Abstract It is advised to use a new formula to calculate the (p, n) reaction cross section at 7.5 MeV. We propose six new parameters for the formula proposed by Broeders, C. and Konobeyev, A.Y. (2008. Systematics of (p,n) reaction crosssection. Radiochim. Acta 96: 387–397) to fit experimental data. It should be noted that our systematics is only applicable to isotopes with a ratio of 7.5 MeV to the reaction threshold above 1.3 (7.5/E th > 1.3). This is based on analytical calculations generated from the semi-empirical mass formula, the evaporation model, and the pre-equilibrium exciton model. We were able to find new parameters for the Broders et al. formula through this inquiry that guarantee a good fit with the revised experimental data (EXFOR2022) and provide a minimum value for the statistical parameters ∑ and χ2.
{"title":"New semi-empirical systematic of (p,n) reaction cross section at 7.5 MeV","authors":"Omeir Lyes, N. Amrani","doi":"10.1515/kern-2022-0091","DOIUrl":"https://doi.org/10.1515/kern-2022-0091","url":null,"abstract":"Abstract It is advised to use a new formula to calculate the (p, n) reaction cross section at 7.5 MeV. We propose six new parameters for the formula proposed by Broeders, C. and Konobeyev, A.Y. (2008. Systematics of (p,n) reaction crosssection. Radiochim. Acta 96: 387–397) to fit experimental data. It should be noted that our systematics is only applicable to isotopes with a ratio of 7.5 MeV to the reaction threshold above 1.3 (7.5/E th > 1.3). This is based on analytical calculations generated from the semi-empirical mass formula, the evaporation model, and the pre-equilibrium exciton model. We were able to find new parameters for the Broders et al. formula through this inquiry that guarantee a good fit with the revised experimental data (EXFOR2022) and provide a minimum value for the statistical parameters ∑ and χ2.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"44 1","pages":"279 - 290"},"PeriodicalIF":0.5,"publicationDate":"2023-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79960430","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract To design a compact heat sink, a simplified geometry, enhanced heat dissipation, and the minimum pressure drop should be taken into consideration. With this objective, an experimental investigation has been conducted with the corrugated plate-fin heat sink by varying the relative radius of corrugation and relative corrugation pitch in the range of 0.16–0.31, and 0.06–0.16, respectively, for the Reynolds number range of 6000–14,000. Experiments were conducted on a corrugated plate-fin heat sink using an open-loop experimental system comprising a test section of a rectangular channel measuring 2300 mm long, 180 mm wide, and 80 mm high. The corrugated fin creates higher disturbances caused by multiple separations and reattachments in the flow and thereby yielding a higher localized heat transfer coefficient and enhanced heat transfer from the system. The maximum fin performance is found to be 5.87 for the corrugated plate-fin heat sink corresponding to the relative radius of corrugation and relative corrugation pitch of 0.16 and 0.125, respectively.
{"title":"Enhanced heat transfer in corrugated plate fin heat sink","authors":"Alen Mathew Jose, M. Kumar, A. Patil","doi":"10.1515/kern-2022-0114","DOIUrl":"https://doi.org/10.1515/kern-2022-0114","url":null,"abstract":"Abstract To design a compact heat sink, a simplified geometry, enhanced heat dissipation, and the minimum pressure drop should be taken into consideration. With this objective, an experimental investigation has been conducted with the corrugated plate-fin heat sink by varying the relative radius of corrugation and relative corrugation pitch in the range of 0.16–0.31, and 0.06–0.16, respectively, for the Reynolds number range of 6000–14,000. Experiments were conducted on a corrugated plate-fin heat sink using an open-loop experimental system comprising a test section of a rectangular channel measuring 2300 mm long, 180 mm wide, and 80 mm high. The corrugated fin creates higher disturbances caused by multiple separations and reattachments in the flow and thereby yielding a higher localized heat transfer coefficient and enhanced heat transfer from the system. The maximum fin performance is found to be 5.87 for the corrugated plate-fin heat sink corresponding to the relative radius of corrugation and relative corrugation pitch of 0.16 and 0.125, respectively.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"2008 1","pages":"262 - 272"},"PeriodicalIF":0.5,"publicationDate":"2023-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86243719","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract To address the issue identified in USNRC’s Generic Letter (GL) 2003-01 that the unfiltered air in-leakage rate through plant’s control room envelope during design basis accident may exceed that assumed in the licensing analysis and thus threat the control room habitability, the control room envelope unfiltered in-leakage Tracer Gas Test (TGT) for Maanshan nuclear power plant (NPP) has been performed in 2020. For future applications, an improved control room dose analysis approach using Alternative Source Terms (AST) has been developed in present study to check whether the TGT results meet regulatory limits and have sufficient safety margins. The AST method follows Regulatory Guide 1.183 (RG 1.183) and the TGT results must be fulfilled the total effective dose criteria set forth in 10 CFR 50.67. Based on the AST approach and the TGT results, the control room personnel dose for Maanshan NPP during Loss of Coolant Accident (LOCA) is 14.35 mSv, and the safety margin is 71.3%. It is sufficient to cover the effects of structural ageing and changes in meteorological data during the control room habitability reassessment and the analysis uncertainty.
{"title":"Post-LOCA control room dose analysis for Maanshan NPP using the AST methodology","authors":"Cheng-Der Wang","doi":"10.1515/kern-2022-0089","DOIUrl":"https://doi.org/10.1515/kern-2022-0089","url":null,"abstract":"Abstract To address the issue identified in USNRC’s Generic Letter (GL) 2003-01 that the unfiltered air in-leakage rate through plant’s control room envelope during design basis accident may exceed that assumed in the licensing analysis and thus threat the control room habitability, the control room envelope unfiltered in-leakage Tracer Gas Test (TGT) for Maanshan nuclear power plant (NPP) has been performed in 2020. For future applications, an improved control room dose analysis approach using Alternative Source Terms (AST) has been developed in present study to check whether the TGT results meet regulatory limits and have sufficient safety margins. The AST method follows Regulatory Guide 1.183 (RG 1.183) and the TGT results must be fulfilled the total effective dose criteria set forth in 10 CFR 50.67. Based on the AST approach and the TGT results, the control room personnel dose for Maanshan NPP during Loss of Coolant Accident (LOCA) is 14.35 mSv, and the safety margin is 71.3%. It is sufficient to cover the effects of structural ageing and changes in meteorological data during the control room habitability reassessment and the analysis uncertainty.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":" 26","pages":"221 - 230"},"PeriodicalIF":0.5,"publicationDate":"2023-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"72380337","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract Spacer grids with mixing vanes have complex geometry and are used to support the fuel rods in nuclear fuel assemblies as well as to improve heat transfer by generating turbulence downstream of the spacer grid in the cores of the pressurized water reactors. To validate the CFD code OpenFOAM for relevant spacer grid geometries, numerical analyses on the OECD/NEA MATiS-H benchmark were performed. The flow behind two different types of spacer grid designs was analysed: split- and swirl-type. Initially, an appropriate inlet velocity profile was generated. In a next step, Computer-Aided Design models of the spacer grids were prepared and then meshed using ANSYS Mesher 19.2. Transient URANS simulations were performed with the k-ω-SST turbulence model and the results were compared with data. Good agreement was obtained for the mean velocity profile and the vorticity in the swirl-type configuration, while the numerical results slightly overestimated the transverse velocity profile at some measurements locations of the split-type configuration. The content of this article was initially presented at the 33rd German CFD Network of Competence Meeting, held on March 22–23 at GRS in Garching, Germany.
{"title":"Numerical investigations of flow in nuclear fuel assembly with spacer grid and OpenFOAM validation","authors":"Hemish Mistry","doi":"10.1515/kern-2022-0109","DOIUrl":"https://doi.org/10.1515/kern-2022-0109","url":null,"abstract":"Abstract Spacer grids with mixing vanes have complex geometry and are used to support the fuel rods in nuclear fuel assemblies as well as to improve heat transfer by generating turbulence downstream of the spacer grid in the cores of the pressurized water reactors. To validate the CFD code OpenFOAM for relevant spacer grid geometries, numerical analyses on the OECD/NEA MATiS-H benchmark were performed. The flow behind two different types of spacer grid designs was analysed: split- and swirl-type. Initially, an appropriate inlet velocity profile was generated. In a next step, Computer-Aided Design models of the spacer grids were prepared and then meshed using ANSYS Mesher 19.2. Transient URANS simulations were performed with the k-ω-SST turbulence model and the results were compared with data. Good agreement was obtained for the mean velocity profile and the vorticity in the swirl-type configuration, while the numerical results slightly overestimated the transverse velocity profile at some measurements locations of the split-type configuration. The content of this article was initially presented at the 33rd German CFD Network of Competence Meeting, held on March 22–23 at GRS in Garching, Germany.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"12 1","pages":"141 - 154"},"PeriodicalIF":0.5,"publicationDate":"2023-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82175668","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract In this study we have explored 6Li + 7Li fusion evaporation reactions cross sections dependencies on both nuclear level density and various spin combination effects. The reaction cross section was calculated in the energy range of 0.1–16 MeV projectile of 6Li on the fixed target of 7Li. The excited compound nucleus (13C) can decay into various channels, and its decay rate in any given channel is proportional to the available phase space, i.e., the corresponding level density of it which is explained in the present study. In the present study, LISE++, PACE4, NRV and GEMINI codes were used to determine cross section of evaporation residues cross sections of 13C.
{"title":"Lithium–lithium fusion evaporation research","authors":"H. Aksakal, E. Yıldız","doi":"10.1515/kern-2022-0104","DOIUrl":"https://doi.org/10.1515/kern-2022-0104","url":null,"abstract":"Abstract In this study we have explored 6Li + 7Li fusion evaporation reactions cross sections dependencies on both nuclear level density and various spin combination effects. The reaction cross section was calculated in the energy range of 0.1–16 MeV projectile of 6Li on the fixed target of 7Li. The excited compound nucleus (13C) can decay into various channels, and its decay rate in any given channel is proportional to the available phase space, i.e., the corresponding level density of it which is explained in the present study. In the present study, LISE++, PACE4, NRV and GEMINI codes were used to determine cross section of evaporation residues cross sections of 13C.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"238 1","pages":"231 - 239"},"PeriodicalIF":0.5,"publicationDate":"2023-03-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76869807","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) with reference to the Advanced Power Reactor 1400 MW (APR1400) for transient and design basis accidents (DBAs) simulation. An experiment for a 4-inch cold leg top-slot break was performed at ATLAS to resolve a safety issue that the loop seal reformation (LSR) of APR1400 could lead to the increase of peak cladding temperature. In addition, the experimental data has been utilized to validate system codes within a framework of domestic standard problem (DSP) program organized by KAERI in collaboration with Korea Institute of Nuclear Safety (KINS). In this study, the experiment has been analyzed by thermal-hydraulic system analysis code, MARS-KS 1.5 and a comparison with experimental and calculation results has been performed. Since the top-slot break is not a typical break geometry for safety analyses, this study aims at examining the applicability of MARS-KS to the top-slot break accident where the LSR occurs repeatedly. The results revealed that overall physical behavior during the accident was predicted by the code, appropriately. MARS-KS showed the excursion of the peak cladding temperature because of the LSR as in experiment. It has been confirmed that the core integrity was maintained because the temperature excursion by the LSR was not large enough to alter the acceptance criteria. In addition, it is presented the results of the sensitivity analysis of parameter that affect the figure of merits.
{"title":"Analysis of 4-inch cold leg top-slot break LOCA in ATLAS experimental facility using MARS-KS","authors":"HyunJoon Jeong, Taewan Kim","doi":"10.1515/kern-2022-0116","DOIUrl":"https://doi.org/10.1515/kern-2022-0116","url":null,"abstract":"Abstract Korea Atomic Energy Research Institute (KAERI) has operated an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS) with reference to the Advanced Power Reactor 1400 MW (APR1400) for transient and design basis accidents (DBAs) simulation. An experiment for a 4-inch cold leg top-slot break was performed at ATLAS to resolve a safety issue that the loop seal reformation (LSR) of APR1400 could lead to the increase of peak cladding temperature. In addition, the experimental data has been utilized to validate system codes within a framework of domestic standard problem (DSP) program organized by KAERI in collaboration with Korea Institute of Nuclear Safety (KINS). In this study, the experiment has been analyzed by thermal-hydraulic system analysis code, MARS-KS 1.5 and a comparison with experimental and calculation results has been performed. Since the top-slot break is not a typical break geometry for safety analyses, this study aims at examining the applicability of MARS-KS to the top-slot break accident where the LSR occurs repeatedly. The results revealed that overall physical behavior during the accident was predicted by the code, appropriately. MARS-KS showed the excursion of the peak cladding temperature because of the LSR as in experiment. It has been confirmed that the core integrity was maintained because the temperature excursion by the LSR was not large enough to alter the acceptance criteria. In addition, it is presented the results of the sensitivity analysis of parameter that affect the figure of merits.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"3 1","pages":"316 - 325"},"PeriodicalIF":0.5,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82150366","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract This work is focused on the development and validation of models and methods for the simulation of wall boiling in nuclear engineering applications with the computational fluid dynamics (CFD) code OpenFOAM. The new chtMultiRegionReactingTwoPhaseEulerFoam solver was developed based on the reactingTwoPhaseEulerFoam solver of OpenFOAM Foundation version 7. The solver is used for the simulation of two-phase flow under consideration of wall boiling and conjugate heat transfer (CHT) between solid structure and two-phase fluid regions. The Euler–Euler approach for two-phase flows was used. The heat flux during wall boiling was calculated with the help of the extended Rensselaer Polytechnic Institute wall heat flux partitioning model, in which the convective heat flux between solid wall and two-phase flow with high void fractions was also considered. The solver was validated against experimental data from the OECD/NEA PWR Subchannel and Bundle Tests benchmark. This Nuclear Power Energy Corporation (NUPEC) database provides data for different fuel assembly subchannel geometries at different thermal-hydraulic conditions. 10 experimental runs with different boundary conditions of the benchmark exercise I-1 were simulated with the chtMultiRegionReactingTwoPhaseEulerFoam solver. The solver showed good numerical stability in all examined cases, which captured different boiling regimes with up to cross-section averaged void fractions of 0.6. The results were compared with measured data for the averaged over the cross-section of the investigated geometry void fractions. Good agreement with experimental data was observed.
摘要本工作的重点是利用计算流体力学(CFD)代码OpenFOAM开发和验证核工程应用中壁沸腾模拟的模型和方法。新的chtMultiRegionReactingTwoPhaseEulerFoam求解器是在OpenFOAM Foundation version 7的reactingTwoPhaseEulerFoam求解器的基础上开发的。利用该求解器对考虑壁面沸腾和固体结构与两相流体区域之间的共轭传热的两相流动进行了模拟。采用欧拉-欧拉法求解两相流。采用扩展的Rensselaer理工学院壁面热流分配模型计算壁面沸腾过程的热流密度，其中考虑了固体壁面与高空隙率两相流之间的对流热流密度。求解器根据OECD/NEA压水堆子通道和束测试基准的实验数据进行了验证。这个核能公司(NUPEC)数据库提供了不同热工条件下不同燃料组件子通道几何形状的数据。利用chtMultiRegionReactingTwoPhaseEulerFoam求解器对基准练习I-1中不同边界条件下的10个实验运行进行了模拟。求解器在所有测试的情况下都表现出良好的数值稳定性，在不同的沸腾状态下，截面平均孔隙分数高达0.6。研究结果与实测数据进行了比较，得到了所研究几何孔隙分数截面上的平均值。与实验数据吻合较好。
{"title":"Numerical simulation of subcooled flow boiling for nuclear engineering applications using OpenFOAM","authors":"Zhi Yang, J. Herb","doi":"10.1515/kern-2022-0112","DOIUrl":"https://doi.org/10.1515/kern-2022-0112","url":null,"abstract":"Abstract This work is focused on the development and validation of models and methods for the simulation of wall boiling in nuclear engineering applications with the computational fluid dynamics (CFD) code OpenFOAM. The new chtMultiRegionReactingTwoPhaseEulerFoam solver was developed based on the reactingTwoPhaseEulerFoam solver of OpenFOAM Foundation version 7. The solver is used for the simulation of two-phase flow under consideration of wall boiling and conjugate heat transfer (CHT) between solid structure and two-phase fluid regions. The Euler–Euler approach for two-phase flows was used. The heat flux during wall boiling was calculated with the help of the extended Rensselaer Polytechnic Institute wall heat flux partitioning model, in which the convective heat flux between solid wall and two-phase flow with high void fractions was also considered. The solver was validated against experimental data from the OECD/NEA PWR Subchannel and Bundle Tests benchmark. This Nuclear Power Energy Corporation (NUPEC) database provides data for different fuel assembly subchannel geometries at different thermal-hydraulic conditions. 10 experimental runs with different boundary conditions of the benchmark exercise I-1 were simulated with the chtMultiRegionReactingTwoPhaseEulerFoam solver. The solver showed good numerical stability in all examined cases, which captured different boiling regimes with up to cross-section averaged void fractions of 0.6. The results were compared with measured data for the averaged over the cross-section of the investigated geometry void fractions. Good agreement with experimental data was observed.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"117 1","pages":"174 - 185"},"PeriodicalIF":0.5,"publicationDate":"2023-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79755582","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Abstract The control drum is usually placed in the periphery of the active zone of the reactor core, and the control drum worth is generally small. This paper discusses different schemes to improve the control drum worth by the combination of neutron reflectors and absorbers, and the design optimization of the drums under the condition that external dimensions of the drums and positions of the drums relative to the core are fixed. The results show that BeO as the reflector, B4C and Gadolinium as the neutron absorber can obtain relatively higher drum worth. In addition, the control drum worth can also be appropriately improved by using a double-layer combination of moderator material and neutron absorber material, but it cannot be effectively improved by enlarging the central angle of the neutron absorber sector.
{"title":"Discussion of options to increase the control drum worth in fast reactor","authors":"Huaping Mei, Chao Chen, Taosheng Li","doi":"10.1515/kern-2022-0093","DOIUrl":"https://doi.org/10.1515/kern-2022-0093","url":null,"abstract":"Abstract The control drum is usually placed in the periphery of the active zone of the reactor core, and the control drum worth is generally small. This paper discusses different schemes to improve the control drum worth by the combination of neutron reflectors and absorbers, and the design optimization of the drums under the condition that external dimensions of the drums and positions of the drums relative to the core are fixed. The results show that BeO as the reflector, B4C and Gadolinium as the neutron absorber can obtain relatively higher drum worth. In addition, the control drum worth can also be appropriately improved by using a double-layer combination of moderator material and neutron absorber material, but it cannot be effectively improved by enlarging the central angle of the neutron absorber sector.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"95 1","pages":"273 - 278"},"PeriodicalIF":0.5,"publicationDate":"2023-03-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"80419042","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
W. Hsu, Jung-Hua Yang, Hsuan-Che Chen, Shao-Wen Chen, Jong-Rong Wang
Abstract The Unit 1 of Kuosheng nuclear power plant is in the decommissioning transition phase and still has spent fuel in the core of reactor now. The time to reach the TAF (top of active fuel) for water level and reach 600/800 °C for cladding temperature are the key parameters in the safety analysis, these affect that the plant has how much time to handle transients. Therefore, to study the water level and cladding temperature for station blackout (SBO) and loss of coolant accident (LOCA) transients in the decommissioning transition phase, the analysis model of Kuosheng nuclear power plant was established by using TRACE code. To evaluate the effect of the decay heat of spent fuel on the water level and cladding temperature, the sensitivity analysis of decay heat was also performed in this study.
{"title":"Using TRACE to establish the analysis model of Kuosheng nuclear power plant for decommissioning transition phase","authors":"W. Hsu, Jung-Hua Yang, Hsuan-Che Chen, Shao-Wen Chen, Jong-Rong Wang","doi":"10.1515/kern-2022-0103","DOIUrl":"https://doi.org/10.1515/kern-2022-0103","url":null,"abstract":"Abstract The Unit 1 of Kuosheng nuclear power plant is in the decommissioning transition phase and still has spent fuel in the core of reactor now. The time to reach the TAF (top of active fuel) for water level and reach 600/800 °C for cladding temperature are the key parameters in the safety analysis, these affect that the plant has how much time to handle transients. Therefore, to study the water level and cladding temperature for station blackout (SBO) and loss of coolant accident (LOCA) transients in the decommissioning transition phase, the analysis model of Kuosheng nuclear power plant was established by using TRACE code. To evaluate the effect of the decay heat of spent fuel on the water level and cladding temperature, the sensitivity analysis of decay heat was also performed in this study.","PeriodicalId":17787,"journal":{"name":"Kerntechnik","volume":"28 1","pages":"213 - 220"},"PeriodicalIF":0.5,"publicationDate":"2023-03-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"82587541","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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