Journal of Nuclear Engineering and Radiation Science最新文献_第6页

Dose Rate Assessment Around the PCFV Release Line During Severe Accident Conditions in Nuclear Power Plant Krsko 核电站严重事故条件下PCFV释放线周围的剂量率评估

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-06-20 DOI: 10.1115/1.4062797

D. Grgić, Paulina Duckic, Vesna Benčik, Siniša Šadek

{"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}

引用次数: 0

Interactions Between Molten Sodium and Standard Pipe Insulation 熔融钠与标准管道保温的相互作用

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-06-20 DOI: 10.1115/1.4062798

D. LaBrier, Jordan Harley, Morgan Robbins

{"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}

引用次数: 0

Effect of Nb On Sintering Process of Gamma Phase Uranium Alloys Fuel Pellets 铌对γ相铀合金燃料球团烧结过程的影响

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-06-20 DOI: 10.1115/1.4062795

Keke Hou, Chao Yan, P. Wang, Changqing Cao, Jun Lin, Yanguang Cui, Junqiang Lu, Libing Zhu

{"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}

引用次数: 0

Uncertainty and Sensitivity Evaluation of the QUENCH-02 Experiment Simulation Using the ASYST Code 基于ASYST代码的淬火-02试验模拟的不确定度和灵敏度评价

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-06-20 DOI: 10.1115/1.4062799

Siniša Šadek, Renato Pavlinac, Karlo Ivanjko, D. Grgić

{"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}

引用次数: 0

Development of Analysis Tools for Heat Pipes Used in the Core Cooling of Small Modular Reactors: Potassium Property Correlations 小型模块堆堆芯冷却用热管分析工具的发展:钾的性质相关性

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Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-06-16 DOI: 10.1115/1.4062752

T. Beuthe, Hooman Jazebizadeh, Aleksandar Vasić, A. Pegarkov, T. Kaya, H. Zahlan

引用次数: 0

A Novel Algorithm for CAD to CSG Conversion in McCAD McCAD中CAD到CSG转换的新算法

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Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-06-15 DOI: 10.3390/jne4020031

M. Harb, D. Leichtle, U. Fischer

引用次数: 0

Reliability Assessment of NPP Safety Class Equipment Considering the Manufacturing Quality Assurance Process 考虑制造质量保证过程的核电厂安全级设备可靠性评估

IF 0.4

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-06-02 DOI: 10.3390/jne4020030

M. Khalaquzzaman, Seung Jun Lee, Muhammed Mufazzal Hossen

{"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}

引用次数: 0

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 Np-Am混合物在VVER-1000反应堆238Pu生产中的应用及中央Np-Am燃料组件冷却剂丢失事故引起的反应性影响

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Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-06-01 DOI: 10.3390/jne4020029

A. Shmelev, N. Geraskin, V. Apse, V. Glebov, E. Kulikov, A. Krasnoborodko

引用次数: 0

Optimized Moderator Design and Analysis of a Pin-Type Supercritical Carbon Dioxide Reactor Based on Reactor Monte Carlo Code 基于蒙特卡罗代码的针式超临界二氧化碳反应器慢化剂优化设计与分析

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-05-31 DOI: 10.1115/1.4056772

Xingyu Zhao, Minyun Liu, Rongyi Cui, Shanfang Huang, Kan Wang, Chuan Lu

{"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}

引用次数: 0

Fabrication Aspects and Performance Characterization of α-Al2O3/ALPO4 Based Sandwich Configuration Flow Channel Inserts and Coatings for High Temperature Liquid Metal Applications 高温液态金属用α-Al2O3/ALPO4夹层结构流道嵌套和涂层的制备方法及性能表征

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Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-05-27 DOI: 10.1115/1.4062646

A. Saraswat, R. Bhattacharyay, P. Chaudhuri, S. Gedupudi

{"title":"Fabrication Aspects and Performance Characterization of α-Al2O3/ALPO4 Based Sandwich Configuration Flow Channel Inserts and Coatings for High Temperature Liquid Metal Applications","authors":"A. Saraswat, R. Bhattacharyay, P. Chaudhuri, S. Gedupudi","doi":"10.1115/1.4062646","DOIUrl":"https://doi.org/10.1115/1.4062646","url":null,"abstract":"\u0000 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.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":"2 1","pages":""},"PeriodicalIF":0.4,"publicationDate":"2023-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77085142","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}

引用次数: 0