Very-high-temperature gas-cooled reactors (VHTR) are being developed to provide higher thermal efficiency and high-temperature process heat. Zirconium carbide (ZrC) has been proposed as a potential coating material for TRistructural-ISOtropic (TRISO) coated fuel particles because of its excellent resistance to fission products corrosion, good thermal stability and higher mechanical strength under elevated temperatures. The integrity and performance of the ZrC coating of the TRISO particles are very important as it provides the main barrier for fission product release. Therefore, the microstructure and property evolution of ZrC coating deserve to be investigated. Fluidized-bed chemical vapor deposition (FB-CVD) has been conducted to fabricate the ZrC coating in a ZrCl4−C3H6-Ar-H2 system. The stoichiometry of ZrC was changed by controlling the feeding rate of ZrCl4 and the flow rate of C3H6. The ZrC coatings were annealed from 1700 °C to 2200 °C to study the possible changes in microstructures and temperature-dependent performances. The effect of stoichiometries on ZrC coating was studied by X-ray diffraction (XRD), scanning electron microscopy (SEM), Raman spectroscopy (Raman), and nanoindenter. Results showed that free carbon prevents grain growth under high-temperature annealing, and it reacts with ZrC1-x at higher temperatures to form pure phase ZrC. In addition, the microstructure evolution mechanism of ZrC at high temperatures was proposed.
{"title":"Effects of Stoichiometry and High-Temperature Annealing on Zirconium Carbide Coating Layer in TRISO Particles","authors":"Xinyu Cheng, Rongzheng Liu, Bing Liu, Xueru Yang, Malin Liu, J. Chang, You-lin Shao","doi":"10.1115/icone29-92858","DOIUrl":"https://doi.org/10.1115/icone29-92858","url":null,"abstract":"\u0000 Very-high-temperature gas-cooled reactors (VHTR) are being developed to provide higher thermal efficiency and high-temperature process heat. Zirconium carbide (ZrC) has been proposed as a potential coating material for TRistructural-ISOtropic (TRISO) coated fuel particles because of its excellent resistance to fission products corrosion, good thermal stability and higher mechanical strength under elevated temperatures. The integrity and performance of the ZrC coating of the TRISO particles are very important as it provides the main barrier for fission product release. Therefore, the microstructure and property evolution of ZrC coating deserve to be investigated. Fluidized-bed chemical vapor deposition (FB-CVD) has been conducted to fabricate the ZrC coating in a ZrCl4−C3H6-Ar-H2 system. The stoichiometry of ZrC was changed by controlling the feeding rate of ZrCl4 and the flow rate of C3H6. The ZrC coatings were annealed from 1700 °C to 2200 °C to study the possible changes in microstructures and temperature-dependent performances. The effect of stoichiometries on ZrC coating was studied by X-ray diffraction (XRD), scanning electron microscopy (SEM), Raman spectroscopy (Raman), and nanoindenter. Results showed that free carbon prevents grain growth under high-temperature annealing, and it reacts with ZrC1-x at higher temperatures to form pure phase ZrC. In addition, the microstructure evolution mechanism of ZrC at high temperatures was proposed.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"35 2 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"83006706","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
H. Yao, Tianzhou Ye, Junmei Wu, Yingwei Wu, C. Yin, Ping Chen
Nuclear fuel cladding is subjected to neutron irradiation in a high-temperature stress environment, and the structural integrity of the cladding is very important for the safe operation of nuclear reactors. FeCrAl alloy has become a promising candidate cladding material for the accident tolerance fuel development in view of its excellent irradiation resistance and high temperature strength. This work aims to study the creep properties of FeCrAl alloy at high temperatures under neutron irradiation. Thermal and irradiation creep behavior in nanocrystalline FeCrAl samples is examined using molecular dynamics simulation method. And the effects of temperature, stress, irradiation dose rate on the creep rate and parameters of the creep constitutive equations are discussed. The results show that the thermal creep rate is greater than irradiation creep rate. The effect of temperature on the thermal creep stress exponent is relatively small at low stress, but is obvious when stress exceeds 0.8 GPa. The higher the temperature, the larger the thermal creep stress exponent. The irradiation creep rate increases almost linearly with the dose rate, that is, the exponent of dose rate for irradiation creep approach 1.0. Irradiation creep stress exponent fluctuates very little around 1.1 within the scope of the present research. Besides, higher temperature accelerates the linear increase of irradiation creep rate with dose rate, and the irradiation creep pre-factor becomes higher.
{"title":"Creep Properties of FeCrAl Alloy at High Temperature Under Neutron Irradiation","authors":"H. Yao, Tianzhou Ye, Junmei Wu, Yingwei Wu, C. Yin, Ping Chen","doi":"10.1115/icone29-89079","DOIUrl":"https://doi.org/10.1115/icone29-89079","url":null,"abstract":"\u0000 Nuclear fuel cladding is subjected to neutron irradiation in a high-temperature stress environment, and the structural integrity of the cladding is very important for the safe operation of nuclear reactors. FeCrAl alloy has become a promising candidate cladding material for the accident tolerance fuel development in view of its excellent irradiation resistance and high temperature strength. This work aims to study the creep properties of FeCrAl alloy at high temperatures under neutron irradiation. Thermal and irradiation creep behavior in nanocrystalline FeCrAl samples is examined using molecular dynamics simulation method. And the effects of temperature, stress, irradiation dose rate on the creep rate and parameters of the creep constitutive equations are discussed. The results show that the thermal creep rate is greater than irradiation creep rate. The effect of temperature on the thermal creep stress exponent is relatively small at low stress, but is obvious when stress exceeds 0.8 GPa. The higher the temperature, the larger the thermal creep stress exponent. The irradiation creep rate increases almost linearly with the dose rate, that is, the exponent of dose rate for irradiation creep approach 1.0. Irradiation creep stress exponent fluctuates very little around 1.1 within the scope of the present research. Besides, higher temperature accelerates the linear increase of irradiation creep rate with dose rate, and the irradiation creep pre-factor becomes higher.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"26 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90446406","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Teng Zhang, Xubo Ma, Gu Jia, Xuan Ma, Bin Zhang, Kui Hu
In order to improve the accuracy of fast reactor physical analysis, two libraries with 1968-group neutron and 21-group photon were generated based on ENDF/B-VIII.0 and ENDF/B-VII.1 data by using NJOY2016. A code, named TXMAT2.0, was developed to process the two libraries to generate ultrafine group neutron and photon cross sections and Kinetic Energy Release in Material (KERMA) factors. To perform the verification of the two libraries, ICSBEP benchmarks for critical verification, and the sample 1D benchmark were selected. Several results were in good agreement with reference data. For the RBEC-M benchmark, the power distribution based on the ultrafine group library was good.
{"title":"Development and Verification of Neutron and Photon Ultrafine Group Library for Fast Reactor Physical Calculation","authors":"Teng Zhang, Xubo Ma, Gu Jia, Xuan Ma, Bin Zhang, Kui Hu","doi":"10.1115/icone29-92091","DOIUrl":"https://doi.org/10.1115/icone29-92091","url":null,"abstract":"\u0000 In order to improve the accuracy of fast reactor physical analysis, two libraries with 1968-group neutron and 21-group photon were generated based on ENDF/B-VIII.0 and ENDF/B-VII.1 data by using NJOY2016. A code, named TXMAT2.0, was developed to process the two libraries to generate ultrafine group neutron and photon cross sections and Kinetic Energy Release in Material (KERMA) factors. To perform the verification of the two libraries, ICSBEP benchmarks for critical verification, and the sample 1D benchmark were selected. Several results were in good agreement with reference data. For the RBEC-M benchmark, the power distribution based on the ultrafine group library was good.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"36 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88600433","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Jieming Huang, Changyuan Gao, Liu-tao Chen, Xu Wang, J. Tan
CZ2 alloy is a new Zr-Nb zirconium alloy independently developed by China Nuclear Power Research and Technology Institute (CNPRI), which has independent intellectual property rights. Zirconium will react with coolant to absorb hydrogen in reactor service, which will degrade the performance of zirconium alloy and lead to failure. Therefore, in many studies of zirconium alloy performance, hydriding is often used to simulate burnup. In this paper, CZ2 alloy with hydrogen content of about 400 ppm was obtained by gaseous hydrogen charging method. The sample was heated to 500°C with the furnace in hydrogen environment and kept for 1.5 h. Then the fatigue tests of hydrided/un-hydrided CZ2 alloy and un-hydrided Zr-4 alloy under asymmetric axial pull-pull loading at 343°C were carried out to study the effect of hydride on the fatigue properties of CZ2 alloy. The results show that with the decrease of stress level, the difference between the fatigue life of CZ2 alloy un-hydrided and that of hydrided becomes larger, and the fatigue life of CZ2 alloy un-hydrided is higher than that of hydrided. The fatigue properties of un-hydrided CZ2 alloys are better than Zr-4 alloys. The two-parameter model and three-parameter model are used to fit the fatigue data of CZ2 alloy. It is found that the three-parameter model can better describe the fatigue life of CZ2 alloy.
{"title":"Study on Fatigue Properties of Hydrided CZ2 Alloy","authors":"Jieming Huang, Changyuan Gao, Liu-tao Chen, Xu Wang, J. Tan","doi":"10.1115/icone29-93763","DOIUrl":"https://doi.org/10.1115/icone29-93763","url":null,"abstract":"\u0000 CZ2 alloy is a new Zr-Nb zirconium alloy independently developed by China Nuclear Power Research and Technology Institute (CNPRI), which has independent intellectual property rights. Zirconium will react with coolant to absorb hydrogen in reactor service, which will degrade the performance of zirconium alloy and lead to failure. Therefore, in many studies of zirconium alloy performance, hydriding is often used to simulate burnup. In this paper, CZ2 alloy with hydrogen content of about 400 ppm was obtained by gaseous hydrogen charging method. The sample was heated to 500°C with the furnace in hydrogen environment and kept for 1.5 h. Then the fatigue tests of hydrided/un-hydrided CZ2 alloy and un-hydrided Zr-4 alloy under asymmetric axial pull-pull loading at 343°C were carried out to study the effect of hydride on the fatigue properties of CZ2 alloy. The results show that with the decrease of stress level, the difference between the fatigue life of CZ2 alloy un-hydrided and that of hydrided becomes larger, and the fatigue life of CZ2 alloy un-hydrided is higher than that of hydrided. The fatigue properties of un-hydrided CZ2 alloys are better than Zr-4 alloys. The two-parameter model and three-parameter model are used to fit the fatigue data of CZ2 alloy. It is found that the three-parameter model can better describe the fatigue life of CZ2 alloy.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"123 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85353883","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
I. L. Simanullang, Katsuki Fukuhara, Keisuke Morita, Y. Fukaya, H. Ho, S. Nagasumi, K. Iigaki, E. Ishitsuka, N. Fujimoto
The ORIGEN2 code has been used for fuel depletion calculations of many kinds of reactor fuels but there is no library for high temperature gas cooled reactors (HTGRs). A set of the ORIGEN2 library for the HTGR has been established to evaluate the fuel burnup characteristics. In this study, the ORIGEN2 library was prepared for the high temperature engineering test reactor (HTTR). The HTTR is the first Japanese prismatic type HTGR. The burnup dependent neutron spectrum is necessary for generating the ORIGEN2 library. A pin-cell burnup calculation was conducted to obtain the burnup dependent neutron spectrum in the fuel compact of HTTR. Then, the ORIGEN2 library was generated based on the neutron spectrum of the pin cell model. The calculation results that were calculated by the ORIGEN2 code was validated by comparison with a detailed calculation with use of the MVP-BURN code. This code-to-code method was used to validate the ORIGEN2 code calculation because of no assay data of HTTR spent fuels. One of the isotopes that evaluated was 239Pu. The calculation results showed that the amount of 239Pu calculated by ORIGEN2 code was higher about 35% than that of calculated by the MVP-BURN code. It showed that the ORIGEN2 library using the neutron spectrum of a pin-cell burnup model was not enough for evaluating burnup characteristics of the HTTR. Therefore, an improvement was performed to evaluate the ORIGEN2 library. In this study, the ORIGEN2 library was generated based on the neutron spectrum of a core burnup calculation. The calculation results showed that the ORIGEN2 code and the MVP-BURN code was in a good agreement. The maximum difference of 239Pu amount between the ORIGEN2 and MVP-BURN became 2.4%.
{"title":"Preparation Method of ORIGEN2 Library for High Temperature Gas-Cooled Reactors","authors":"I. L. Simanullang, Katsuki Fukuhara, Keisuke Morita, Y. Fukaya, H. Ho, S. Nagasumi, K. Iigaki, E. Ishitsuka, N. Fujimoto","doi":"10.1115/icone29-90802","DOIUrl":"https://doi.org/10.1115/icone29-90802","url":null,"abstract":"\u0000 The ORIGEN2 code has been used for fuel depletion calculations of many kinds of reactor fuels but there is no library for high temperature gas cooled reactors (HTGRs). A set of the ORIGEN2 library for the HTGR has been established to evaluate the fuel burnup characteristics. In this study, the ORIGEN2 library was prepared for the high temperature engineering test reactor (HTTR). The HTTR is the first Japanese prismatic type HTGR. The burnup dependent neutron spectrum is necessary for generating the ORIGEN2 library. A pin-cell burnup calculation was conducted to obtain the burnup dependent neutron spectrum in the fuel compact of HTTR. Then, the ORIGEN2 library was generated based on the neutron spectrum of the pin cell model. The calculation results that were calculated by the ORIGEN2 code was validated by comparison with a detailed calculation with use of the MVP-BURN code. This code-to-code method was used to validate the ORIGEN2 code calculation because of no assay data of HTTR spent fuels. One of the isotopes that evaluated was 239Pu. The calculation results showed that the amount of 239Pu calculated by ORIGEN2 code was higher about 35% than that of calculated by the MVP-BURN code. It showed that the ORIGEN2 library using the neutron spectrum of a pin-cell burnup model was not enough for evaluating burnup characteristics of the HTTR. Therefore, an improvement was performed to evaluate the ORIGEN2 library. In this study, the ORIGEN2 library was generated based on the neutron spectrum of a core burnup calculation. The calculation results showed that the ORIGEN2 code and the MVP-BURN code was in a good agreement. The maximum difference of 239Pu amount between the ORIGEN2 and MVP-BURN became 2.4%.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"66 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82790007","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
China implements the established policy of closed nuclear fuel cycle for the sustainable development of nuclear power. However, there seems no feasible development plan and roadmap to initiate and deploy a commercial closed fuel cycle in China up to now. The industrialization of the nuclear fuel cycle requires gradual and phased progresses. Since most of the operating nuclear power plants in China are PWR units, and China implements and promotes a commercial closed nuclear fuel cycle, how to initiate a closed nuclear fuel cycle from the mature commercial pressurized water reactors is an unavoidable reality. Different from the implementation of closed nuclear fuel cycle reactors in countries such as France and Russia, the operating status and modes of PWRs in China are varied significantly. Most PWRs in China have implemented different plant modifications such as reactor power upgrading, core design and fuel management improvements with different new fuel types, different burnups and different cycle lengths, which have consumed certain degree of safety margins. These characteristics and differences bring challenges and difficulties to the implementation of a closed nuclear fuel cycle in China. Based on international experiences and China’s situation, this paper discusses the necessity of initiating a closed nuclear fuel cycle from mature commercial nuclear power plants in China as the initial stage of the closed fuel cycle to lay the foundation for the future advanced nuclear fuel cycle, analyze and discuss the initiating mode of China’s commercial closed nuclear fuel cycle, review the nuclear fuel types to be utilized in the closed nuclear fuel cycle, and discuss the possible configuration and development path of China’s closed fuel cycle in the future.
{"title":"Research on Scenarios and Development Paths of China’s Commercial Closed Nuclear Fuel Cycle","authors":"M. Xiao, Xiaojun Xiao","doi":"10.1115/icone29-89223","DOIUrl":"https://doi.org/10.1115/icone29-89223","url":null,"abstract":"\u0000 China implements the established policy of closed nuclear fuel cycle for the sustainable development of nuclear power. However, there seems no feasible development plan and roadmap to initiate and deploy a commercial closed fuel cycle in China up to now.\u0000 The industrialization of the nuclear fuel cycle requires gradual and phased progresses. Since most of the operating nuclear power plants in China are PWR units, and China implements and promotes a commercial closed nuclear fuel cycle, how to initiate a closed nuclear fuel cycle from the mature commercial pressurized water reactors is an unavoidable reality.\u0000 Different from the implementation of closed nuclear fuel cycle reactors in countries such as France and Russia, the operating status and modes of PWRs in China are varied significantly. Most PWRs in China have implemented different plant modifications such as reactor power upgrading, core design and fuel management improvements with different new fuel types, different burnups and different cycle lengths, which have consumed certain degree of safety margins. These characteristics and differences bring challenges and difficulties to the implementation of a closed nuclear fuel cycle in China.\u0000 Based on international experiences and China’s situation, this paper discusses the necessity of initiating a closed nuclear fuel cycle from mature commercial nuclear power plants in China as the initial stage of the closed fuel cycle to lay the foundation for the future advanced nuclear fuel cycle, analyze and discuss the initiating mode of China’s commercial closed nuclear fuel cycle, review the nuclear fuel types to be utilized in the closed nuclear fuel cycle, and discuss the possible configuration and development path of China’s closed fuel cycle in the future.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"1959 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87775597","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
The dual-cooled annular nuclear fuel is an advanced design that is expected to greatly lower fuel temperature even under high linear power density, as compared to traditional cylindrical fuel pin. Although fuel temperature can be much lower, the annular pellet also receives much higher neutron fluence, which may induce severe cracking during normal operation. This work deals with quasi-static cracking of dual-cooled annular UO2 pellet under neutron radiation. The analysis is based on the phase-field fracture model coupled with an oxygen diffusion model, heat conduction model and mechanical equilibrium model. The considered thermo-mechanical properties and irradiation behaviors of the nuclear fuel are both temperature and irradiation dependent. Especially, the acceleration of fuel creep due to oxygen redistribution is included. The fracture is represented by a scalar phase-field variable governed by a cohesive phase-field fracture method. These models are numerically implemented in the multi-physics coupling simulation framework MOOSE. For the first time, the diffusion-thermo-mechanical coupled fracture model is applied to the dual-cooled annular UO2 fuel pellet cracking during reactor startup, power ramp and reactor shutdown. Preliminarily, UO2 irradiation creep is found to play an important role on the fuel pellet fragmentation. The developed capability supports interpretation of experimental data and can guide material design of advanced ceramic nuclear fuel.
{"title":"Phase-Field Fracture Simulation of Dual-Cooled Annular Fuel Pellet","authors":"Wei Li","doi":"10.1115/icone29-92230","DOIUrl":"https://doi.org/10.1115/icone29-92230","url":null,"abstract":"\u0000 The dual-cooled annular nuclear fuel is an advanced design that is expected to greatly lower fuel temperature even under high linear power density, as compared to traditional cylindrical fuel pin. Although fuel temperature can be much lower, the annular pellet also receives much higher neutron fluence, which may induce severe cracking during normal operation. This work deals with quasi-static cracking of dual-cooled annular UO2 pellet under neutron radiation. The analysis is based on the phase-field fracture model coupled with an oxygen diffusion model, heat conduction model and mechanical equilibrium model. The considered thermo-mechanical properties and irradiation behaviors of the nuclear fuel are both temperature and irradiation dependent. Especially, the acceleration of fuel creep due to oxygen redistribution is included. The fracture is represented by a scalar phase-field variable governed by a cohesive phase-field fracture method. These models are numerically implemented in the multi-physics coupling simulation framework MOOSE. For the first time, the diffusion-thermo-mechanical coupled fracture model is applied to the dual-cooled annular UO2 fuel pellet cracking during reactor startup, power ramp and reactor shutdown. Preliminarily, UO2 irradiation creep is found to play an important role on the fuel pellet fragmentation. The developed capability supports interpretation of experimental data and can guide material design of advanced ceramic nuclear fuel.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"83 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90517505","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Lin Shi, Changyuan Gao, Guoliang Zhang, Guocheng Sun, Xu Wang, Liu-tao Chen, J. Tan
This paper is dealing with the iodine-induced stress corrosion cracking behavior of CZ2 zirconium alloy which is developed by China General Nuclear Power Group. The alloy in this study was fabricated with four different final annealing temperature in the range of 450 °C to 600 °C. In order to investigate the iodine-induced stress corrosion cracking behavior of CZ2 alloy, the slow strain rate tensile tests of four different CZ2 were conducted with three different iodine partial pressure of 0Pa, 10Pa and 10000Pa. The temperature of the tests was 350 °C and the strain rate was 1.4 × 10−6s−1. Also, the sensitivity index of iodine-induced stress corrosion cracking was calculated. The iodine-induced stress corrosion cracking sensitivity index of recrystallized CZ2 alloy is lower than that of stress-relieved CZ2 alloy, and with the increase of final annealing temperature, the sensitivity index decreases gradually. Finally, the fracture surface of CZ2 alloy was observed by scanning electron microscopy. The fracture feature of all four different CZ2 alloy changes from ductile fracture morphology to brittle fracture morphology with the increase of iodine partial pressure. Under the condition of 10000 Pa iodine partial, the fracture feature of stress-relieved CZ2 shows obvious brittle cleavage fracture, the fracture feature of partially recrystallized CZ2 is partly ductile fracture and partly cleavage fracture morphology. For recrystallized CZ2, there are many dimples in the fracture morphology, and shows obvious ductile fracture.
{"title":"Study of Iodine-Induced Stress Corrosion Cracking of CZ2 Alloy","authors":"Lin Shi, Changyuan Gao, Guoliang Zhang, Guocheng Sun, Xu Wang, Liu-tao Chen, J. Tan","doi":"10.1115/icone29-93835","DOIUrl":"https://doi.org/10.1115/icone29-93835","url":null,"abstract":"\u0000 This paper is dealing with the iodine-induced stress corrosion cracking behavior of CZ2 zirconium alloy which is developed by China General Nuclear Power Group. The alloy in this study was fabricated with four different final annealing temperature in the range of 450 °C to 600 °C. In order to investigate the iodine-induced stress corrosion cracking behavior of CZ2 alloy, the slow strain rate tensile tests of four different CZ2 were conducted with three different iodine partial pressure of 0Pa, 10Pa and 10000Pa. The temperature of the tests was 350 °C and the strain rate was 1.4 × 10−6s−1. Also, the sensitivity index of iodine-induced stress corrosion cracking was calculated. The iodine-induced stress corrosion cracking sensitivity index of recrystallized CZ2 alloy is lower than that of stress-relieved CZ2 alloy, and with the increase of final annealing temperature, the sensitivity index decreases gradually. Finally, the fracture surface of CZ2 alloy was observed by scanning electron microscopy. The fracture feature of all four different CZ2 alloy changes from ductile fracture morphology to brittle fracture morphology with the increase of iodine partial pressure. Under the condition of 10000 Pa iodine partial, the fracture feature of stress-relieved CZ2 shows obvious brittle cleavage fracture, the fracture feature of partially recrystallized CZ2 is partly ductile fracture and partly cleavage fracture morphology. For recrystallized CZ2, there are many dimples in the fracture morphology, and shows obvious ductile fracture.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"39 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"81560672","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Comparing with deterministic methods, Monte Carlo method has high precision but huge time-consuming when using for shielding design. For real deep-penetration problems, a series of variance reduction methods have been proposed and applied in related software (e.g. MCMP, SERPENT) in recent decades to overcome the drawbacks of Monte Carlo method. However, these methods still have troubles, such as the selection of correction factors and function model in biasing method. The important region division method also has time and memory consuming issues in complicated models. At present, the Consistent Adjoint-Driven Importance Sampling (CADIS) and Forward-Weighted CADIS (FW-CADIS) methods are implemented well in deeply penetrating problems. This paper presents a new efficient Monte Carlo method to solve deep-penetration problems. Contrary to traditional Monte Carlo methods, in this method, the particle trajectories that contributes to the tallies most are first determined, then the occurrence probability of the corresponding trajectory is calculated and counted. The pre-determined tracks are obtained through a serious of geometric transformations from standard tracks generated in a simple medium. The geometric transformations of tracks include rotation and stretching/shortening. Moreover, the weight correction is performed to assure the weight is unbiased. Preliminary numerical results on monolayer medium demonstrate that this method can significantly reduce calculation consumptions while retaining decent accuracies.
{"title":"The Study of a New Efficient Monte Carlo Method for Deep-Penetration Transport","authors":"Tao Zhang, Zhihong Liu, D. She, Jingxia Zhao","doi":"10.1115/icone29-92621","DOIUrl":"https://doi.org/10.1115/icone29-92621","url":null,"abstract":"\u0000 Comparing with deterministic methods, Monte Carlo method has high precision but huge time-consuming when using for shielding design. For real deep-penetration problems, a series of variance reduction methods have been proposed and applied in related software (e.g. MCMP, SERPENT) in recent decades to overcome the drawbacks of Monte Carlo method. However, these methods still have troubles, such as the selection of correction factors and function model in biasing method. The important region division method also has time and memory consuming issues in complicated models. At present, the Consistent Adjoint-Driven Importance Sampling (CADIS) and Forward-Weighted CADIS (FW-CADIS) methods are implemented well in deeply penetrating problems. This paper presents a new efficient Monte Carlo method to solve deep-penetration problems. Contrary to traditional Monte Carlo methods, in this method, the particle trajectories that contributes to the tallies most are first determined, then the occurrence probability of the corresponding trajectory is calculated and counted. The pre-determined tracks are obtained through a serious of geometric transformations from standard tracks generated in a simple medium. The geometric transformations of tracks include rotation and stretching/shortening. Moreover, the weight correction is performed to assure the weight is unbiased. Preliminary numerical results on monolayer medium demonstrate that this method can significantly reduce calculation consumptions while retaining decent accuracies.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"5 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89339226","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Lei Shi, Chao Chen, Hongjun Liu, Jiqiang Su, Honglin Zhang, Qun Liu, Yanrui Li, Jian Hu
Spent nuclear fuel is an inevitable product from the development of nuclear energy. Almost all of the fuel content is radioactive, and long systematic process are required for the safety management, which has always been an important global issue. In order to make sure that spent nuclear fuel should be safely managed in different countries developing nuclear power, IAEA is establishing a sharing system of spent nuclear fuel management by concluding joint conventions and issuing safety standards. For different countries, the United States, France and Russia with nuclear power have all established a complete policy and legislative system for spent fuel management. In the US, policy decision of open-cycle has been made, and no commercial reprocessing is being conducted. In France and Russia, closed-cycle strategy is implemented with industrial-scale reprocessing plant in operation. At present, China has become the country with the largest scale of nuclear power under construction in the world. There will be a large number of spent nuclear fuel requiring properly and safely managed. The lessons-learning of how developed countries managing spent nuclear fuel arising is important for China. The authors suggest that it is necessary to combine the top-level design to the legal practice, so that there are laws to respect during all steps of spent fuel management, and responsibilities of all parties are clear.
{"title":"Study on the Policy and Legislative System of Spent Nuclear Fuel Management in Different Developed Countries","authors":"Lei Shi, Chao Chen, Hongjun Liu, Jiqiang Su, Honglin Zhang, Qun Liu, Yanrui Li, Jian Hu","doi":"10.1115/icone29-92948","DOIUrl":"https://doi.org/10.1115/icone29-92948","url":null,"abstract":"\u0000 Spent nuclear fuel is an inevitable product from the development of nuclear energy. Almost all of the fuel content is radioactive, and long systematic process are required for the safety management, which has always been an important global issue. In order to make sure that spent nuclear fuel should be safely managed in different countries developing nuclear power, IAEA is establishing a sharing system of spent nuclear fuel management by concluding joint conventions and issuing safety standards. For different countries, the United States, France and Russia with nuclear power have all established a complete policy and legislative system for spent fuel management. In the US, policy decision of open-cycle has been made, and no commercial reprocessing is being conducted. In France and Russia, closed-cycle strategy is implemented with industrial-scale reprocessing plant in operation. At present, China has become the country with the largest scale of nuclear power under construction in the world. There will be a large number of spent nuclear fuel requiring properly and safely managed. The lessons-learning of how developed countries managing spent nuclear fuel arising is important for China. The authors suggest that it is necessary to combine the top-level design to the legal practice, so that there are laws to respect during all steps of spent fuel management, and responsibilities of all parties are clear.","PeriodicalId":36762,"journal":{"name":"Journal of Nuclear Fuel Cycle and Waste Technology","volume":"39 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2022-08-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88402910","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
京公网安备 11010802042870号
京ICP备2023020795号-1
扫码关注我们
求助内容：
应助结果提醒方式：