D. Grgić, Paulina Duckic, Vesna Benčik, Siniša Šadek
Passive Containment Filtered Vent (PCFV) was installed in Nuclear Power Plant (NPP) Krsko in 2013 as part of the safety upgrade program. It is intended for severe accident consequences prevention and mitigation by ensuring the containment integrity. In this paper, dose rates around the exhaust line of the PCFV system resulting from radioactivity release in case of a severe accident were determined in a four step methodology. The assumed severe accident scenario is a beyond design basis station blackout in NPP Krsko, which was simulated using the MELCOR code. Its results were input for the RADTRAD radiological calculations to obtain the activities released in the containment. These activities were then transformed into the gamma source intensity and spectrum using the ORIGEN-S libraries. This form of the source term is required for Monte Carlo calculations which were performed using the MCNP6.2. Two Monte Carlo calculations were performed. One for which the radiation source was modeled to emanate from the containment atmosphere and the other from the PCFV duct fluid. The main reason for the calculation was to assess limiting dose rates around PCFV duct (radiation monitor location) during actuation after severe accident. That is why the model is simple and conservative. The other task was to demonstrate that this location is not suitable for longer personnel presence in case of equipment failure during the PCFV actuation. Due to conservative assumptions, predicted dose rates are the highest expected at that location for any severe accident scenario.
{"title":"Dose Rate Assessment Around the PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko","authors":"D. Grgić, Paulina Duckic, Vesna Benčik, Siniša Šadek","doi":"10.1115/1.4062797","DOIUrl":"https://doi.org/10.1115/1.4062797","url":null,"abstract":"\u0000 Passive Containment Filtered Vent (PCFV) was installed in Nuclear Power Plant (NPP) Krsko in 2013 as part of the safety upgrade program. It is intended for severe accident consequences prevention and mitigation by ensuring the containment integrity. In this paper, dose rates around the exhaust line of the PCFV system resulting from radioactivity release in case of a severe accident were determined in a four step methodology. The assumed severe accident scenario is a beyond design basis station blackout in NPP Krsko, which was simulated using the MELCOR code. Its results were input for the RADTRAD radiological calculations to obtain the activities released in the containment. These activities were then transformed into the gamma source intensity and spectrum using the ORIGEN-S libraries. This form of the source term is required for Monte Carlo calculations which were performed using the MCNP6.2. Two Monte Carlo calculations were performed. One for which the radiation source was modeled to emanate from the containment atmosphere and the other from the PCFV duct fluid. The main reason for the calculation was to assess limiting dose rates around PCFV duct (radiation monitor location) during actuation after severe accident. That is why the model is simple and conservative. The other task was to demonstrate that this location is not suitable for longer personnel presence in case of equipment failure during the PCFV actuation. Due to conservative assumptions, predicted dose rates are the highest expected at that location for any severe accident scenario.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"76 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"74224902","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}
Safety system design and implementation is critical to the operation of any nuclear plant. For sodium cooled nuclear reactors, hazards external to the reactor core are present in the form of molten sodium that leaks through degraded piping structures. These structures are often clad in high-temperature insulation to preserve the heat needed to keep the sodium molten in the piping. While large sodium leaks are quite noticeable and often result in hazardous fire situations, small leaks of molten sodium are often masked by the shroud of insulation until a large pool of material has collected outside of the failed pipe. This study concentrated on the physical and chemical interactions between molten sodium and standard fiberglass insulation in temperatures ranging from 100 ? to 500 ?. The degradation of the insulation material begins with the volatilization of the organic binder around 250 ?, thereafter the insulation deteriorates at an advanced rate in areas that are in direct contact with the sodium. Chemical profile data was collected for a variety of samples locations that were in contact with the molten sodium, with only a slight increase in the amount of sodium present that can be attributed to the external sodium source. In this way, the molten sodium does not chemically degrade the insulation, but rather accelerates the thermal degradation of the insulation on a local scale, acting as a concentrated heat source to the insulation.
{"title":"Interactions Between Molten Sodium and Standard Pipe Insulation","authors":"D. LaBrier, Jordan Harley, Morgan Robbins","doi":"10.1115/1.4062798","DOIUrl":"https://doi.org/10.1115/1.4062798","url":null,"abstract":"\u0000 Safety system design and implementation is critical to the operation of any nuclear plant. For sodium cooled nuclear reactors, hazards external to the reactor core are present in the form of molten sodium that leaks through degraded piping structures. These structures are often clad in high-temperature insulation to preserve the heat needed to keep the sodium molten in the piping. While large sodium leaks are quite noticeable and often result in hazardous fire situations, small leaks of molten sodium are often masked by the shroud of insulation until a large pool of material has collected outside of the failed pipe. This study concentrated on the physical and chemical interactions between molten sodium and standard fiberglass insulation in temperatures ranging from 100 ? to 500 ?. The degradation of the insulation material begins with the volatilization of the organic binder around 250 ?, thereafter the insulation deteriorates at an advanced rate in areas that are in direct contact with the sodium. Chemical profile data was collected for a variety of samples locations that were in contact with the molten sodium, with only a slight increase in the amount of sodium present that can be attributed to the external sodium source. In this way, the molten sodium does not chemically degrade the insulation, but rather accelerates the thermal degradation of the insulation on a local scale, acting as a concentrated heat source to the insulation.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"159 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"75057733","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}
Keke Hou, Chao Yan, P. Wang, Changqing Cao, Jun Lin, Yanguang Cui, Junqiang Lu, Libing Zhu
As a candidate material for metallic fuel, U-Mo metal fuel pellets are the most promising. U-Mo and U-Mo-Nb alloy pellets with a certain porosity were successfully prepared by the process of hydrogenation/dehydrogenation - compression molding - argon liquid-phase sintering. In order to study the effect of Nb addition on γ phase uranium alloy fuel pellets, microstructure and thermo-properties of the samples were observed by XRD/SEM etc. Results showed that with the increase of Nb content in the pellets from the non-add to micro-adding, Nb can facilitate the diffusion of Mo into the U matrix, resulting in the formation of a metastable γ-U phase. Meanwhile, during the same liquid phase sintering process of U-Mo fuel pellets, with the increase of Nb content, the number of secondary phases in U-Mo fuel pellets gradually decreased, while the size and number of voids of the secondary phases decreased. And the distribution of voids is more uniform. The specific heat capacity and thermal diffusivity of porous γ phase uranium alloys fuel pellets with different density were measured and thermal conductivity from 373K to 873K were calculated according to the experiment results. It is suggested that the thermal conductivity will increase with the density of pellets.
{"title":"Effect of Nb On Sintering Process of Gamma Phase Uranium Alloys Fuel Pellets","authors":"Keke Hou, Chao Yan, P. Wang, Changqing Cao, Jun Lin, Yanguang Cui, Junqiang Lu, Libing Zhu","doi":"10.1115/1.4062795","DOIUrl":"https://doi.org/10.1115/1.4062795","url":null,"abstract":"\u0000 As a candidate material for metallic fuel, U-Mo metal fuel pellets are the most promising. U-Mo and U-Mo-Nb alloy pellets with a certain porosity were successfully prepared by the process of hydrogenation/dehydrogenation - compression molding - argon liquid-phase sintering. In order to study the effect of Nb addition on γ phase uranium alloy fuel pellets, microstructure and thermo-properties of the samples were observed by XRD/SEM etc. Results showed that with the increase of Nb content in the pellets from the non-add to micro-adding, Nb can facilitate the diffusion of Mo into the U matrix, resulting in the formation of a metastable γ-U phase. Meanwhile, during the same liquid phase sintering process of U-Mo fuel pellets, with the increase of Nb content, the number of secondary phases in U-Mo fuel pellets gradually decreased, while the size and number of voids of the secondary phases decreased. And the distribution of voids is more uniform. The specific heat capacity and thermal diffusivity of porous γ phase uranium alloys fuel pellets with different density were measured and thermal conductivity from 373K to 873K were calculated according to the experiment results. It is suggested that the thermal conductivity will increase with the density of pellets.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"17 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90023383","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}
Siniša Šadek, Renato Pavlinac, Karlo Ivanjko, D. Grgić
Uncertainty and sensitivity methods are increasingly used in safety analyzes of nuclear power plants to address the unreliability of input data, numerical models and, in general, the lack of knowledge regarding certain physical phenomena, in determining safety margins and acceptance criteria. The ASYST code, developed as part of an international nuclear technology ASYST Development and Training Program (ADTP) managed by Innovative Systems Software (ISS), is used to perform an uncertainty analysis of the QUENCH-02 experiment conducted at the Karlsruhe Institute of Technology. The code uses a probabilistic methodology based on the propagation of input uncertainties. The QUENCH facility contains electrically heated PWR fuel rod simulators and the aim of the experiment is to examine hydrogen source term and the behavior of the fuel rod cladding during core reflood. For selected input parameters, such as steam/water flow, electrical power and other relevant boundary conditions, it is necessary to define their probability density functions. Input databases are then prepared for individual calculations based on the selected confidence level and confidence interval. The number of performed calculations is 60, large enough to ensure at least 95% coverage of expected output results and uncertainty limits. The results of the calculations are compared with the experimental measurements. The Pearson correlation coefficient is used to obtain correlation between the input uncertain parameters and the output data. Sensitivity analyses cover the influence of variations in the heater electrical power and the steam flow rate on the hydrogen production and the maximum cladding temperature.
不确定性和敏感性方法越来越多地用于核电厂的安全分析，以解决输入数据、数值模型的不可靠性，以及在确定安全裕度和接受标准方面缺乏对某些物理现象的了解。ASYST代码是由Innovative Systems Software (ISS)管理的国际核技术ASYST开发和培训计划(ADTP)的一部分，用于对卡尔斯鲁厄理工学院进行的QUENCH-02实验进行不确定性分析。该代码使用基于输入不确定性传播的概率方法。QUENCH设施包含电加热的压水堆燃料棒模拟器，实验的目的是检查堆芯再灌注过程中氢源项和燃料棒包壳的行为。对于选定的输入参数，如蒸汽/水流、电功率等相关边界条件，需要定义其概率密度函数。然后根据所选的置信水平和置信区间准备输入数据库进行单独的计算。执行的计算次数为60，足以确保至少95%的预期输出结果和不确定性限制的覆盖率。计算结果与实验测量结果进行了比较。使用Pearson相关系数来获得输入不确定参数与输出数据之间的相关性。灵敏度分析包括加热器电功率和蒸汽流量变化对产氢量和最高包层温度的影响。
{"title":"Uncertainty and Sensitivity Evaluation of the QUENCH-02 Experiment Simulation Using the ASYST Code","authors":"Siniša Šadek, Renato Pavlinac, Karlo Ivanjko, D. Grgić","doi":"10.1115/1.4062799","DOIUrl":"https://doi.org/10.1115/1.4062799","url":null,"abstract":"\u0000 Uncertainty and sensitivity methods are increasingly used in safety analyzes of nuclear power plants to address the unreliability of input data, numerical models and, in general, the lack of knowledge regarding certain physical phenomena, in determining safety margins and acceptance criteria. The ASYST code, developed as part of an international nuclear technology ASYST Development and Training Program (ADTP) managed by Innovative Systems Software (ISS), is used to perform an uncertainty analysis of the QUENCH-02 experiment conducted at the Karlsruhe Institute of Technology. The code uses a probabilistic methodology based on the propagation of input uncertainties. The QUENCH facility contains electrically heated PWR fuel rod simulators and the aim of the experiment is to examine hydrogen source term and the behavior of the fuel rod cladding during core reflood.\u0000 For selected input parameters, such as steam/water flow, electrical power and other relevant boundary conditions, it is necessary to define their probability density functions. Input databases are then prepared for individual calculations based on the selected confidence level and confidence interval. The number of performed calculations is 60, large enough to ensure at least 95% coverage of expected output results and uncertainty limits. The results of the calculations are compared with the experimental measurements. The Pearson correlation coefficient is used to obtain correlation between the input uncertain parameters and the output data. Sensitivity analyses cover the influence of variations in the heater electrical power and the steam flow rate on the hydrogen production and the maximum cladding temperature.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"3 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-06-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"87323080","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}
T. Beuthe, Hooman Jazebizadeh, Aleksandar Vasić, A. Pegarkov, T. Kaya, H. Zahlan
Alkali metal heat pipes can be used for core cooling in advanced micro and small modular reactors. This article provides a detailed review of currently available saturation property correlations for potassium that can be utilized in heat pipe simulation models. Using these properties, numerical models will be developed and employed to simulate and compare the performance of heat pipes models to experimental results. A comprehensive comparative overview of the best available potassium saturation property correlations developed over the past century has been assembled. The results show most of the potassium property correlations needed to model a heat pipe are relatively well defined and recommendations for their use can be provided, but the findings also suggest a significant disparity in some cases.
{"title":"Development of Analysis Tools for Heat Pipes Used in the Core Cooling of Small Modular Reactors: Potassium Property Correlations","authors":"T. Beuthe, Hooman Jazebizadeh, Aleksandar Vasić, A. Pegarkov, T. Kaya, H. Zahlan","doi":"10.1115/1.4062752","DOIUrl":"https://doi.org/10.1115/1.4062752","url":null,"abstract":"\u0000 Alkali metal heat pipes can be used for core cooling in advanced micro and small modular reactors. This article provides a detailed review of currently available saturation property correlations for potassium that can be utilized in heat pipe simulation models. Using these properties, numerical models will be developed and employed to simulate and compare the performance of heat pipes models to experimental results. A comprehensive comparative overview of the best available potassium saturation property correlations developed over the past century has been assembled. The results show most of the potassium property correlations needed to model a heat pipe are relatively well defined and recommendations for their use can be provided, but the findings also suggest a significant disparity in some cases.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-06-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"90890359","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}
Modeling and simulation lie at the heart of the design process of any nuclear application. An accurate representation of the radiation environment ensures not only the feasibility of new technologies, but it also aids in operation, maintenance, and even decommissioning. With increasingly complex designs, high-fidelity models have become a necessity for design maturity. McCAD has been under development for many years at Karlsruhe Institute of Technology (KIT) to facilitate the process of generating suitable models for nuclear analyses. In this paper, an overview of the major advances in the new version of the code is presented. A novel conversion algorithm has proven to be robust in significantly reducing the processing time to generate radiation transport models, making it easier to iterate on design details. A first-of-a-kind capability to generate hierarchical void cells is also discussed with preliminary analysis showing performance gains for particle tracking.
{"title":"A Novel Algorithm for CAD to CSG Conversion in McCAD","authors":"M. Harb, D. Leichtle, U. Fischer","doi":"10.3390/jne4020031","DOIUrl":"https://doi.org/10.3390/jne4020031","url":null,"abstract":"Modeling and simulation lie at the heart of the design process of any nuclear application. An accurate representation of the radiation environment ensures not only the feasibility of new technologies, but it also aids in operation, maintenance, and even decommissioning. With increasingly complex designs, high-fidelity models have become a necessity for design maturity. McCAD has been under development for many years at Karlsruhe Institute of Technology (KIT) to facilitate the process of generating suitable models for nuclear analyses. In this paper, an overview of the major advances in the new version of the code is presented. A novel conversion algorithm has proven to be robust in significantly reducing the processing time to generate radiation transport models, making it easier to iterate on design details. A first-of-a-kind capability to generate hierarchical void cells is also discussed with preliminary analysis showing performance gains for particle tracking.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"38 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-06-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79217950","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}
M. Khalaquzzaman, Seung Jun Lee, Muhammed Mufazzal Hossen
Quality and safety are intensely related and go hand in hand. Quality of the safety-grade equipment is very important for the safety of a nuclear power plant (NPP) and achieving production goals. During manufacturing of plant components or equipment, deviation from the design might occur at different stages of manufacturing for various reasons, such as a lack of skilled manpower, deviation of materials, human errors, malfunction of equipment, violation of manufacturing procedure, etc. These deviations can be assessed cautiously and taken into consideration in the final safety analysis report (FSAR) before issuing an operating license. In this paper, we propose a Bayesian belief network for quality assessment of safety class equipment of NPPs with a few examples. The proposed procedure is a holistic approach for estimation of equipment failure probability considering manufacturing deviations and errors. Case studies for safety-class dry transformers and reactor pressurizers employing the proposed method are also presented in this article. This study provides insights for probabilistic safety assessment engineers and nuclear plant regulators for improved assessment of NPP safety.
{"title":"Reliability Assessment of NPP Safety Class Equipment Considering the Manufacturing Quality Assurance Process","authors":"M. Khalaquzzaman, Seung Jun Lee, Muhammed Mufazzal Hossen","doi":"10.3390/jne4020030","DOIUrl":"https://doi.org/10.3390/jne4020030","url":null,"abstract":"Quality and safety are intensely related and go hand in hand. Quality of the safety-grade equipment is very important for the safety of a nuclear power plant (NPP) and achieving production goals. During manufacturing of plant components or equipment, deviation from the design might occur at different stages of manufacturing for various reasons, such as a lack of skilled manpower, deviation of materials, human errors, malfunction of equipment, violation of manufacturing procedure, etc. These deviations can be assessed cautiously and taken into consideration in the final safety analysis report (FSAR) before issuing an operating license. In this paper, we propose a Bayesian belief network for quality assessment of safety class equipment of NPPs with a few examples. The proposed procedure is a holistic approach for estimation of equipment failure probability considering manufacturing deviations and errors. Case studies for safety-class dry transformers and reactor pressurizers employing the proposed method are also presented in this article. This study provides insights for probabilistic safety assessment engineers and nuclear plant regulators for improved assessment of NPP safety.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"1 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-06-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"89622535","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}
A. Shmelev, N. Geraskin, V. Apse, V. Glebov, E. Kulikov, A. Krasnoborodko
This paper presents the results obtained from numerical evaluations for the possibility of large-scale 238Pu production in the light-water VVER-1000 reactor and the reactivity effect caused by the loss-of-coolant accident in the central fuel assembly of the reactor core. This fuel assembly containing the Np–Am-component of minor actinides was placed in the center of the reactor core and intended for intense production of 238Pu. Optimal conditions were found for large-scale production of plutonium with an isotope composition suitable for application in radioisotope thermoelectric generators. The reactivity effect from the loss-of-coolant accident in the central Np–Am fuel assembly was evaluated, and the perturbation theory was used to determine the contributions of some neutron processes (leakage, absorption, and moderation) to the total variation of the effective neutron multiplication factor.
{"title":"Application of Np–Am Mixture in Production of 238Pu in a VVER-1000 Reactor and the Reactivity Effect Caused by Loss-of-Coolant Accident in the Central Np–Am Fuel Assembly","authors":"A. Shmelev, N. Geraskin, V. Apse, V. Glebov, E. Kulikov, A. Krasnoborodko","doi":"10.3390/jne4020029","DOIUrl":"https://doi.org/10.3390/jne4020029","url":null,"abstract":"This paper presents the results obtained from numerical evaluations for the possibility of large-scale 238Pu production in the light-water VVER-1000 reactor and the reactivity effect caused by the loss-of-coolant accident in the central fuel assembly of the reactor core. This fuel assembly containing the Np–Am-component of minor actinides was placed in the center of the reactor core and intended for intense production of 238Pu. Optimal conditions were found for large-scale production of plutonium with an isotope composition suitable for application in radioisotope thermoelectric generators. The reactivity effect from the loss-of-coolant accident in the central Np–Am fuel assembly was evaluated, and the perturbation theory was used to determine the contributions of some neutron processes (leakage, absorption, and moderation) to the total variation of the effective neutron multiplication factor.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"50 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77731252","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}
Xingyu Zhao, Minyun Liu, Rongyi Cui, Shanfang Huang, Kan Wang, Chuan Lu
Abstract This study analyzed an yttrium hydride (YH2) moderated supercritical carbon dioxide cooled reactor loaded with pin-type, beryllium oxide diluted oxide fuel elements to reduce the critical enrichment. The impact of the YH2 on the coolant void reactivity was studied along with a moderator zoning scheme to flatten the radial power distribution. The YH2 was added as hexagonal moderating rods at the center of the fuel assemblies. The core was modeled using the continuous-energy Reactor Monte Carlo code (RMC) with the on-the-fly cross sections treatment. The results showed that the YH2 moderator increased the thermal fission and reduced the critical enrichment of the core with the same diluent volume fraction by more than 30%. The YH2 moderator significantly softened the neutron energy spectrum and reduced the neutron leakage upon core voiding, resulting in both a weaker positive spectral reactivity feedback and a weaker negative leakage reactivity feedback during core depressurization. For an UO2-loaded core, the YH2 gave a lower negative coolant void reactivity, while for a mixed oxide fuel (MOX)-loaded core with diluent volume fractions smaller than 35%, the spectral feedback was more important and the YH2 strongly reduced the positive coolant void reactivity to less than $1. Arranging the YH2 in the peripheral assemblies reduced the radial power peaking factor to 1.319. The study shows that the YH2 moderator can reduce the critical enrichment, make the core less sensitive to voiding, and can flatten the radial power distribution of a single-enrichment core through moderator zoning.
{"title":"Optimized Moderator Design and Analysis of a Pin-Type Supercritical Carbon Dioxide Reactor Based on Reactor Monte Carlo Code","authors":"Xingyu Zhao, Minyun Liu, Rongyi Cui, Shanfang Huang, Kan Wang, Chuan Lu","doi":"10.1115/1.4056772","DOIUrl":"https://doi.org/10.1115/1.4056772","url":null,"abstract":"Abstract This study analyzed an yttrium hydride (YH2) moderated supercritical carbon dioxide cooled reactor loaded with pin-type, beryllium oxide diluted oxide fuel elements to reduce the critical enrichment. The impact of the YH2 on the coolant void reactivity was studied along with a moderator zoning scheme to flatten the radial power distribution. The YH2 was added as hexagonal moderating rods at the center of the fuel assemblies. The core was modeled using the continuous-energy Reactor Monte Carlo code (RMC) with the on-the-fly cross sections treatment. The results showed that the YH2 moderator increased the thermal fission and reduced the critical enrichment of the core with the same diluent volume fraction by more than 30%. The YH2 moderator significantly softened the neutron energy spectrum and reduced the neutron leakage upon core voiding, resulting in both a weaker positive spectral reactivity feedback and a weaker negative leakage reactivity feedback during core depressurization. For an UO2-loaded core, the YH2 gave a lower negative coolant void reactivity, while for a mixed oxide fuel (MOX)-loaded core with diluent volume fractions smaller than 35%, the spectral feedback was more important and the YH2 strongly reduced the positive coolant void reactivity to less than $1. Arranging the YH2 in the peripheral assemblies reduced the radial power peaking factor to 1.319. The study shows that the YH2 moderator can reduce the critical enrichment, make the core less sensitive to voiding, and can flatten the radial power distribution of a single-enrichment core through moderator zoning.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"24 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2023-05-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135195823","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}
A. Saraswat, R. Bhattacharyay, P. Chaudhuri, S. Gedupudi
Liquid metals (LMs) exhibit several key characteristics justifying their utilization as coolants/breeders for nuclear fusion reactors and advanced fission reactors. In a fusion reactor, LMs confront a large flow retarding MHD force, imposing significant demands on pumping power and designs of ancillary systems. Corrosion of structural materials and coolant chemistry control are vital issues common to both fusion and fission reactors employing liquid lead (Pb) and its alloys. To address these concerns, technological solutions such as Flow Channel Inserts and corrosion resistant coatings are being investigated to provide a chemical/electrical isolation between LM and structural material. In this study, three prototype geometries (circular, square and 90 bend) of steel-insulator-steel FCIs are fabricated and an electrical insulation characterization is performed over a temperature range of 100C - 600C. Welding trials and pressure tests are performed to validate the electrical and mechanical integrity over typical fusion reactor operational regime. This paper presents detailed fabrication aspects along with quantitative estimations of insulation filling density, electrical insulation performance and, for the first time, a systematic study of insulation degradation owing to combined effects of TIG welding, pressure and machining operations. Critical details derived from metallurgical examinations and destructive tests are also presented. From implementation perspective towards LFRs, a feasibility assessment of a-Al2O3/AlPO4 thin film coating deposition on planar and non-planar substrates is performed followed by its mechanical characterizations. Detailed metallurgical analyses are presented to assess Pb ingress after 700 hour exposure to molten Pb alloy.
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