deionized water since reactor discharge in 1964. Data from the current metallurgical examinations enable a direct assessment of extended pool storage effects because the metallurgical condition of similar fuel rods was investigated and documented soon after reactor discharge. Data from current and past examinations were compared, and no significant degradation of the Zircaloy cladding was indicated after almost 21 yr in water storage. The cladding dimensions and mechanical properties, fission gas release, hydrogen contents of the cladding, and external oxide film thicknesses that were measured during the current examinations were all within the range of measurements made on fuel bundles soon after reactor discharge. The appearance of the external surfaces and the microstructures of the fuel and cladding were also similar to those reported previously. In addition, no evidence of accelerated corrosion or hydride redistribution in the cladding was observed
Management of cladding hulls and fuel hardware International Nuclear Information System [INIS] 1985-01-01 The reprocessing of spent fuel from power reactors based on chop-leach technology produces a solid waste product of cladding hulls and other metallic residues. This report describes the current situation in the management of fuel cladding hulls and hardware. Information is presented on the material composition of such waste together with the heating effects due to neutron-induced activation products and fuel contamination. As no country has established a final disposal route and the corresponding repository, this report also discusses possible disposal routes and various disposal options under consideration at present
Fuel-clad heat transfer coefficient of a defected fuel rod International Nuclear Information System [INIS] Bruet, M.; Stora, J.P. 1976-01-01 A special rod has been built with a stack of UO 2 pellets inside a thick zircaloy clad. The atmosphere inside the fuel rod can be changed and particularly the introduction of water is possible. The capsule was inserted in the Siloe pool reactor in a special device equipped with a neutron flux monitor. The fuel centerline temperature and the temperature at a certain radius of the clad were recorded by two thermocouples. The temperature profiles in the fuel and in the cladding have been calculated and then the heat transfer coefficient. In order to check the proper functioning of the device, two runs were successively achieved with a helium atmosphere. Then the helium atmosphere inside the fuel rod was removed and replaced by water. The heat transfer coefficients derived from the measurements at low power level are in agreement with the values given by the model based on thermal conductivity. However, for higher power levels, the heat transfer coefficients become higher than those based on the calculated gap
Irradiation effects on mechanical properties of fuel element cladding from thermal reactors International Nuclear Information System [INIS] Chatterjee, S. 2005-01-01 During reactor operation, UO 2 expands more than the cladding tube [Zirconium alloys for thermal reactors], is hotter, cracks and swells. The fuel therefore will interact with the cladding, resulting in straining of the later. To minimize the possibility of rupture of the cladding, ideally it should have good ductility as well as high strength. However, the ductility reduces with increase in fuel element burn-up. Increased burn-up also increases swelling of the fuel, leading to increased contact pressure between the fuel and the cladding tube. This would cause strains to be concentrated over localized regions of the cladding. For fuel elements burnup exceeding 40 GWd/T, the contribution of embrittlement due to hydriding, and the increased possibility of embrittlement due to stress corrosion cracking, also need to be considered. In addition to the tensile properties, the other mechanical properties of interest to the performance of cladding tube in an operating fuel element are creep rate and fatigue endurance. Irradiation is reported to have insignificant effect on high cycle endurance limit, and fatigue from fuel element vibration is most unlikely, to be life limiting. Even though creep rates due to irradiation are reported to increase by an order of magnitude, the cladding creep ductility would be so high that creep type failures in fuel element would be most improbable. Thus, the most important limiting aspect of mechanical performance of fuel element cladding has been recognized as the tensile ductility resulting from the stress conditions experienced by the cladding. Some specific fission products of threshold amount [if] deposited on the cladding, and hydride morphology [e.g. hydride lenses]. The presentation will brief about irradiation damage in cladding materials and its significance, background of search for better Zirconium alloys as cladding materials, and elaborate on the types of mechanical tests need to be conducted for the evaluation of claddings
Delayed hydride cracking of Zircaloy-4 fuel cladding International Nuclear Information System [INIS] Pizarro, Luis M.; Fernandez, Silvia; Lafont, Claudio; Mizrahi, Rafael; Haddad, Roberto 2007-01-01 Crack propagation rates, grown by the delayed hydride cracking mechanism, were measured in Zircaloy-4 fuel cladding, according to a Coordinated Research Project [CRP] sponsored by the International Atomic Energy Agency [IAEA]. During the first stage of the program a Round Robin Testing was performed on fuel cladding samples provided by Studsvik [Sweden], of the type used in PWR reactors. Crack growth in the axial direction is obtained through the specially developed 'pin load testing' [PLT] device. In these tests, crack propagation rates were determined at 250 C degrees on several samples of the material described above, obtaining a mean value of about 2.5 x 10 -8 m s -1 . The results were analyzed and compared satisfactorily with those obtained by the other laboratories participating in the CRP. At the present moment, similar tests on CANDU and Atucha I type fuel cladding are being performed. It is thought that the obtained results will give valuable information concerning the analysis of possible failures affecting fuel cladding under reactor operation. [author] [es
Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels International Nuclear Information System [INIS] Lu, Hongbing; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott 2014-01-01 Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method [MPM]; 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear
Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels Energy Technology Data Exchange [ETDEWEB] Lu, Hongbing [Univ. of Texas, Austin, TX [United States]; Bukkapatnam, Satish; Harimkar, Sandip; Singh, Raman; Bardenhagen, Scott 2014-01-09 Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method [MPM]; 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear
The prediction problems of VVER fuel element cladding failure theory International Nuclear Information System [INIS] Pelykh, S.N.; Maksimov, M.V.; Ryabchikov, S.D. 2016-01-01 Highlights: • Fuel cladding failure forecasting is based on the fuel load history and the damage distribution. • The limit damage parameter is exceeded, though limit stresses are not reached. • The damage parameter plays a significant role in predicting the cladding failure. • The proposed failure probability criterion can be used to control the cladding tightness. - Abstract: A method for forecasting of VVER fuel element [FE] cladding failure due to accumulation of deformation damage parameter, taking into account the fuel assembly [FA] loading history and the damage parameter distribution among FEs included in the FA, has been developed. Using the concept of conservative FE groups, it is shown that the safety limit for damage parameter is exceeded for some FA rearrangement, though the limits for circumferential and equivalent stresses are not reached. This new result contradicts the wide-spread idea that the damage parameter value plays a minor role when estimating the limiting state of cladding. The necessary condition of rearrangement algorithm admissibility and the criterion for minimization of the probability of cladding failure due to damage parameter accumulation have been derived, for using in automated systems controlling the cladding tightness.
Fuel cladding mechanical properties for transient analysis International Nuclear Information System [INIS] Johnson, G.D.; Hunter, C.W.; Hanson, J.E. 1976-01-01 Out-of-pile simulated transient tests have been conducted on irradiated fast-reactor fuel pin cladding specimens at heating rates of 10 0 F/s [5.6 0 K/s] and 200 0 F/s [111 0 K/s] to generate mechanical property information for use in describing cladding behavior during off-normal events. Mechanical property data were then analyzed, applying the Larson-Miller Parameter to the effects of heating rate and neutron fluence. Data from simulated transient tests on TREAT-tested fuel pins demonstrate that Plant Protective System termination of 3$/s transients prevents significant damage to cladding. The breach opening produced during simulated transient testing is shown to decrease in size with increasing neutron fluence

Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor International Nuclear Information System [INIS] Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B. 2009-03-01 The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide [MOX] fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However
Characteristics of hydride precipitation and reorientation in spent-fuel cladding International Nuclear Information System [INIS] Chung, H. M.; Strain, R. V.; Billone, M. C. 2000-01-01 The morphology, number density, orientation, distribution, and crystallographic aspects of Zr hydrides in Zircaloy fuel cladding play important roles in fuel performance during all phases before and after discharge from the reactor, i.e., during normal operation, transient and accident situations in the reactor, temporary storage in a dry cask, and permanent storage in a waste repository. In the past, partly because of experimental difficulties, hydriding behavior in irradiated fuel cladding has been investigated mostly by optical microscopy [OM]. In the present study, fundamental metallurgical and crystallographic characteristics of hydride precipitation and reorientation were investigated on the microscopic level by combined techniques of OM and transmission electron and scanning electron microscopy [TEM and SEM] of spent-fuel claddings discharged from several boiling and pressurized water reactors [BWRs and PWRs]. Defueled sections of standard and Zr-lined Zircaloy-2 fuel claddings, irradiated to fluences of ∼3.3 x 10 21 n cm -2 and ∼9.2 x 10 21 n cm -2 [E > 1 MeV], respectively, were obtained from spent fuel rods discharged from two BWRs. Sections of standard and low-tin Zircaloy-4 claddings, irradiated to fluences of ∼4.4 x 10 21 n cm -2 , ∼5.9 x 10 21 n cm -2 , and ∼9.6 x 10 21 n cm -2 [E > 1 MeV] in three PWRs, were also obtained. Microstructural characteristics of hydrides were analyzed in as-irradiated condition and after gas-pressurization-burst or expanding-mandrel tests at 292-325 C in Ar for some of the spent-fuel claddings. Analyses were also conducted of hydride habit plane, morphology, and reorientation characteristics on unirradiated Zircaloy-4 cladding that contained dense radial hydrides. Reoriented hydrides in the slowly cooled unirradiated cladding were produced by expanding-mandrel loading
Corrosion effect of fast reactor fuel claddings on their mechanical properties International Nuclear Information System [INIS] Davydov, E.F.; Krykov, F.N.; Shamardin, V.K. 1985-01-01 Fast reactor fuel cladding corrosion effect on its mechanical properties was investigated. UO 2 fuel elements were irradiated in the BOP-60 reactor at the linear heat rate of 42 kw/m. Fuel cladding is made of stainless steel OKh16N15M3BR. Calculated maximum cladding temperature is 920 K. Neutron fluence in the central part of fuel elements is 6.3x10 26 m+H- 2 . To investigate the strength changes temperature dependence of corrossion depth, cladding strength reduction factors was determined. Samples plasticity reduction with corrosion layer increase is considered to be a characteristic feature
General considerations on the oxide fuel-cladding chemical interaction International Nuclear Information System [INIS] Pascard, R. 1977-01-01 Since the very first experimental irradiations in thermal reactors, performed in view of the future Rapsodie fuel general study, corrosion cladding anomalies were observed. After 10 years of Rapsodie and more than two years of Phenix, performance brought definite confirmation of the chemical reactions between the irradiated fuel and cladding. That is the reason for which the fuel designers express an urgent need for determining the corrosion rates. Semi-empirical laws and mechanisms describing corrosion processes are proposed. Erratic conditions for appearance of the oxide-cladding corrosion are stressed upon. Obviously such a problem can be fully appreciated only by a statistical approach based on a large number of observations on the true LMFBR fuel pins
Temperature measurements of the aluminium claddings of fuel elements in nuclear reactor International Nuclear Information System [INIS] Chen Daolong 1986-01-01 A method for embedding the sheathed thermocouples in the aluminium claddings of some fuel elements of experimental reactors by ultrasonic welding technique is described. The measurement results of the cladding temperature of fuel elements in reactors are given. By means of this method, the joint between the sheathed thermocouples and the cladding of fuel elements can be made very tight, there are no bulges on the cladding surfaces, and the sheathed thermocouples are embedded strongly and reliably. Therefore an essential means is provided for acquiring the stable and dynamic state data of the cladding temperature of in-core fuel elements
Technical basis for storage of Zircaloy-clad spent fuel in inert gases International Nuclear Information System [INIS] Johnson, A.B. Jr.; Gilbert, E.R. 1983-09-01 The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. Dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570 0 C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed [one PWR rod exposed to an air cover gas failed at about 270 0 C]. Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380 0 C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400 0 C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved
Scientific basis for storage criteria for interim dry storage of aluminum-clad fuels International Nuclear Information System [INIS] Sindelar, R.L.; Peacock, H.B. Jr.; Lam, P.S.; Iyer, N.C.; Louthan, M.R. Jr.; Murphy, J.R. 1996-01-01 An engineered system for dry storage of aluminum-clad foreign and domestic research reactor spent fuel owned by the US Department of Energy is being considered to store the fuel up to a nominal period of 40 years prior to ultimate disposition. Scientifically-based criteria for environmental limits to drying and storing the fuels for this system are being developed to avoid excessive degradation in sealed and non-sealed [open to air] dry storage systems. These limits are based on consideration of degradation modes that can cause loss of net section of the cladding, embrittlement of the cladding, distortion of the fuel, or release of fuel and fission products from the fuel/clad system. Potential degradation mechanisms include corrosion mechanisms from exposure to air and/or sources of humidity, hydrogen blistering of the aluminum cladding, distortion of the fuel due to creep, and interdiffusion of the fuel and fission products with the cladding. The aluminum-clad research reactor fuels are predominantly highly-enriched aluminum uranium alloy fuel which is clad with aluminum alloys similar to 1100, 5052, and 6061 aluminum. In the absence of corrodant species, degradation due to creep and diffusion mechanisms limit the maximum fuel storage temperature to 200 C. The results of laboratory scale corrosion tests indicate that this fuel could be stored under air up to 200 C at low relative humidity levels [< 20%] to limit corrosion of the cladding and fuel [exposed to the storage environment through assumed pre-existing pits in the cladding]. Excessive degradation of fuels with uranium metal up to 200 C can be avoided if the fuel is sufficiently dried and contained in a sealed system; open storage can be achieved if the temperature is controlled to avoid excessive corrosion even in dry air
Modelling of pellet-cladding interaction for PWRs reactors fuel rods International Nuclear Information System [INIS] Esteves, A.M. 1991-01-01 The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyzes the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. Linear and non-linear material behaviors are allowed. Elastic, plastic and creep behaviors are considered for the cladding materials. The modelling is applied to Angra-II fuel rod design. The results are analyzed and compared. [author]
Probabilistic assessment of spent-fuel cladding breach International Nuclear Information System [INIS] Foadian, H.; Rashid, Y.R.; Seager, K.D. 1991-01-01 A methodology for determining the probability spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated Zircaloy properties such as ductility and fracture toughness, and for fuel rod initial conditions such as manufacturing flaws and PCI part-wall cracks. Example analyses are used to illustrate the implementation of this methodology for a BWR [GE 7 x 7] and a PWR [B ampersand W 15 x 15] assembly. The cladding breach probabilities for each assembly are tabulated for regulatory normal and accident transport conditions including fire
Probabilistic assessment of spent-fuel cladding breach International Nuclear Information System [INIS] Foadian, H.; Rashid, Y.R.; Seager, K.D. 1992-01-01 In this paper a methodology for determining the probability of spent-fuel cladding breach due to normal and accident class B cask transport conditions is introduced. This technique uses deterministic stress analysis results as well as probabilistic cladding material properties, initial flaws, and breach criteria. Best estimates are presented for the probability distributions of irradiated Zircaloy properties such as ductility and fracture toughness, and for fuel rod initial conditions such as manufacturing flaws and PCI part-wall cracks. Example analyses are used to illustrate the implementation of this methodology for a BWR [GE 7 x 7] and a PWR [B and W 15 x 15] assembly. The cladding breach probabilities for each assembly are tabulated for regulatory normal and accident transport conditions including fire
Preliminary study of mechanical behavior for Cr coated Zr-4 Fuel Cladding Energy Technology Data Exchange [ETDEWEB] Kim, Do-Hyoung; Kim, Hak-Sung [Hanyang Univ., Seoul [Korea, Republic of]; Kim, Hyo-Chan; Yang, Yong-Sik [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2015-05-15 To decrease the oxidation rate of Zr-based alloy components, many concepts of accident tolerant fuel [ATF] such as Mo-Zr cladding, SiC/SiCf cladding and iron-based alloy cladding are under development. One of the promised concept is the coated cladding which can remarkably increase the corrosion and wear resistance. Recently, KAERI is developing the Cr coated Zircaloy cladding as accident tolerance cladding. To coat the Cr powder on the Zircaloy, 3D laser coating technology has been employed because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. Therefore, for this work, the mechanical integrity of Cr coated Zircaloy should be evaluated to predict the safety of fuel cladding during the operating or accident of nuclear reactor. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis [FEA]. The ring compression test [RCT] of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr, which were referred from the literatures and experimental reports. In this work, the mechanical behavior of the Cr coated Zircaloy cladding has been studied by using finite element analysis [FEA]. The ring compression test [RCT] of fuel cladding was simulated to evaluate the validity of mechanical properties of Zr-4 and Cr. The pellet-clad mechanical interaction [PCMI] properties of Cr coated Zr-4 cladding were investigated by thermo-mechanical finite element analysis [FEA] simulation. The mechanical properties of Zr-4 and Cr was validated by simulation of ring compression test [RCT] of fuel cladding.
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].
Fabrication of oxide dispersion strengthened ferritic clad fuel pins International Nuclear Information System [INIS] Zirker, L.R.; Bottcher, J.H.; Shikakura, S.; Tsai, C.L. 1991-01-01 A resistance butt welding procedure was developed and qualified for joining ferritic fuel pin cladding to end caps. The cladding are INCO MA957 and PNC ODS lots 63DSA and 1DK1, ferritic stainless steels strengthened by oxide dispersion, while the end caps are HT9 a martensitic stainless steel. With adequate parameter control the weld is formed without a residual melt phase and its strength approaches that of the cladding. This welding process required a new design for fuel pin end cap and weld joint. Summaries of the development, characterization, and fabrication processes are given for these fuel pins. 13 refs., 6 figs., 1 tab
Creep collapse of TAPS fuel cladding International Nuclear Information System [INIS] Chaudhry, S.M.; Anand, A.K. 1975-01-01 Densification of UO 2 can cause axial gaps between fuel pelets and cladding in unsupported [internally] at these regions. An analysis is carried out regarding the possibility of creep collapse in these regions. The analysis is based on Timoshenko's theory of collapse. At various times during the residence of fuel in reactor following parameters are calculated : [1] inelastic collapse of perfectly circular tubes [2] plastic instability in oval tubes [3] effect of creep on ovality. Creep is considered to be a non-linear combination of the following : [a] thermal creep [b] intresenic creep [c] stress aided radiation enhanced [d] stress free growth [4] Critical pressure ratio. The results obtained are compared with G.E. predictions. The results do not predict collapse of TAPS fuel cladding for five year residence time. [author]
Gap conductance in Zircaloy-clad LWR fuel rods International Nuclear Information System [INIS] Ainscough, J.B. 1982-04-01 This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed
Influence of the fuel operational parameters on the aluminium cladding quality of discharged fuel Energy Technology Data Exchange [ETDEWEB] Chwaszczewski, S.; Czajkowski, W.; Borek-Kruszewska, E. [Institute of Atomic Energy, Otwock Swierk [POLAND] 2002-07-01 In the last two years, the new MR6 type fuel containing 1550 g of U with 36% enrichment has been loaded into MARIA reactor core. Its aluminium cladding thickness is 0,6 mm and typical burnup -about 4080 MWh [as compared to 2880 MWh for the 80% enriched fuel used]. However, increased fission product release from these assemblies was observed near the end of its operational time. The results presented earlier [1] show that the corrosion behaviour of aluminium cladding depends on the conditions of fuel operation in the reactor. The corrosion process in the aluminum of fuel cladding proceeds faster then in the aluminum of constructional elements. This tendency was also observed in MR-6/80% and in WWR- SM fuel assemblies. Therefore the visual tests of discharged MR-6/36% fuel elements were performed. Some change of appearance of aluminum cladding was observed, especially in the regions with large energy generation i.e. in the centre of reactor core and in the strong horizontal gradient of neutron flux. In the present paper, the results of visual investigation of discharged fuel assemblies are presented. The results of the investigation are correlated with the operational parameters. [author]
Out-pile Test of Double Cladding Fuel Rod Mockups for a Nuclear Fuel Irradiation Test Energy Technology Data Exchange [ETDEWEB] Sohn, Jaemin; Park, Sungjae; Kang, Younghwan; Kim, Harkrho; Kim, Bonggoo; Kim, Youngki [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2008-05-15 An instrumented capsule for a nuclear fuel irradiation test has been developed to measure fuel characteristics, such as a fuel temperature, internal pressure of a fuel rod, a fuel pellet elongation and a neutron flux during an irradiation test at HANARO. In the future, nuclear fuel irradiation tests under a high temperature condition are expected from users. To prepare for this request, we have continued developing the technology for a high temperature nuclear fuel irradiation test at HANARO. The purpose of this paper is to verify the possibility that the temperature of a nuclear fuel can be controlled at a high temperature during an irradiation test. Therefore we designed and fabricated double cladding fuel rod mockups. And we performed out-pile tests using these mockups. The purposes of a out-pile test is to analyze an effect of a gap size, which is between an outer cladding and an inner cladding, on the temperature and the effect of a mixture ratio of helium gas and neon gas on the temperature. This paper presents the design and fabrication of double cladding fuel rod mockups and the results of the out-pile test.
WWER water chemistry related to fuel cladding behaviour Energy Technology Data Exchange [ETDEWEB] Kysela, J; Zmitko, M [Nuclear Research Inst. plc., Rez [Czech Republic]; Vrtilkova, V [Nuclear Fuel Inst., Prague [Czech Republic] 1997-02-01 Operational experience in WWER primary water chemistry and corrosion related to the fuel cladding is reviewed. Insignificant corrosion of fuel cladding was found which is caused by good corrosion resistance of Zr1Nb material and relatively low coolant temperature at WWER-440 reactor units. The differences in water chemistry control is outlined and an attention to the question of compatibility of Zircaloys with WWER water chemistry is given. Some results of research and development in field of zirconium alloy corrosion behaviour are discussed. Experimental facility for in-pile and out-of-pile cladding material corrosion testing is shown. [author]. 14 refs, 5 figs, 3 tabs.
Development of metallic fuel fabrication - A study on the interdiffusion behavior between ternary metallic fuel and cladding materials Energy Technology Data Exchange [ETDEWEB] Lee, Byung Soo; Seol, Kyung Won; Shon, In Jin [Chonbuk National University, Chonju [Korea] 1999-04-01 To study a new ternary metallic fuel for liquid metal reactor, various U-Zr-X alloys have been made by induction melting. The specimens were prepared for thermal stability tests at 630 deg. C upto 5000 hours in order to estimate the decomposition of the lamellar structure. Interdiffusion studies were carried out at 700 deg. C for 200 hours for the diffusion couples assembled with U-Zr-X ternary fuel versus austenitic stainless steel D9 and martensitic stainless steel HT9, respectively, to investigate the fuel-cladding compatibility. The ternary alloy, especially U-Zr-Mo and U-Zr-Nb alloys showed relatively good thermal stability as long as 5000hrs at 630 deg. C. From the composition profiles of the interdiffusion study, Fe penetrated deeper to the fuel side than other cladding elements such as Ni and Cr, whereas U did to the cladding side of fuel elements in the fuel/D9 couples. On the contrary, the reaction layers of Fuel/HT9 couple were thinner than that of Fuel/D9 couples and were less affected by cladding element, which was believed to be due to Zr rich layer between the fuel-cladding interface. HT9 is considered to be superior to D9 and a favorable choice as a cladding material in terms of fuel-cladding compatibility. 21 refs., 24 figs., 7 tabs. [Author]
BISON Fuel Performance Analysis of FeCrAl cladding with updated properties Energy Technology Data Exchange [ETDEWEB] Sweet, Ryan [Oak Ridge National Lab. [ORNL], Oak Ridge, TN [United States]; George, Nathan M. [Oak Ridge National Lab. [ORNL], Oak Ridge, TN [United States]; Terrani, Kurt A. [Oak Ridge National Lab. [ORNL], Oak Ridge, TN [United States]; Wirth, Brian [Oak Ridge National Lab. [ORNL], Oak Ridge, TN [United States] 2016-08-30 In order to improve the accident tolerance of light water reactor [LWR] fuel, alternative cladding materials have been proposed to replace zirconium [Zr]-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum [FeCrAl] alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys [namely Alkrothal 720 and APMT] as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime [~4 cycles] for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and

Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses Energy Technology Data Exchange [ETDEWEB] Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David 2016-04-17 In 2015, as part of a Regulatory Technology Development Plan [RTDP] effort for sodium-cooled fast reactors [SFRs], Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment [SRE] facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II [EBR-II] from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility [TREAT] in the mid-1980s.
Corrosion of research reactor aluminium clad spent fuel in water International Nuclear Information System [INIS] 2009-12-01 A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools [or basins] close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project [CRP] on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water [Phase I] initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water [Phase II], to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors [2001-2006] was corrosion monitoring and surveillance of research
First results on the effect of fuel-cladding eccentricity International Nuclear Information System [INIS] Panka, I.; Kereszturi, A. 2009-01-01 In the traditional fuel-behaviour or hot channel calculations it is assumed that the fuel pellet is centered within the clad. However, in the real life the pellet could be positioned asymmetrically within the clad, which leads to asymmetric gap conductance and therefore it is worthwhile to investigate the magnitude of the effect on maximal fuel temperature and surface heat flux. In this paper our first experiences are presented on this topic. [Authors]
Demonstration of fuel resistant to pellet-cladding interaction. Phase I. Final report International Nuclear Information System [INIS] Rosenbaum, H.S. 1979-03-01 This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction [PCI]. Two fuel concepts are being developed for possible demonstration within this program: [a] Cu-barrier fuel, and [b] Zr-liner fuel. These advanced fuels [known collectively as barrier fuels] have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress, and reactive fission products during reactor service. This is the final report for PHASE 1 of this program. Support tests have shown that the barrier fuel resists PCI far better than does the conventional Zircaloy-clad fuel. Power ramp tests thus far have shown good PCI resistance for Cu-barrier fuel at burnup > 12 MWd/kg-U and for Zr-liner fuel \> 16 MWd/kg-U. The program calls for continued testing to still higher burnup levels in PHASE 2
Thermochemical aspects of fuel-cladding and fuel-coolant interactions in LMFBR oxide fuel pins International Nuclear Information System [INIS] Adamson, M.G.; Aitken, E.A.; Caputi, R.W.; Potter, P.E.; Mignanelli, M.A. 1979-01-01 This paper examines several thermochemical aspects of the fuel-cladding, fuel-coolant and fuel-fission product interactions that occur in LMFBR austenitic stainless steel-clad mixed [U,Pu]-oxide fuel pins during irradiation under normal operating conditions. Results are reported from a variety of high temperature EMF cell experiments in which continuous oxygen activity measurements on reacting and equilibrium mixtures of metal oxides and [excess] liquid alkali metal [Na, K, Cs] were performed. Oxygen potential and 0:M thresholds for Na-fuel reactions are re-evaluated in the light of new measurements and newly-assessed thermochemical data, and the influence on oxygen potential of possible U-Pu segregation between oxide and urano-plutonate [equilibrium] phases has been analyzed. [orig./RW] [de
Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study Energy Technology Data Exchange [ETDEWEB] Kristine Barrett; Shannon Bragg-Sitton 2012-09-01 The Advanced Light Water Reactor [LWR] Nuclear Fuel Development Research and Development [R&D] Pathway encompasses strategic research focused on improving reactor core economics and safety margins through the development of an advanced fuel cladding system. To achieve significant operating improvements while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are required in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up allowance while potentially improving safety margin through the adoption of an “accident tolerant” fuel system that would offer improved coping time under accident scenarios. With a development time of about 20 – 25 years, advanced fuel designs must be started today and proven in current reactors if future reactor designs are to be able to use them with confidence.
Fuel Performance Calculations for FeCrAl Cladding in BWRs Energy Technology Data Exchange [ETDEWEB] George, Nathan [Univ. of Tennessee, Knoxville, TN [United States]. Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN [United States]. Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN [United States]. Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN [United States]. Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. [ORNL], Oak Ridge, TN [United States]; Worrall, Andrew [Oak Ridge National Lab. [ORNL], Oak Ridge, TN [United States] 2015-01-01 This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor [BWR] cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL [formerly Peregrine] expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod [e.g., strains, centerline fuel temperature, and time to gap closure] were investigated and contrasted.
Study of pellet clad interaction defects in Dresden-3 fuel rods International Nuclear Information System [INIS] Pasupathi, V.; Perrin, J.S. 1979-01-01 During Cycle-3 operation of Dresden-3, fuel rod failures occurred following a transient power increase. Ten fuel rods from five of the leaking fuel assemblies were examined at Battelle's Columbus Laboratory and General Electric-Vallecitos Nuclear Center. Examinations consisted of nondestructive and destructive methods including metallography and scanning electron microscopy [SEM]. Results showed the cause of fuel rod failure to be pellet clad interaction involving stress corrosion cracking. Results of SEM studies of the cladding crack surfaces and deposits on clad inner surfaces were in agreement with those reported by other investigators
Fuel chemistry and pellet-clad interaction related to high burnup fuel. Proceedings of the technical committee International Nuclear Information System [INIS] 2000-10-01 The purpose of the meeting was to review new developments in clad failures. Major findings regarding the causes of clad failures are presented in this publication, with the main topics being fuel chemistry and fission product behaviour, swelling and pellet-cladding mechanical interaction, cladding failure mechanism at high burnup, thermal properties and fuel behaviour in off-normal conditions. This publication contains 17 individual presentations delivered at the meeting; each of them was indexed separately
Performance of HT9 clad metallic fuel at high temperature International Nuclear Information System [INIS] Pahl, R.G.; Lahm, C.E.; Hayes, S.L. 1992-01-01 Steady-state testing of HT9 clad metallic fuel at high temperatures was initiated in EBR-II in November of 1987. At that time U-10 wt. % Zr fuel clad with the low-swelling ferritic/martensitic alloy HT9 was being considered as driver fuel options for both EBR-II and FFTF. The objective of the X447 test described here was to determine the lifetime of HT9 cladding when operated with metallic fuel at beginning of life inside wall temperatures approaching ∼660 degree C. Though stress-temperature design limits for HT9 preclude its use for high burnup applications under these conditions due to excessive thermal creep, the X447 test was carried out to obtain data on high temperature breach phenomena involving metallic fuel since little data existed in that area
The quest for safe and reliable fuel cladding materials Energy Technology Data Exchange [ETDEWEB] Pino, Eddy S.; Abe, Alfredo Y., E-mail: eddypino132@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares [IPEN/CNEN-SP], Sao Paulo, SP [Brazil]; Giovedi, Claudia, E-mail: claudia.giovedi@labrisco.usp.br [Universidade de Sao Paulo [POLI/USP], Sao Paulo, SP [Brazil]. Lab. de Analise, Avaliacao e Gerenciamento de Risco 2015-07-01 The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels [ATF]. This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide [SiC] materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR [Pressurized Water Reactor] registering a good performance. Then, alternated cladding materials such as iron based alloys [304, 310, 316, 347] should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. [author]
The quest for safe and reliable fuel cladding materials International Nuclear Information System [INIS] Pino, Eddy S.; Abe, Alfredo Y.; Giovedi, Claudia 2015-01-01 The tragic Fukushima Daiichi Nuclear Plant accident of March, 2011, has brought great unrest and challenge to the nuclear industry, which, in collaboration with universities and nuclear research institutes, is making great efforts to improve the safety in nuclear reactors developing accident tolerant fuels [ATF]. This involves the study of different materials to be applied as cladding and, also, the improvement in the fuel properties in order to enhance the fuel performance and safety, specifically under accident conditions. Related to the cladding, iron based alloys and silicon carbide [SiC] materials have been studied as a good alternative. In the case of austenitic stainless steel, there is the advantage that the austenitic stainless steel 304 was used as cladding material in the first PWR [Pressurized Water Reactor] registering a good performance. Then, alternated cladding materials such as iron based alloys [304, 310, 316, 347] should be used to replace the zirconium-based alloys in order to improve safety. In this paper, these cladding materials are evaluated in terms of their physical and chemical properties; among them, strength and creep resistance, thermal conductivity, thermal stability and corrosion resistance. Additionally, these properties are compared with those of conventional zirconium-based alloys, the most used material in actual PWR, to assess the advantages and disadvantages of each material concerning to fuel performance and safety contribution. [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
Influence of pellet-clad-gap-size on LWR fuel rod performance International Nuclear Information System [INIS] Brzoska, B.; Fuchs, H.P.; Garzarolli, F.; Manzel, R. 1979-01-01 The as-fabricated pellet-clad-gap size varies due to fabricational tolerances of the cladding inner diameter and the pellet outer diameter. The consequences of these variations on the fuel rod behaviour are analyzed using the KWU fuel rod code CARO. The code predictions are compared with experimental results of special pathfinder test fuel rods irradiated in the OBRIGHEIM nuclear power plant. These test fuel rods include gap sizer in the range of 140 μm to 270 μm, prepressurization between 13 bar to 36 bar and Helium and Argon fill gases irradiated up to a local burnup of 35 MWd/kg[U]. Post irradiation examination were performed at different burnups. CARC calculations have been performed with special emphasis in cladding creep down, fission gas release and pellet clad gap closure. [orig.]
In-reactor performance of methods to control fuel-cladding chemical interaction International Nuclear Information System [INIS] Weber, E.T.; Gibby, R.L.; Wilson, C.N.; Lawrence, L.A.; Adamson, M.G. 1979-01-01 Inner surface corrosion of austenitic stainless steel cladding by oxygen and reactive fission product elements requires a 50 μm wastage allowance in current FBR reference oxide fuel pin design. Elimination or reduction of this wastage allowance could result in better reactor efficiency and economics through improvements in fuel pin performance and reliability. Reduction in cladding thickness and replacement of equivalent volume with fuel result in improved breeding capability. Of the factors affecting fuel-cladding chemical interaction [FCCI], oxygen activity within the fuel pin can be most readily controlled and/or manipulated without degrading fuel pin performance or significantly increasing fuel fabrication costs. There are two major approaches to control oxygen activity within an oxide fuel pin: [1] control of total oxygen inventory and chemical activity [Δ anti GO 2 ] by use of low oxygen-to-metal ratio [O/M] fuel; and [2] incorporation of a material within the fuel pin to provide in-situ control of oxygen activity [Δ anti GO 2 ] and fixation of excess oxygen prior to, or in preference to reaction with the cladding. The paper describes irradiation tests which were conducted in EBR-II and GETR incorporating oxygen buffer/getter materials and very low O/M fuel to control oxygen activity in sealed fuel pins
Mechanical and temperature contact in fuel rod cladding International Nuclear Information System [INIS] Fredriksson, B.E.; Rydholm, S.G. 1977-01-01 The paper presents results for the effect of different types of slip rules on the contact stress distribution. It is shown that the contact shear stress is smaller for the hardening model than for the ideal model. It is also shown that a crack in the fuel increases the contact stresses and that at temperature decrease high tensile stresses arise after eventual welding. It is also shown how particles between fuel and cladding influence the stresses. Also here the effect of eventual welding is studied. The present method is well suited to study cracks and crack propagation. The surfaces of the existing cracks are defined as contact surfaces and the crack extension work is calculated by releasing the nodes at the crack tip. As the crack surfaces are defined as contact surfaces eventual crack closure is automatically taken into account. Crack extension work is calculated for existing cracks in the cladding. It is shown that cracks in the fuel and particles between fuel and cladding will increase the crack extension work
Fuel-cladding chemical interaction correlation for mixed-oxide fuel pins International Nuclear Information System [INIS] Lawrence, L.A. 1986-10-01 A revised wastage correlation was developed for FCCI with fabrication and operating parameters. The expansion of the data base to 305 data sets provided sufficient data to employ normal statistical techniques for calculation of confidence levels without unduly penalizing predictions. The correlation based on 316 SS cladding also adequately accounts for limited measured depths of interaction for fuel pins with D9 and HTq cladding
Review of the Conceptual Design for In-Vessel Fuel Handling Machines in SFR International Nuclear Information System [INIS] Kim, S. H.; Koo, G. H. 2012-01-01 The main in-vessel fuel handling machines in sodium cooled fast reactor[SFR] are composed of the in-vessel transfer machine[IVTM] and the rotating plug. These machines perform the function to handle fuel assemblies inside the reactor core during the refueling time. The IVTM should be able to access all areas above the reactor core and the fuel transfer port which can discharge the fuel assembly by the rotation of the rotating plug. In the 600 MWe demonstration reactor, the conceptual design of the in-vessel fuel handling machines was carried out. As shown in Fig. 1, the invessel fuel handling machines of the demonstration reactor are the double rotating plug type. With reference to the given core configuration of the demonstration reactor, the arrangement design of the rotating plug was carried out by using the developed simulation program. At present, the conceptual design of SFR prototype reactor which has small capacity of about 100 MWe is being started. Thus, it is necessary the economical efficiency and the reliability of the in-vessel fuel handling machines are reviewed according to the reduction of the power capacity. In this study, the preliminary design concepts of the main invessel fuel handling machines according to the fuel handling type are compared. Also, the design characteristics for the driving mechanism of the IVTM in the demonstration reactor and the recovery concept from the malfunction are reviewed
In-reactor cladding breach of EBR-II driver-fuel elements International Nuclear Information System [INIS] Seidel, B.R.; Einziger, R.E. 1977-01-01 Knowledge of performance and minimum useful element lifetime of Mark-II driver-fuel elements is required to maintain a high plant operating capacity factor with maximum fuel utilization. To obtain such knowledge, intentional cladding breach has been obtained in four run-to-cladding-breach Mark-II experimental driver-fuel subassemblies operating under normal conditions in EBR-II. Breach and subsequent fission-product release proved benign to reactor operations. The breaches originated on the outer surface of the cladding in the root of the restrainer dimples and were intergranular. The Weibull distribution of lifetime accurately predicts the observed minimum useful element lifetime of 10 at.% burnup, with breach ensuing shortly thereafter
FRACAS: a subcode for the analysis of fuel pellet-cladding mechanical interaction International Nuclear Information System [INIS] Bohn, M.P. 1977-04-01 This report describes FRACAS [Fuel Rod and Cladding Analysis Subcode], a computer code which performs the mechanical analysis in the FRAP fuel rod codes. At each loadstep, FRACAS obtains a complete elastic-plastic-creep solution for the stresses, strains, and displacements in the fuel rod cladding. The cladding is modeled as a thin cylindrical shell with prescribed temperature, pressures, and radial displacement of the inside surface. The displacement of the fuel pellets is assumed to be due to thermal gradients only. Three different regimes of pellet-cladding mechanical interaction are considered: [a] open gap, [b] closed gap, and [c] trapped stack. Both transient and steady state creep calculations are performed. The capabilities of the code are illustrated by an example problem, and comparisons are made with data obtained from two experimental fuel rods

Chemical interaction of fuel and cladding tubes International Nuclear Information System [INIS] Kirihara, Tomoo; Yamawaki, Michio; Obata, Naomi; Handa, Muneo. 1983-01-01 It was attempted to take up the behavior of nuclear fuel in cores and summarize it by the expert committee on the irradiation behavior of nuclear fuel from fiscal 1978 to fiscal 1980 from the following viewpoints. The behavior of nuclear fuel in cores has been treated separately according to each reactor type, accordingly this point is reconsidered. The clearly understood points and the uncertain points are discriminated. It is made more easily understandable for people in other fields of atomic energy. This report is that of the group on the chemical interaction, and the first report of this committee. The chemical interaction as the behavior of fuel in cores is in the unseparable relation to the mechanical interaction, but this relation is not included in this report. The chemical interaction of fuel and cladding tubes under irradiation shows different phenomena in LWRs and FBRs, and is called SCC and FCC, respectively. But this point of causing the difference must be understood to grasp the behavior of fuel. The mutual comparison of oxide fuels for FBRs and LWRs, the stress corrosion cracking of zircaloy tubes, and fuel-cladding chemical interaction in FBRs are reported. [Kako, I.]
Irradiation experience with HT9-clad metallic fuel International Nuclear Information System [INIS] Pahl, R.G.; Lahm, C.E.; Tsai, H.; Billone, M.C. 1991-01-01 The safe and reliable performance of metallic fuel is currently under study and demonstration in the Integral Fast Reactor program. In-reactor tests of HT9-clad metallic fuel have now reached maturity and have all shown good performance characteristics to burnups exceeding 17.5 at. % in the lead assembly. Because this low-swelling tempered martensitic alloy is the cladding of choice for high fluence applications, the experimental observations and performance modelling efforts reported in this paper play an important role in demonstrating reliability
Nuclear-powered pacemaker fuel cladding study International Nuclear Information System [INIS] Shoup, R.L. 1976-01-01 The composite of metals and alloys used in the fabrication of 238 Pu cardiac pacemaker fuel capsules resists the effects of high temperatures, high mechanical forces, and chemical corrosives and provides more than adequate protection to the fuel pellet even from deliberate attempts to dissolve the cladding in inorganic acids. This does not imply that opening a pacemaker fuel capsule by inorganic acids is impossible but that it would not be a wise choice
Facility for in-reactor creep testing of fuel cladding International Nuclear Information System [INIS] Kohn, E.; Wright, M.G. 1976-11-01 A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. [author]
Microbial biofilm growth on irradiated, spent nuclear fuel cladding International Nuclear Information System [INIS] Bruhn, D.F.; Frank, S.M.; Roberto, F.F.; Pinhero, P.J.; Johnson, S.G. 2009-01-01 A fundamental criticism regarding the potential for microbial influenced corrosion in spent nuclear fuel cladding or storage containers concerns whether the required microorganisms can, in fact, survive radiation fields inherent in these materials. This study was performed to unequivocally answer this critique by addressing the potential for biofilm formation, the precursor to microbial-influenced corrosion, in radiation fields representative of spent nuclear fuel storage environments. This study involved the formation of a microbial biofilm on irradiated spent nuclear fuel cladding within a hot cell environment. This was accomplished by introducing 22 species of bacteria, in nutrient-rich media, to test vessels containing irradiated cladding sections and that was then surrounded by radioactive source material. The overall dose rate exceeded 2 Gy/h gamma/beta radiation with the total dose received by some of the bacteria reaching 5 x 10 3 Gy. This study provides evidence for the formation of biofilms on spent-fuel materials, and the implication of microbial influenced corrosion in the storage and permanent deposition of spent nuclear fuel in repository environments
High performance fuel technology development : Development of high performance cladding materials International Nuclear Information System [INIS] Park, Jeongyong; Jeong, Y. H.; Park, S. Y. 2012-04-01 The superior in-pile performance of the HANA claddings have been verified by the successful irradiation test and in the Halden research reactor up to the high burn-up of 67GWD/MTU. The in-pile corrosion and creep resistances of HANA claddings were improved by 40% and 50%, respectively, over Zircaloy-4. HANA claddings have been also irradiated in the commercial reactor up to 2 reactor cycles, showing the corrosion resistance 40% better than that of ZIRLO in the same fuel assembly. Long-term out-of-pile performance tests for the candidates of the next generation cladding materials have produced the highly reliable test results. The final candidate alloys were selected and they showed the corrosion resistance 50% better than the foreign advanced claddings, which is beyond the original target. The LOCA-related properties were also improved by 20% over the foreign advanced claddings. In order to establish the optimal manufacturing process for the inner and outer claddings of the dual-cooled fuel, 18 different kinds of specimens were fabricated with various cold working and annealing conditions. Based on the performance tests and various out-of-pile test results obtained from the specimens, the optimal manufacturing process was established for the inner and outer cladding tubes of the dual-cooled fuel
Development of advanced zirconium fuel cladding International Nuclear Information System [INIS] Jeong, Young Hwan; Park, S. Y.; Lee, M. H. 2007-04-01 This report includes the manufacturing technology developed for HANA TM claddings, a series of their characterization results as well as the results of their in-pile and out-of pile performances tests which were carried out to develop some fuel claddings for a high burn-up [70,000MWd/mtU] which are competitive in the world market. Some of the HANA TM claddings, which had been manufactured based on the results from the 1st and 2nd phases of the project, have been tested in a research reactor in Halden of Norway for an in-pile performance qualification. The results of the in-pile test showed that the performance of the HANA TM claddings for corrosion and creep was better than 50% compared to that of Zircaloy-4 or A cladding. It was also found that the out-of pile performance of the HANA TM claddings for such as LOCA and RIA in some accident conditions corrosion creep, tensile, burst and fatigue was superior or equivalent to that of the Zircaloy-4 or A cladding. The project also produced the other many data which were required to get a license for an in-pile test of HANA TM claddings in a commercial reactor. The data for the qualification or characterization were provided for KNFC to assist their activities to get the license for the in-pile test of HANA TM Lead Test Rods[LTR] in a commercial reactor
Siemens advance PWR fuel assemblies [HTP] and cladding International Nuclear Information System [INIS] Stout, R. B.; Woods, K. N. 1997-01-01 This paper describes the key features of the Siemens HTP [High Thermal Performance] fuel design, the current in-reactor performance of this advanced fuel assembly design, and the advanced cladding types available
Temperature distribution determination of JPSR power reactor fuel element and cladding International Nuclear Information System [INIS] Sudarmono 1996-01-01 In order to utilize of fuel rod efficiency, a concept of JAERI passive Safety Reactor [JPSR] has been developed in Japan Atomic Energy Research Institute. In the JPSR design, UO 2 . are adopted as a fuel rod. The temperature distribution in the fuel rod and cladding in the hottest channel is a potential limiting design constraint of the JPSR. In the present determination, temperature distribution of the fuel rod and cladding for JPSR were PET:formed using COBRA-IV-I to evaluate the safety margin of the present JPSR design. In this method, the whole core was represented by the 1/4 sector and divided into 50 subchannels and 40 axial nodes. The temperature become maximum at the elevation of 1.922 and 2.196 m in the typical cell under operating condition. The maximum temperature in the center of the fuel rod surface of the fuel rod and cladding were 1620,4 o C, 722,8 o C, and 348,6 o C. The maximum results of temperature in the center of the fuel rod and cladding; were 2015,28 o C and 550 o C which were observed at 3.1 second in the typical cell
Technical committee meeting on fuel and cladding interaction. Summary report Energy Technology Data Exchange [ETDEWEB] NONE 1977-04-01 Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors [most frequently LMFBRs]. This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases.
Technical committee meeting on fuel and cladding interaction. Summary report International Nuclear Information System [INIS] 1977-04-01 Experiments and experiences concerning fuel-cladding interaction in thermal and fast neutron flux burnup are dealt with. A number of results from in-pile and out-of pile experiments with different fuel pins with cladding made of different stainless steels showed the importance of corrosion process, dependent on the burnup, core temperature, metal-oxide ratio, and other steady state parameters in the core of fast reactors [most frequently LMFBRs]. This is of importance for fuel pins design and fabrication. Mixed oxide fuel is treated in many cases
Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor International Nuclear Information System [INIS] Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom 2009-01-01 The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory [ORNL] High Flux Isotope Reactor [HFIR] and the newly expanded post-irradiation examination [PIE] capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions [i.e., for both fast-spectrum conditions and light-water-reactor conditions]. This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.
Acceptance criteria for interim dry storage of aluminum-clad fuels International Nuclear Information System [INIS] Sindelar, R.L.; Peacock, H.B. Jr.; Iyer, N.C.; Louthan, M.R. Jr. 1994-01-01 Direct repository disposal of foreign and domestic research reactor fuels owned by the United States Department of Energy is an alternative to reprocessing [together with vitrification of the high level waste and storage in an engineered barrier] for ultimate disposition. Neither the storage systems nor the requirements and specifications for acceptable forms for direct repository disposal have been developed; therefore, an interim storage strategy is needed to safely store these fuels. Dry storage [within identified limits] of the fuels received from wet-basin storage would avoid excessive degradation to assure post-storage handleability, a full range of ultimate disposal options, criticality safety, and provide for maintaining confinement by the fuel/clad system. Dry storage requirements and technologies for US commercial fuels, specifically zircaloy-clad fuels under inert cover gas, are well established. Dry storage requirements and technologies for a system with a design life of 40 years for dry storage of aluminum-clad foreign and domestic research reactor fuels are being developed by various groups within programs sponsored by the DOE
Chemical dissolution of spent fuel and cladding using complexed fluoride species International Nuclear Information System [INIS] Rance, P.J.W.; Freeman, G.A.; Mishin, V.; Issoupov, V. 2001-01-01 The dissolution of LWR fuel cladding using two fluoride ion donors, HBF 4 and K 2 ZrF 6 , in combination with nitric acid has been investigated as a potential reprocessing head-end process suitable for chemical decladding and fuel dissolution in a single process step. Maximum zirconium concentrations in the order of 0,75 to 1 molar have been achieved and dissolution found to continue to low F:Zr ratios albeit at ever decreasing rates. Dissolution rates of un-oxidised zirconium based fuel claddings are fast, whereas oxidised materials exhibit an induction period prior to dissolution. Data is presented relating to the rates of dissolution of cladding and UO 2 fuels under various conditions. [author]
Technology readiness level [TRL] assessment of cladding alloys for advanced nuclear fuels International Nuclear Information System [INIS] Shepherd, Daniel 2015-01-01 Reliable fuel claddings are essential for the safe, sustainable and economic operation of nuclear stations. This paper presents a worldwide TRL assessment of advanced claddings for Gen III and IV reactors following an extensive literature review. Claddings include austenitic, ferritic/martensitic [F/M], reduced activation [RA] and oxide dispersion strengthened [ODS] steels as well as advanced iron-based alloys [Kanthal alloys]. Also assessed are alloys of zirconium, nickel [including Hastelloy R ], titanium, chromium, vanadium and refractory metals [Nb, Mo, Ta and W]. Comparison is made with Cf/C and SiCf/SiC composites, MAX phase ceramics, cermets and TRISO fuel particle coatings. The results show in general that the higher the maximum operating temperature of the cladding, the lower the TRL. Advanced claddings were found to have lower TRLs than the corresponding fuel materials, and therefore may be the limiting factor in the deployment of advanced fuels and even possibly the entire reactor in the case of Gen IV. [authors]
Performance testing of refractory alloy-clad fuel elements for space reactors International Nuclear Information System [INIS] Dutt, D.S.; Cox, C.M.; Karnesky, R.A.; Millhollen, M.K. 1985-01-01 Two fast reactor irradiation tests, SP-1 and SP-2, provide a unique and self-consistent data set with which to evaluate the technical feasibility of potential fuel systems for the SP-100 space reactor. Fuel pins fabricated with leading cladding candidates [Nb-1Zr, PWC-11, and Mo-13Re] and fuel forms [UN and UO 2 ] are operated at temperatures typical of those expected in the SP-100 design. The first US fast reactor irradiated, refractory alloy clad fuel pins, from the SP-1 test, reached 1 at. % burnup in EBR-II in March 1985. At that time selected pins were discharged for interim examination. These examinations confirmed the excellent performance of the Nb-1Zr clad uranium oxide and uranium nitride fuel elements, which are the baseline fuel systems for two SP-100 reactor concepts
Circumferential nonuniformity of cladding radiation swelling of fast reactor peripheral fuel elements International Nuclear Information System [INIS] Reutov, V.F.; Farkhutdinov, K.G. 1977-01-01 The results are presented of the investigation into the perimeter radiation swelling of Kh18N10T stainless steel cladding in different cross sections of a peripheral fuel element of the BR-5 reactor. The fluence on the cladding is 1.8-2.9 x 10 22 fast neutr/cm 2 , the operating temperatures in different parts of the fuel element being 430 deg to 585 deg C. There has been observed circumferential non-uniformity of the distribution, concentration, and of the total volume of radiation cavities, which is due to temperature non-uniformity along the cladding perimeter. It is shown that such non-uniformity of radiation swelling of the cladding material may result in bending of the peripheral fuel element with regard to the fuel assembly sheath walls
Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins International Nuclear Information System [INIS] Roake, W.E.; Adamson, M.G.; Hilbert, R.F.; Langer, S. 1977-01-01 Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction [FCCI] has been recognized as an important factor in the ability to achieve goal peak burnups of 8% [80.MWd/kg] in FFTF and in excess of 10% [100.MWd/kg] in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to ∼60 deg. C [1100 deg. F]. In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II [EBR-II] as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor [GETR] and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin
Investigations of fuel cladding chemical interaction in irradiated LMFBR type oxide fuel pins Energy Technology Data Exchange [ETDEWEB] Roake, W E [Westinghouse-Hanford Co., Richland, WA [United States]; Adamson, M G [General Electric Company, Vallecitos Nuclear Center, Pleasanton, CA [United States]; Hilbert, R F; Langer, S 1977-04-01 Understanding and controlling the chemical attack of fuel pin cladding by fuel and fission products are major objectives of the U.S. LMFBR Mixed Oxide Irradiation Testing Program. Fuel-cladding chemical interaction [FCCI] has been recognized as an important factor in the ability to achieve goal peak burnups of 8% [80.MWd/kg] in FFTF and in excess of 10% [100.MWd/kg] in the LMFBR demonstration reactors while maintaining coolant bulk outlet temperatures up to {approx}60 deg. C [1100 deg. F]. In this paper we review pertinent parts of the irradiation program and describe recent observation of FCCI in the fuel pins of this program. One goal of the FCCI investigations is to obtain a sufficiently quantitative understanding of FCCI such that correlations can be developed relating loss of effective cladding thickness to irradiation and fuel pin fabrication parameters. Wastage correlations being developed using different approaches are discussed. Much of the early data on FCCI obtained in the U.S. Mixed Oxide Fuel Program came from capsule tests irradiated in both fast and thermal flux facilities. The fast flux irradiated encapsulated fuel pins continue to provide valuable data and insight into FCCI. Currently, however, bare pins with prototypic fuels and cladding irradiated in the fast flux Experimental Breeder Reactor-II [EBR-II] as multiple pin assemblies under prototypic powers, temperatures and thermal gradients are providing growing quantities of data on FCCI characteristics and cladding thickness losses from FCCI. A few special encapsulated fuel pin tests are being conducted in the General Electric Test Reactor [GETR] and EBR-II, but these are aimed at providing specific information under irradiation conditions not achievable in the fast flux bare pin assemblies or because EBR-II Operation or Safety requirements dictate that the pins be encapsulated. The discussion in this paper is limited to fast flux irradiation test results from encapsulated pins and multiple pin
Potential for fuel melting and cladding thermal failure during a PCM event in LWRs International Nuclear Information System [INIS] El-Genk, M.S.; Croucher, D.W. 1979-01-01 The primary concern in nuclear reactor safety is to ensure that no conceivable accident, whether initiated by a failure of the reactor system or by incorrect operation, will lead to a dangerous release of radiation to the environment. A number of hypothesized off-normal power or cooling conditions, generally termed as power-cooling-mismatch [PCM] accidents, are considered in the safety analysis of light water reactors [LWRs]. During a PCM accident, film boiling may occur at the cladding surface and cause a rapid temperature increase in the fuel and the cladding, perhaps producing embrittlement of the zircaloy cladding by oxidation. Molten fuel may be produced at the center of the pellets, extrude radially through open cracks in the outer, unmelted portion of the pellet and relocate in the fuel-cladding gap. If the amount of extruded molten fuel is sufficient to establish contact with the cladding, which is at a high temperature during film boiling, the zircaloy cladding may melt. The present work assesses the potential for central fuel melting and thermal failure of the zircaloy cladding due to melting upon being contacted by extruded molten UO 2 -fuel during a PCM event

Early implementation of SiC cladding fuel performance models in BISON Energy Technology Data Exchange [ETDEWEB] Powers, Jeffrey J. [Oak Ridge National Lab. [ORNL], Oak Ridge, TN [United States] 2015-09-18 SiC-based ceramic matrix composites [CMCs] [5–8] are being developed and evaluated internationally as potential LWR cladding options. These development activities include interests within both the DOE-NE LWR Sustainability [LWRS] Program and the DOE-NE Advanced Fuels Campaign. The LWRS Program considers SiC ceramic matrix composites [CMCs] as offering potentially revolutionary gains as a cladding material, with possible benefits including more efficient normal operating conditions and higher safety margins under accident conditions [9]. Within the Advanced Fuels Campaign, SiC-based composites are a candidate ATF cladding material that could achieve several goals, such as reducing the rates of heat and hydrogen generation due to lower cladding oxidation rates in HT steam [10]. This work focuses on the application of SiC cladding as an ATF cladding material in PWRs, but these work efforts also support the general development and assessment of SiC as an LWR cladding material in a much broader sense.
Fuel-to-cladding heat transfer coefficient into reactor fuel element International Nuclear Information System [INIS] Lassmann, K. 1979-01-01 Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. [orig.] [de
Study on the improvement of nuclear fuel cladding reliability International Nuclear Information System [INIS] Rheem, Karp Soon; Han, Jung Ho; Jeong, Yong Hwan; Lee, Deok Hyun 1987-12-01 In order to improve the nuclear fuel cladding reliability for high burn-up fuels, the corrosion resistance of laser beam surface treated and β-quenched zircaloys and the mechanical characteristics including fatigue, burst, and out-of-pile PCMI characteristics of heat treated zircaloys were investigated. In addition, the inadiation characteristics of Ko-Ri reactor fuel claddings was examined. It was found that the wasteside corrosion resistance of commercial zircaloys was improved remarkably by laser beam surface treatment. The out-of-pile transient cladding failures were investigated in terms of hoop stress versus time-to-failures by means of mandrel loading units at 25 deg C and 325 deg C. Fatigue characteristics of the β-quenched and as-received zircaloy cladding were investigated by using an internal oil pressurization method which can simulate the load-following operation cycle. The results were in good agreement with the existing data obtained by conventional methods for commercial zircaloys. Burst tests were performed with commercial and the β-quenched zircaloys in high pressure argon gas atmosphere as a function of burst temperature. The burst stress decreased linearly in the α phase region up to 600 deg C and hereafter the decrement of the burst stress decreased gradually with temperature in the β-phase region. For the first time, the burst characteristic of the irradiated zircaloy-4 cladding tubes released from Ko-Ri nuclear power unit 1 was investigated, and attempts were made to trace the cause of cladding failures by examining the failed structure and fret marks by debris. [Author]
Statistical mechanical analysis of LMFBR fuel cladding tubes International Nuclear Information System [INIS] Poncelet, J.-P.; Pay, A. 1977-01-01 The most important design requirement on fuel pin cladding for LMFBR's is its mechanical integrity. Disruptive factors include internal pressure from mixed oxide fuel fission gas release, thermal stresses and high temperature creep, neutron-induced differential void-swelling as a source of stress in the cladding and irradiation creep of stainless steel material, corrosion by fission products. Under irradiation these load-restraining mechanisms are accentuated by stainless steel embrittlement and strength alterations. To account for the numerous uncertainties involved in the analysis by theoretical models and computer codes statistical tools are unavoidably requested, i.e. Monte Carlo simulation methods. Thanks to these techniques, uncertainties in nominal characteristics, material properties and environmental conditions can be linked up in a correct way and used for a more accurate conceptual design. First, a thermal creep damage index is set up through a sufficiently sophisticated clad physical analysis including arbitrary time dependence of power and neutron flux as well as effects of sodium temperature, burnup and steel mechanical behavior. Although this strain limit approach implies a more general but time consuming model., on the counterpart the net output is improved and e.g. clad temperature, stress and strain maxima may be easily assessed. A full spectrum of variables are statistically treated to account for their probability distributions. Creep damage probability may be obtained and can contribute to a quantitative fuel probability estimation
Performance of IN-706 and PE-16 cladding in mixed-oxide fuel pins International Nuclear Information System [INIS] Makenas, B.J.; Lawrence, L.A.; Jensen, B.W. 1982-05-01 Iron-nickel base, precipitation-strengthened alloys, IN-706 and PE-16, advanced alloy cladding considered for breeder reactor applications, were irradiated in mixed-oxide fuel pins in the HEDL-P-60 subassembly in EBR-II. Initial selection of candidate advanced alloys was done using only nonfueled materials test results. However, to establish the performance characteristics of the candidate cladding alloys, i.e., dimensional stability and structural integrity under conditions of high neutron flux, elevated temperature, and applied stress, it was necessary to irradiate fuel pins under typical operating conditions. Fuel pins were clad with solution treated IN-706 and PE-16 and irradiated to peak fluences of 6.1 x 10 22 n/cm 2 [E > .1 MeV] and 8.8 x 10 22 n/cm 2 [E > .1 MeV] respectively. Fabrication and operating parameters for the fuel pins with the advanced cladding alloy candidates are summarized. Irradiation of HEDL-P-60 was interrupted with the breach of a pin with IN-706 cladding at 5.1 at % and the test was terminated with cladding breach in a pin with PE-16 cladding at 7.6 at %
Corrosion of research reactor aluminium clad spent fuel in water. Additional information International Nuclear Information System [INIS] 2009-12-01 A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools [or basins] close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project [CRP] on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water [Phase I] initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water [Phase II], to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors [2001-2006] was corrosion monitoring and surveillance of research
Cladding failure margins for metallic fuel in the integral fast reactor International Nuclear Information System [INIS] Bauer, T.H.; Fenske, G.R.; Kramer, J.M. 1987-01-01 The Integral Fast Reactor [IFR] concept being developed at Argonne National Laboratory has prompted a renewed interest in uranium-based metal alloys as a fuel for sodium-cooled fast reactors. In this paper we will present recent measurements of cladding eutectic penetration rates for the ternary IFR alloy and will compare these results with earlier eutectic penetration data for other fuel and cladding materials. A method for calculating failure of metallic fuel pins is developed by combining cladding deformation equations with a large strain analysis where the hoop stress is calculated using the instantaneous wall thickness as determined from correlations of the eutectic penetration-rate data. This method is applied to analyze the results of in-reactor and out-of-reactor fuel pin failure tests on uranium-fissium alloy EBR-II Mark-II driver fuel. In the final section of this paper we extend the calculations to consider the failure of IFR ternary fuel under reactor accident conditions. [orig./GL]
The fuel to clad heat transfer coefficient in advanced MX-type fuel pins International Nuclear Information System [INIS] Caligara, F.; Campana, M.; Mandler, R.; Blank, H. 1979-01-01 Advanced fuels [mixed carbides, nitrides and carbonitrides] are characterised by a high thermal conductivity compared to that of oxide fuels [5 times greater] and their behaviour under irradiation [amount of swelling, fracture behaviour, restructuring] is far more sensitive to the design parameters and to the operating temperature than that of oxide fuels. The use of advanced fuels is therefore conditioned by the possibility of mastering the above phenomena, and the full exploitation of their favorable neutron characteristics depends upon a good understanding of the mutual relationships of the various parameters, which eventually affect the mechanical stability of the pin. By far the most important parameter is the radial temperature profile which controls the swelling of the fuel and the build-up of stress fields within the pin. Since the rate of fission gas swelling of these fuels is relatively large, a sufficient amount of free space has to be provided within the pin. This space originally appears as fabrication porosity and as fuel-to-clad clearance. Due to the large initial gap width and to the high fuel thermal conductivity, the range of the fuel operating temperatures is mainly determined by the fuel-to-clad heat transfer coefficient h, whose correct determination becomes one of the central points in modelling. During the many years of modelling activity in the field of oxide fuels, several theoretical models have been developed to calculate h, and a large amount of experimental data has been produced for the empirical adjustment of the parameters involved, so that the situation may be regarded as rather satisfactory. The analysis lead to the following conclusions. A quantitative comparison of experimental h-values with existing models for h requires rather sophisticated instrumented irradiation capsules, which permit the measurement of mechanical data [concerning fuel and clad] together with heat rating and temperatures. More and better well
Pie technique of LWR fuel cladding fracture toughness test International Nuclear Information System [INIS] Endo, Shinya; Usami, Koji; Nakata, Masahito; Fukuda, Takuji; Numata, Masami; Kizaki, Minoru; Nishino, Yasuharu 2006-01-01 Remote-handling techniques were developed by cooperative research between the Department of Hot Laboratories in the Japan Atomic Energy Research Institute [JAERI] and the Nuclear Fuel Industries Ltd. [NFI] for evaluating the fracture toughness on irradiated LWR fuel cladding. The developed techniques, sample machining by using the electrical discharge machine [EDM], pre-cracking by fatigue tester, sample assembling to the compact tension [CT] shaped test fixture gave a satisfied result for a fracture toughness test developed by NFL. And post-irradiation examination [PIE] using the remote-handling techniques were carried out to evaluate the fracture toughness on BWR spent fuel cladding in the Waste Safety Testing Facility [WASTEF]. [author]
Effect of reactor chemistry and operating variables on fuel cladding corrosion in PWRs International Nuclear Information System [INIS] Park, Moon Ghu; Lee, Sang Hee 1997-01-01 As the nuclear industry extends the fuel cycle length, waterside corrosion of zircaloy cladding has become a limiting factor in PWR fuel design. Many plant chemistry factors such as, higher lithium/boron concentration in the primary coolant can influence the corrosion behavior of zircaloy cladding. The chemistry effect can be amplified in higher duty fuel, particularlywhen surface boiling occurs. Local boiling can result in increased crud deposition on fuel cladding which may induce axial power offset anomalies [AOA], recently reported in several PWR units. In this study, the effect of reactor chemistry and operating variables on Zircaloy cladding corrosion is investigated and simulation studies are performed to evaluate the optimal primary chemistry condition for extended cycle operation. [author]. 8 refs., 3 tabs., 16 figs
Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod Science.gov [United States] Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K. 2018-02-01 SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.
Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks International Nuclear Information System [INIS] Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland; Helmut Kuhl 2015-01-01 While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies [FA] from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods [SFR], normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs
Quivers For Special Fuel Rods-Disposal Of Special Fuel Rods In CASTOR V Casks Energy Technology Data Exchange [ETDEWEB] Bannani, Amin; Cebula, Wojciech; Buchmuller, Olga; Huggenberg, Roland [GNS, Essen [Germany]; Helmut Kuhl [WTI, Julich [Germany] 2015-05-15 While GNS casks of the CASTOR family are a suitable means to transfer fuel assemblies [FA] from the NPP to an interim dry storage site, Germanys phase-out of nuclear energy has triggered the demand for an additional solution to dispose of special fuel rods [SFR], normally remaining in the fuel pond until the final shutdown of the NPP. SFR are fuel rods that had to be removed from fuel assemblies mainly due to their special condition, e. g. damages in the cladding of the fuel rods which may have occurred during reactor operations. SFR are usually stored in the spent fuel pond after they are removed from the FA. The quiver for special fuel rods features a robust yet simple design, with a high mechanical stability, a reliable leak-tightness and large safety margins for future requirements on safety analysis. The quiver for special fuel rods can be easily adapted to a large variety of different damaged fuel rods and tailored to the specific need of the customer. The quiver for special fuel rods is adaptable e.g. in length and diameter for use in other types of transport and storage casks and is applicable in other countries as well. The overall concept presented here is a first of its kind solution for the disposal of SFRs via Castor V-casks. This provides an important precondition in achieving the status 'free from nuclear fuel' of the shut down German NPPs.
A thermodynamic model for the attack behaviour in stainless steel clad oxide fuel pins International Nuclear Information System [INIS] Goetzmann, O. 1979-01-01 So far, post irradiation examination of burnt fuel pins has not revealed a clear cut picture of the cladding attack situation. For seemingly same conditions sometimes attack occurs, sometimes not. This model tries to depict the reaction possibilities along the inner cladding wall on the basis of thermodynamic facts in the fuel pin. It shows how the thermodynamic driving force for attack changes along the fuel column, and with different initial and operational conditions. Two criteria for attack are postulated: attack as a result of the direct reaction of reactive elements with cladding components; and attack as a result of the action of a special agent [CsOH]. In defining a reaction potenial the oxygen potential, the temperature conditions [cladding temperature and fuel surface temperature], and the fission products are involved. For the determination of the oxygen potential at the cladding, three models for the redistribution of oxygen across the fuel/clad gap are offered. The effect of various parameters, like rod power, gap conductance, oxygen potential, inner wall temperature, on the thermodynamic potential for attack is analysed. [Auth.]
MODELLING OF NUCLEAR FUEL CLADDING TUBES CORROSION Miroslav Cech 2016-12-01 Full Text Available This paper describes materials made of zirconium-based alloys used for nuclear fuel cladding fabrication. It is focused on corrosion problems their theoretical description and modeling in nuclear engineering.
Out-of-pile experiments of fuel-cladding chemical interaction, [2] International Nuclear Information System [INIS] Konashi, Kenji; Yato, Tadao; Kaneko, Hiromitsu; Honda, Yutaka 1980-01-01 Cesium seems to be one of the most important fission products in the fuel-cladding chemical interaction of fuel pins for LMFBRs. However the FCCI under irradiation cannot always be explained by considering only cesium-oxygen system as the corrosive, since attack does not occur in the cesium-oxygen system unless oxygen potential is sufficiently high. Cesium-tellurium-oxygen system has been proposed to account for heavy cladding attack which was sometimes found in hypostoichiometric mixed oxide fuel pins. In this paper, the experiment on the reaction of liquid tellurium with stainless steel is reported. The type 316 stainless steel claddings for Monju type fuel pins were used as the test specimens. Tellurium was contained into the cladding tubes with end plugs. The temperature dependence of the attack by tellurium was examined in the range from 450 to 900 deg C for 30 min, and the heating time dependence was examined from 5 min to 200 hr at 725 deg C. An infrared lamp furnace was used for the experiment within 7 hr, and a resistance furnace for longer experiment. The character of corrosion was matrix attack, and the reaction products on the stainless steel surfaces consisted of chrome rich inner phase and iron and nickel rich outer phase. The results are reported. [Kako, I.]
A Preliminary Design Study of Ultra-Long-Life SFR Cores having Heterogeneous Fuel Assemblies Energy Technology Data Exchange [ETDEWEB] Jung, GeonHee; You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin [Korea, Republic of] 2016-10-15 The PWR and CANDU reactors have provided electricity for several decades in our country but they have produced lots of spent fuels and so the safe and efficient disposal of these spent fuels is one of the main issues in nuclear industry. This type ultra-long-life cores are quite efficient in terms of the amount of spent fuel generation per electricity production and they can be used as an interim storage for PWR or CANDU spent fuel over several tens of years if they use the PWR or CANDU spent fuel as the initial fuel. Typically, the previous works have considered radially homogeneous fuel assemblies in which only blanket or driver fuel rods are employed and they considered axially or radially heterogeneous core configurations with the radially homogeneous fuel assemblies. These core configurations result in the propagation of the power distribution which can lead to the significant temperature changes for each fuel assembly over the time. In this work, the radially heterogeneous fuel assemblies are employed in new ultra-long-life SFR [Sodium-cooled Fast Reactor] cores to minimize the propagation of power distribution by allowing the power propagation in the fuel assemblies. In this work, new small ultra-long life SFR cores were designed with heterogeneous fuel assemblies having both blanket and driver fuel rods to minimize the propagation of power distribution over the core by allowing power propagation from driver rods to blanket rods in fuel assemblies. In particular, high fidelity depletion calculation coupled with heterogeneous Monte Carlo neutron transport calculation was performed to assess the neutronic feasibility of the ultralong life cores. The results of the analysis showed that the candidate core has the cycle length of 77 EFPYs, a small burnup reactivity swing of 1590 pcm and acceptably small SVRs both at BOC and EOC.
Critical stability conditions of the fuel element cladding; Kriticni uslovi stabilnosti kosuljice G.E Energy Technology Data Exchange [ETDEWEB] Pavlovic, M; Savic, D [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd [Serbia and Montenegro] 1968-12-15 The role of the fuel element cladding being the first safety barrier, is to prevent contamination by the fission products. Construction of the fuel element cladding depends on the reactor type, coolant type, fuel type, technology of material fabrication, influence of the material on the neutron economy, thermal conditions, etc. That is why an optimum solution has to be found. This paper deals with mechanical properties of ceramic natural UO{sub 2} sintered fuel pellets in the zircaloy-2 cladding. This type of fuel is used in heavy water reactors.
Degradation resistant fuel cladding materials and manufacturing Energy Technology Data Exchange [ETDEWEB] Marlowe, M.O. [GE Nuclear Energy, Wilmington, NC [United States]; Montes, J. [ENUSA, Madrid [Spain] 1995-12-31 GE has been producing the degradation resistant cladding [zirconium liner and zircaloy-2 surface larger] described here with the cooperation of its primary zirconium vendors since the beginning of 1994. Approximately 24 fuel reloads, or in excess of 250,000 fuel rods, have been produced using this material by GE. GE has also produced tubing for one reload of fuel that is currently being produced by its technology affiliate ENUSA. [orig./HP]
Transitioning aluminum clad spent fuels from wet to interim dry storage International Nuclear Information System [INIS] Louthan, M.R. Jr.; Iyer, N.C.; Sindelar, R.L.; Peacock, H.B. Jr. 1994-01-01 The United States Department of Energy [DOE] currently owns several hundred metric tons of aluminum clad, spent nuclear fuel and target assemblies. The vast majority of these irradiated assemblies are currently stored in water basins that were designed and operated for short term fuel cooling prior to fuel reprocessing. Recent DOE decisions to severely limit the reprocessing option have significantly lengthened the time of storage, thus increasing the tendency for corrosion induced degradation of the fuel cladding and the underlying core material. The portent of continued corrosion, coupled with the age of existing wet storage facilities and the cost of continuing basin operations, including necessary upgrades to meet current facility standards, may force the DOE to transition these wet stored, aluminum clad spent fuels to interim dry storage. The facilities for interim dry storage have not been developed, partially because fuel storage requirements and specifications for acceptable fuel forms are lacking. In spite of the lack of both facilities and specifications, current plans are to dry store fuels for approximately 40 to 60 years or until firm decisions are developed for final fuel disposition. The transition of the aluminum clad fuels from wet to interim dry storage will require a sequence of drying and canning operations which will include selected fuel preparations such as vacuum drying and conditioning of the storage atmosphere. Laboratory experiments and review of the available literature have demonstrated that successful interim dry storage may also require the use of fuel and canister cleaning or rinsing techniques that preclude, or at least minimize, the potential for the accumulation of chloride and other potentially deleterious ions in the dry storage environment. This paper summarizes an evaluation of the impact of fuel transitioning techniques on the potential for corrosion induced degradation of fuel forms during interim dry storage

Demonstration of fuel resistant to pellet-cladding interaction. Phase 2. First semiannual report, January-June 1979 International Nuclear Information System [INIS] Rosenbaum, H.S. 1979-08-01 This program has as its ultimate objective the demonstration of an advanced fuel design that is resistant to the failure mechanism known as fuel pellet-cladding interaction [PCI]. Two fuel concepts are being developed for possible demonstration within this program: [a] Cu-barrier fuel and [b] Zr-liner fuel. These advanced fuels [known collectively as barrier fuels] have special fuel cladding designed to protect the Zircaloy cladding tube from the harmful effects of localized stress and reactive fission products during reactor service. This is the first semiannual progress report for Phase 2 of this program [January-June 1979]. Progress in the irradiation testing of barrier fuel and of unfueled barrier cladding specimens is reported
On LMFBR corrosion. Part II: Consideration of the in-reactor fuel-cladding system International Nuclear Information System [INIS] Bradbury, M.H.; Pickering, S.; Walker, C.T.; Whitlow, W.H. 1976-05-01 The scientific and technological aspects of LMFBR cladding corrosion are discussed in detail. Emphasis is placed on the influence of the irradiation environment and the effect of fuel and filler-gas impurities on the corrosion process. These studies are complemented by a concise review of out-of-pile simulation experiments that endeavour to clarify the role of the aggressive fission products cesium, tellurium and iodine. The principal models for cladding corrosion are presented and critically assessed. Areas of uncertainty are exposed and some pertinent experiments are suggested. Consideration is also given to some new observations regarding the role of stress in fuel-cladding reactions and the formation of ferrite in the corrosion zone of the cladding during irradiation. Finally, two technological solutions to the problem of cladding corrosion are proposed. These are based on the use of an oxygen buffer in the fuel and the application of a protective coating to the inner surface of the cladding
Fuel-pin cladding transient failure strain criterion International Nuclear Information System [INIS] Bard, F.E.; Duncan, D.R.; Hunter, C.W. 1983-01-01 A criterion for cladding failure based on accumulated strain was developed for mixed uranium-plutonium oxide fuel pins and used to interpret the calculated strain results from failed transient fuel pin experiments conducted in the Transient Reactor Test [TREAT] facility. The new STRAIN criterion replaced a stress-based criterion that depends on the DORN parameter and that incorrectly predicted fuel pin failure for transient tested fuel pins. This paper describes the STRAIN criterion and compares its prediction with those of the stress-based criterion
Fuel cladding tube leak detection device International Nuclear Information System [INIS] Naito, Makoto. 1992-01-01 The device of the present invention can detect even a minute leakage or a continuous leakage during reactor operation. That is, the device of the present invention comprises a detector for analyzing nuclides of gases incorporated in a gas waste processing system, and a calculation device connected to the detector and detecting leakage from a fuel cladding tube by calculation for variation coefficient of long-life nuclides. By using theses devices, radioactivity contained in gases incorporated in the gas waste processing system is analyzed for the nuclides. Among the analized nuclides, if the amount of the long-life nuclides exceeds a predetermined value, it is judged as leakage of the fuel cladding tube. For example, the long-life nuclides include Xe-133. The device of the present invention can certainly detect occurrence of leakage even when it is minute or continues leakage. Accordingly, countermeasures can be taken in an early stage, thereby enabling to contribute improvement for the safety of a nuclear power plant. [I.S.]
DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS International Nuclear Information System [INIS] Rudisill, T; John Mickalonis, J 2006-01-01 The reprocessing of commercial spent nuclear fuel [SNF] generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide [ZrO 2 ] which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO 2 layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste [LLW]. Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium [Zr] metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride [NH 4 F]/ammonium nitrate [NH 4 NO 3 ] solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid [HF] process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO 2 layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm [thick] oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate [[NH 4 ] 2 ZrF 6 ] on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of
Investigation and recovery of unrecovered fuel pellets and cladding tube pieces International Nuclear Information System [INIS] Kobayashi, Keiji 1980-01-01 The total weight of the fuel pellets lost due to break was about 1206 g, and cladding tube pieces were about 217 g. Among these, the pellets of about 527 g and the cladding tube pieces of about 152 g were recovered when broken fuel rods were discovered. It is not desirable to leave these broken pieces as unrecovered in view of safety and the management of nuclear fuel materials. Kansai Electric Power Co., Inc., investigated the position and the amount of these pellets and cladding tube pieces for about a year, and recovered a part of them. The results were written in two reports. The objects of the investigation and recovery, and the method of recovery are explained. The UO 2 and zirconium recovered were 58.52 g and 369.58 g, respectively. The solid pellets were recovered from the reactor, fuel assemblies, a spent fuel pit and canals, and the content in sludge was recovered from other installations. The amounts of unrecovered pellets and cladding tube pieces in primary cooling water, coolant filters, sealing water filters, primary cooling pipes, waste resins and fuel assemblies were estimated. The problems concerning the recovery and estimation are pointed out. The results of estimating the amount of uranium in coolant filters and sealing water filters are useful to know the time of the occurrence of accident. [Kako, I.]
Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel International Nuclear Information System [INIS] Keiser, D.D. Jr.; Mariani, R.D. 1999-01-01 Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr alloy fuel elements irradiated in the experimental breeder reactor II [EBR-II]. We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments [hulls] that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr [and U] found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining. [orig.]
Design characteristics of pantograph type in vessel fuel handling system in SFR International Nuclear Information System [INIS] Kim, S. H.; Koo, G. H. 2012-01-01 The pantograph type in vessel fuel handling system in a sodium cooled fast reactor [SFR], which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine [IVTM], a single rotating plug, in vessel storage, and a fuel transfer port [FTP]. The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied
Design characteristics of pantograph type in vessel fuel handling system in SFR Energy Technology Data Exchange [ETDEWEB] Kim, S. H.; Koo, G. H. [KAERI, Daejeon [Korea, Republic of] 2012-10-15 The pantograph type in vessel fuel handling system in a sodium cooled fast reactor [SFR], which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine [IVTM], a single rotating plug, in vessel storage, and a fuel transfer port [FTP]. The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.
Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability International Nuclear Information System [INIS] Scheglov, A. 1994-01-01 In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders [CCC] model as: axial asymmetry of fuel-cladding system [due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release], gaps between the pallets [and heat release peaking in fuel near the gap], chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets [characterized by a large central hole], temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs
Influence of fuel-cladding system deviations from the model of continuous cylinders on the parameters of WWER fuel element working ability Energy Technology Data Exchange [ETDEWEB] Scheglov, A [Russian Research Centre Kurchatov Inst., Moscow [Russian Federation] 1994-12-31 In the programs of fuel rod computation, fuel and cladding are usually presented in the form of coaxial cylinders, which can change their sizes, mechanical and thermal-physical properties. The real fuel element has some typical deviations from this continuous coaxial cylinders [CCC] model as: axial asymmetry of fuel-cladding system [due to the oval form of the cladding, cracking and other type of fuel pallet damage, axial asymmetry of the volumetric heat release], gaps between the pallets [and heat release peaking in fuel near the gap], chambers in the pallets. As a result of these deviations actual fuel rod parameters of working ability - temperature, stresses, thermal fluxes relieved from the cladding, geometry changes - in some locations can greatly vary from the ones calculated according to CCC model. The influence of these deviations is extremely important while calculating the fuel rod, because they are a part of the mechanical excess coefficient. The author reviews the influence of these factors using specific examples. He applies his own two-dimensional codes based on the Finite Elements Method for calculations of temperature fields, stresses and deformation in the fuel rod elements. It is shown that consideration of these deviations, as a rule, leads to the increase of the maximum fuel temperature in the WWER pellets [characterized by a large central hole], temperature of the cladding, thermal flux, relieved by the coolant from the cladding, and stresses in the cladding. It is necessary to consider these factors for both validation of the fuel element working ability and interpretation of the experimental results. 4 tabs., 3 figs., 5 refs.
Failure probabilities of SiC clad fuel during a LOCA in public acceptable simple SMR [PASS] Energy Technology Data Exchange [ETDEWEB] Lee, Youho, E-mail: euo@kaist.ac.kr; Kim, Ho Sik, E-mail: hskim25@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr 2015-10-15 Highlights: • Graceful operating conditions of SMRs markedly lower SiC cladding stress. • Steady-state fracture probabilities of SiC cladding is below 10{sup −7} in SMRs. • PASS demonstrates fuel coolability [T < 1300 °C] with sole radiation in LOCA. • SiC cladding failure probabilities of PASS are ∼10{sup −2} in LOCA. • Cold gas gap pressure controls SiC cladding tensile stress level in LOCA. - Abstract: Structural integrity of SiC clad fuels in reference Small Modular Reactors [SMRs] [NuScale, SMART, IRIS] and a commercial pressurized water reactor [PWR] are assessed with a multi-layered SiC cladding structural analysis code. Featured with low fuel pin power and temperature, SMRs demonstrate markedly reduced incore-residence fracture probabilities below ∼10{sup −7}, compared to those of commercial PWRs ∼10{sup −6}–10{sup −1}. This demonstrates that SMRs can serve as a near-term deployment fit to SiC cladding with a sound management of its statistical brittle fracture. We proposed a novel SMR named Public Acceptable Simple SMR [PASS], which is featured with 14 × 14 assemblies of SiC clad fuels arranged in a square ring layout. PASS aims to rely on radiative cooling of fuel rods during a loss of coolant accident [LOCA] by fully leveraging high temperature tolerance of SiC cladding. An overarching assessment of SiC clad fuel performance in PASS was conducted with a combined methodology—[1] FRAPCON-SiC for steady-state performance analysis of PASS fuel rods, [2] computational fluid dynamics code FLUENT for radiative cooling rate of fuel rods during a LOCA, and [3] multi-layered SiC cladding structural analysis code with previously developed SiC recession correlations under steam environments for both steady-state and LOCA. The results show that PASS simultaneously maintains desirable fuel cooling rate with the sole radiation and sound structural integrity of fuel rods for over 36 days of a LOCA without water supply. The stress level of
Pellet-clad interaction observations in boiling water reactor fuel elements International Nuclear Information System [INIS] Sahoo, K.C.; Bahl, J.K.; Sivaramakrishnan, K.S.; Roy, P.R. 1981-01-01 Under a programme to assess the performance of fuel elements of Tarapur Atomic Power Station, post-irradiation examination has been carried out on 18 fuel elements in the first phase. Pellet-clad mechanical interaction behaviour in 14 elements with varying burnup and irradiation history has been studied using eddy current testing technique. The data has been analysed to evaluate the role of pellet-clad mechanical interaction in PCI/SCC failure in power reactor operating conditions. [author]
Investigation on fuel-cladding chemical interaction in metal fuel for FBR International Nuclear Information System [INIS] Inagaki, Kenta; Nakamura, Kinya; Ogata, Takanari; Uwaba, Tomoyuki 2013-01-01 During steady-state irradiation of metallic fuel in fast reactors, rare-earth fission products can react with stainless steel cladding at the fuel-cladding interface. The authors conducted isothermal annealing tests with some diffusion couples to investigate the structure of the wastage layer formed at the interface. Candidate cladding alloys, ferritic-martensitic steel [PNC-FMS] and oxide-dispersion-strengthened [ODS] steel were assembled with rare-earth alloys, RE5 : La-Ce-Pr-Nd-Sm, which simulate the fission yield of rare-earth fission products. The diffusion couples were isothermally annealed in the temperature range of 500-650°C for up to 170 h. In both RE5/ODS-steel and RE5/PNC-FMS couples, the wastage layer of the two-phase region of the [Fe, Cr] 17 RE 2 matrix phase with the precipitation of the [Fe, RE, Cr] phase was formed. The structure was similar to that formed in RE5/Fe-12Cr and RE5/HT9 couples, which implies that the reaction between REs and steel is not significantly influenced by the minor alloying elements within the candidate cladding materials. It was also clarified that the increase in the wastage layer thickness was diffusion-controlled. The temperature dependence of the reaction rate constants were formulated, which can be the basis for the quantification of the wastage layer growth. [author]
Compatibility study between U-UO{sub 2} cermet fuel and T91 cladding Energy Technology Data Exchange [ETDEWEB] Mishra, Sudhir, E-mail: sudhir@barc.gov.in [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 [India]; Kaity, Santu; Khan, K.B. [Radiometallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 [India]; Sengupta, Pranesh; Dey, G.K. [Materials Science Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 [India] 2016-12-01 Cermet is a new fuel concept for the fast reactor system and is ideally designed to combine beneficial properties of both ceramic and metal. In order to understand fuel clad chemical compatibility, diffusion couples were prepared with U-UO{sub 2} cermet fuel and T91 cladding material. These diffusion couples were annealed at 923–1073 K for 1000 h and 1223 K for 50 h, subsequently their microstructures were examined using scanning electron microscope [SEM], X-ray energy dispersive spectroscope [EDS] and electron probe microanalyser [EPMA]. It was observed that the interaction between the fuel and constituents of T91 clad was limited to a very small region up to the temperature 993 K and discrete U{sub 6}[Fe,Cr] and U[Fe,Cr]{sub 2} intermetallic phases developed. Eutectic microstructure was observed in the reaction zone at 1223 K. The activation energy for reaction at the fuel clad interface was determined.
Mechanical behavior of fast reactor fuel pin cladding subjected to simulated overpower transients International Nuclear Information System [INIS] Johnson, G.D.; Hunter, C.W. 1978-06-01 Cladding mechanical property data for analysis and prediction of fuel pin transient behavior were obtained under experimental conditions in which the temperature ramps of reactor transients were simulated. All cladding specimens were 20% CW Type 316 stainless steel and were cut from EBR-II irradiated fuel pins. It was determined that irradiation degraded the cladding ductility and failure strength. Specimens that had been adjacent to the fuel exhibited the poorest properties. Correlations were developed to describe the effect of neutron fluence on the mechanical behavior of the cladding. Metallographic examinations were conducted to characterize the failure mode and to establish the nature of internal and external surface corrosion. Various mechanisms for the fuel adjacency effect were examined and results for helium concentration profiles were presented. Results from the simulated transient tests were compared with TREAT test results
Protection of spent aluminum-clad research reactor fuels during extended wet storage International Nuclear Information System [INIS] Fernandes, Stela M.C.; Correa, Olandir V.; Souza, Jose A.; Ramanathan, Lalgudi V.; Antunes, Renato A. 2013-01-01 Aluminum-clad spent nuclear fuel from research reactors [RR] is stored in light water filled pools or basins worldwide. Many incidences of pitting corrosion of the fuel cladding has been reported and attributed to synergism in the effect of certain water parameters. Protection of spent Al-clad RR fuel with a conversion coating was proposed in 2008. Preliminary results revealed increased pitting corrosion resistance of cerium oxide coated aluminum alloys AA 1050 and AA 6061, used as RR fuel plate cladding. Further development of conversion coatings for Al alloys was carried out and this paper presents: [a] the preparation and characterization of hydrotalcite [HTC] coatings; [b] the results of laboratory tests in which the corrosion behavior of coated Al alloys in NaCl solutions was determined; [c] the results of field tests in which un-coated, boehmite coated, HTC coated and cerium modified boehmite / HTC coated AA 1050 and AA 6061 coupons were exposed to the IEA-R1 reactor spent fuel basin for extended periods. In these field tests the coupons coated with HTC from a high temperature [HT] bath and subsequently modified with Ce were the most resistant to pitting corrosion. In laboratory tests also, HT- hydrotalcite + Ce coated specimens were the most corrosion resistant in 0.01 M NaCl. The role of cerium in increasing the corrosion resistance imparted by the different conversion coatings of spent Al-clad RR fuel elements is presented. [author]
Corrosion issues in the long term storage of aluminum-clad spent nuclear fuels International Nuclear Information System [INIS] Louthan, M.R. Jr.; Peacock, H.B. Jr.; Sindelar, R.L.; Iyer, N.C. 1996-01-01 Approximately 8% of the spent nuclear fuel owned by the US Department of Energy is clad with aluminum alloys. The spent fuel must be either reprocessed or temporarily stored in wet or dry storage systems until a decision is made on final disposition in a repository. There are corrosion issues associated with the aluminum cladding regardless of the disposition pathway selected. This paper discusses those issues and provides data and analysis to demonstrate that control of corrosion induced degradation in aluminum clad spent fuels can be achieved through relatively simple engineering practices
Deformation and collapse of zircaloy fuel rod cladding into plenum axial gaps International Nuclear Information System [INIS] Pfennigwerth, P.L.; Gorscak, D.A.; Selsley, I.A. 1983-01-01 To minimize support structure, blanket and reflector fuel rods of the thoria urania-fueled Light Water Breeder Reactor [LWBR] were designed with non-freestanding Zircaloy-4 cladding. An analytical model was developed to predict deformation of unirradiated cladding into axial gaps of fuel rod plenum regions where it is unsupported. This model uses the ACCEPT finite element computer program to calculate elastic-plastic deformation of cladding due to external pressure. The finite element is 20-node, triquadratic, isoparametric, and 3-dimensional. Its curved surface permits accurate modeling of the tube geometry, including geometric nonuniformities such as circumferential wall thickness variation and initial tube out-of-roundness. Progressive increases in axial gap length due to cladding elongation and fuel stack shrinkage are modeled, as are deformations of fuel pellets and stainless steel support sleeves which bound plenum axial gaps in LWBR type blanket fuel rods. Zircaloy-4 primary and secondary thermal creep representations were developed from uniaxial creep testing of fuel rod tubing. Creep response to multi-axial loading is modeled with a variation of Hill's formulation for anisotropic materials. Coefficients accounting for anisotropic thermal creep in Zircaloy-4 tubes were developed from creep testing of externally pressurized tubes having fixed axial gaps in the range 2.5 cm to 5.7 cm and radial clearances over simulated fuel pellets ranging from zero to 0.089 mm. [orig./RW]
Interdiffusion between U-Pu-Zr fuel and HT9 cladding International Nuclear Information System [INIS] Keiser, D.D. Jr.; Petri, M.C. 1994-01-01 As part of systematic interdiffusion studies of fuel-cladding compatibility in the integral Fast Reactor, a solid-solid diffusion couple was assembled with U-22Pu-23 1 Zr fuel and HT9 2 cladding and annealed at 650 degrees C for 100 hours. The couple was examined for diffusion structure development using a scanning electron microscope equipped with an energy dispersive x-ray analyzer [SEM-EDX]. Point-by-point and linescan analysis was used to generate composition profiles and diffusion paths. From the composition profiles, average effective interdiffusion coefficients were calculated for individual components on both sides of the Matano plane. Results from this investigation indicate that the same types of phases as would be expected from binary U-Fe, Pu-Fe, and Zr-Fe phase diagrams develop in this couple; and U and Pu are the fastest diffusing fuel components and Fe is the fastest diffusing cladding component. Compared with diffusion couples with binary [U-Zr] fuel, the addition of Pu greatly enhanced the extent of diffusion and affected the types of phases observed

Method for automatic filling of nuclear fuel rod cladding tubes International Nuclear Information System [INIS] Bezold, H. 1979-01-01 Prior to welding the zirconium alloy cladding tubes with end caps, they are automatically filled with nuclear fuel tablets and ceramic insulating tablets. The tablets are introduced into magazine drums and led through a drying oven to a discharging station. The empty cladding tubes are removed from this discharging station and filled with tablets. A filling stamp pushes out the columns of tablets in the magazine tubes of the magazine drum into the cladding tube. Weight and measurement of length determine the filled state of the cladding tube. The cladding tubes are then led to the welding station via a conveyor belt. [DG] [de
An internal conical mandrel technique for fracture toughness measurements on nuclear fuel cladding Energy Technology Data Exchange [ETDEWEB] Sainte Catherine, C.; Le Boulch, D.; Carassou, S. [CEA Saclay, DEN/DMN, Bldg 625 P, Gif-Sur-Yvette, F-91191 [France]; Lemaignan, C. [CEA Grenoble, 17 rue des Martyrs, Grenoble, F-38054 [France]; Ramasubramanian, N. [ECCATEC Inc., 92 Deburn Drive, Toronto, Ontario [Canada] 2006-07-01 An understanding of the limiting stress level for crack initiation and propagation in a fuel cladding material is a fundamental requirement for the development of water reactor clad materials. Conventional tests, in use to evaluate fracture properties, are of limited help, because they are adapted from ASTM standards designed for thick materials, which differ significantly from fuel cladding geometry [small diameter thin-walled tubing]. The Internal Conical Mandrel [ICM] test described here is designed to simulate the effect of fuel pellet diametrical increase on a cladding with an existing axial through-wall crack. It consists in forcing a cone, having a tapered increase in diameter, inside the Zircaloy cladding with an initial axial crack. The aim of this work is to quantify the crack initiation and propagation criteria for fuel cladding material. The crack propagation is monitored by a video system for obtaining crack extension {delta}a. A finite-element [FE] simulation of the ICM test is performed in order to derive J integrals. A node release technique is applied during the FE simulation for crack propagation and the J-resistance curves [J-{delta}a] are generated. This paper presents the test methodology, the J computation validation, and results for cold-worked stress relieved Zircaloy-4 cladding at 20 deg. and 300 deg. C and also for Al 7050-T7651 aluminum alloy tubing at 20 deg. C. [authors]
Application of Coating Technology for Accident Tolerant Fuel Cladding Energy Technology Data Exchange [ETDEWEB] Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2014-10-15 To commercialize the ATF cladding concepts, various factors are considered, such as safety under normal and accident conditions, economy for the fuel cycle, and developing development challenges, and schedule. From the proposed concepts, it is known that the cladding coating, FeCrAl alloy, and Zr-Mo claddings are considered as a near/mid-term application, whereas the SiC material is considered as a long-term application. Among them, the benefit of cladding coating on Zr-based alloys is the fuel cycle economy regarding the manufacturing, neutron cross section, and high tritium permeation characteristics. However, the challenge of cladding coating on Zr-based alloys is the lower oxidation resistance and mechanical strength at high-temperature than other concepts. Another important point is the adhesion property between the Zr-based alloy and coating materials. As an improved coating technology compared to a previous study, a 3D laser coating technology supplied with Cr powders is considered to make a coated cladding because it is possible to make a coated layer on the tubular cladding surface by controlling the 3-diminational axis. We are systematically studying the laser beam power, inert gas flow, cooling of the cladding tube, and powder control as key points to develop 3D laser coating technology. After Cr-coating on the Zr-based cladding, ring compression and ring tensile tests were performed to evaluate the adhesion property between a coated layer and Zr-based alloy tube at room temperature [RT], and a high-temperature oxidation test was conducted to evaluate the oxidation behavior at 1200 .deg. C of the coated tube samples. A 3D laser coating method supplied with Cr powders was developed to decrease the high-temperature oxidation rate in a steam environment through a systematic study for various coating parameters, and a Cr-coated Zircaloy-4 cladding tube of 100 mm in length to the axial direction can be successfully manufactured.
Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly Energy Technology Data Exchange [ETDEWEB] Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik [Korea Atomic Energy Research Institute, Daejeon [Korea, Republic of] 2016-10-15 These affect the mechanical design of the fuel assembly components. And thus, appropriate structural design criteria should also be chosen to incorporate the specific design conditions of the SFR fuel assemblies. Among them, the temperature is one of the most crucial conditions to be concerned because the sodium coolant temperature is normally more than 500ºC which is much higher than that of the LWR [< 350ºC]. This implies that a thermal creep should be significantly considered in the SFR fuel assembly mechanical design. In addition to the high temperature condition, an irradiation swelling is also an important behavior that the SFR fuel assembly material should accommodate. To incorporate the temperature and irradiation impacts, the material of the fuel assembly components is presently determined to be made of HT-9, the ferriticmartensitic steel. In this paper, the ASME Sec. III Div. 5 [referred to as ‘Div. 5’ hereinafter], which was developed for a ‘high temperature reactor’, is considered as one of the structural design criteria for the mechanical design of SFR fuel assemblies. In this paper, the stress intensity limits, S{sub m} and S{sub t} of HT-9 were built for the structural criteria of an SFR fuel assembly. S{sub m} is obtained from the ultimate strength. As for S{sub t}, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of S{sub mt}, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as S{sub mt} under the temperature about 470ºC which is relatively low temperature range and over 470ºC with relatively short time duration as 1000 hours. And the S{sub t} is adopted as Smt at over 470ºC and long time duration over 34800 hours, and over 520ºC and 10{sup 4} hours too. And at over 570ºC and 1000 hours, and at over 630ºC and 100 hours, S{sub t} is also adopted for S{sub mt}.
Nuclear fuel rod with burnable plate and pellet-clad interaction fix International Nuclear Information System [INIS] Boyle, R.F. 1987-01-01 This patent describes a nuclear fuel rod comprising a metallic tubular cladding containing nuclear fuel pellets, the pellets containing enriched uranium-235. The improvement described here comprises: ceramic wafers, each wafter comprising a sintered mixture of gadolinium oxide and uranium dioxide, the uranium oxide having no more uranium-235 than is present in natural uranium dioxide. Each of the wafers is axially disposed between a major portion of adjacent the nuclear fuel pellets, whereby the wafers freeze out volatile fission products produced by the nuclear fuel and prevent interaction of the fission products with the metallic tubing cladding
FUMAC-a new model for light water reactor fuel relocation and pellet-cladding interaction International Nuclear Information System [INIS] Walton, L.A.; Matheson, J.E. 1984-01-01 An improved approach to the mechanical modeling of fuel rod performance is presented. Previous computer modeling has centered around a unified finite element approach with both fuel pellets and cladding being represented by ring elements. The fuel mechanical analysis code [FUMAC] departs from these approaches in two areas. The pellet model is an empirically based deterministic algorithm, while the cladding model uses both plane stress and plane strain finite elements. The work describes a semiempirical fuel cracking and fragment relocation model, which is burnup and power-level dependent. The interaction of the pellet with the cladding is treated classically. The resulting thick cylinder stresses are used in conjunction with an orthotropic creep model to predict cladding ridging. The resulting ridging compares well with experimental data for both steady-state and transient operating conditions. Future work planned includes the integration of the finite element cladding model with the pellet model and refinement of the pellet relocation and thermal models. Transient performance predictions will be emphasized
Method for the protection of the cladding tubes of fuel rods International Nuclear Information System [INIS] Steinberg, E. 1978-01-01 To present stress crack corrosion and to protect the cladding tubes of the fuel rods made of a circonium alloy from attack by iodine, the inward surfaces are provided with protective coatings. Therefore the casting tubes already filled with fuel element pellets are put under over-pressure at a temperature range between 300 and 500 0 C, until almost yield-point is reached. A small amount of H 2 O or H 2 O 2 , filled in, reacts with the cladding tube material to form the Zr-O 2 protective coating. Afterwards comes a pressure relief, and the cladding tube reaches its original dimensions. [DG] [de
Future possibilities of SUSEN technologies for R&D of nuclear fuel cladding International Nuclear Information System [INIS] Mikloš, M. 2015-01-01 R&D possibilities with nuclear fuel cladding were discussed in this paper. The availability of 10 MWT reactor with BWR and PWR loops having chemistry control was described. Activity transport and fuel cladding corrosion can be investigated in this facility including PIE. The facility has hot cells and the laboratory is expected to start in 2017
A state of the Art report on Manufacturing technology of high burn-up fuel cladding Energy Technology Data Exchange [ETDEWEB] Kim, Kyeong Ho; Nam, Cheol; Baek, Jong Hyuk; Choi, Byung Kwon; Park, Sang Yoon; Lee, Myung Ho; Jeong, Yong Hwan 1999-09-01 In order to manufacturing the prototype fuel cladding, overall manufacturing processes and technologies should be thoroughly understood on the manufacturing processes and technologies of foreign cladding tubes. Generally, the important technology related to fuel cladding tube manufacturing processes for PWRs/PHWRs is divided into three stages. The first stage is to produce the zirconium sponge from zirconium sand, the second stage is to produce the zircaloy shell or TREX from zirconium sponge ingot and finally, cladding is produced from TREX or zircaloy shell. Therefore, the manufacturing processes including the first and second stages are described in brief in this technology report in order to understand the whole fuel cladding manufacturing processes. [author]
Progress in Understanding of Fuel-Cladding Chemical interaction in Metal Fuel International Nuclear Information System [INIS] Inagaki, Okenta; Nakamura, Kinya; Ogata, Takanari 2013-01-01 Conclusion: Representative phases formed in FCCI were identified: • The reaction between lanthanide elements and cladding; • The reaction between U-PU-Zr and cladding [Fe]. Characteristics of the wastage layer were clarified: • Time and temperature dependency of the growth ratio of the wastage layer formed by lanthanide elements; • Threshold temperature of the liquid phase formation in the reaction between U-Pu-Zr and Fe. These results are used: - as a basis for the FCCI modeling; - as a reference data in post-irradiation examination of irradiated metallic fuels
Interim report on the creepdown of Zircaloy fuel cladding International Nuclear Information System [INIS] Hobson, D.O.; Dodd, C.V. 1977-01-01 This report describes the creepdown phenomenon in Zircaloy fuel cladding and the methods by which it will be measured and analyzed. Instrumentation for monitoring radial deformation in the cladding is described in detail--in terms of theory, design, and stability. The programs that control the microcomputer are listed, both to document the level of sophistication of the instrumentation and to indicate the flexibility of the test equipment
Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10 International Nuclear Information System [INIS] Smith, E.; Ranjan, G.V.; Cipolla, R.C. 1976-11-01 Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs [i.e., ''uniform'' conditions]. Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking
PCI resistant light water reactor fuel cladding International Nuclear Information System [INIS] Foster, J.P.; Sabol, G.P. 1988-01-01 A tubular nuclear fuel element cladding tube is described, the fuel element cladding tube forming the entire fuel element cladding and consisting of: a single continuous wall, the single continuous wall consisting of a single alloy selected from the group consisting of zirconium base alloys, A, B, C, D, and E; the single continuous wall characterized by a cold worked and stress relieved microstructure throughout; wherein the zirconium base alloy A contains 0.2 - 0.6 w/o Sn, 0.03 - 0.11 w/o sum of Fe and Cr, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy B contains 0.1 - 0.6 w/oo Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O and section 1500 ppm total impurities; the zirconium base alloy C contains 1.2 - 1.7 w/o Sn, 0.04 - 0.24 w/o Fe, 0.05 - 0.15 w/o Cr, section 0.08 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities; the zirconium base alloy D contains 0.15 - 0.6 w/o Sn, 0.15 - 0.5 w/o Fe, section 600 ppm O, and section 1500 ppm total impurities; and the zirconium base alloy E contains 0.4 - 0.6 w/o Sn, 0.1 - 0.3 w/o Fe, 0.03 - 0.07 w/o Ni, section 600 ppm O, and section 1500 ppm total impurities
Improving Accident Tolerance of Nuclear Fuel with Coated Mo-alloy Cladding Bo Cheng 2016-02-01 Full Text Available In severe loss of coolant accidents [LOCA, similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconium alloy fuel cladding materials are rapidly heated due to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel [ATF design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute [EPRI is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in 1,200–1,500°C steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstrated corrosion resistance. As these composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Mo alloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are
About criteria of inadmissible embrittlement of zirconium fuel cladding during LOCA in the PWRs International Nuclear Information System [INIS] Osmachkin, V.S. 1999-01-01 According the licensing procedures the designers of the PWRs reactor have to prove the meeting of special safety requirements. One criteria on effectiveness of the Emergency Core Cooling System is not to exceeding some limited conditions of the fuel cladding during LOCA accidents [typical example T m ax o C, ECR

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