Journal of Nuclear Engineering and Radiation Science最新文献_第4页

Greetings From the Chair of the Jsme Power Energy Systems Division 来自Jsme电力能源系统部主席的问候

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-21 DOI: 10.1115/1.4063882

Takashi Kiga

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Application of Level 3 Probabilistic Safety Analysis In UK HPR1000 三级概率安全分析在英国HPR1000核电站中的应用

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-21 DOI: 10.1115/1.4063874

Jinkai Wang, Qi Wang

{"title":"Application of Level 3 Probabilistic Safety Analysis In UK HPR1000","authors":"Jinkai Wang, Qi Wang","doi":"10.1115/1.4063874","DOIUrl":"https://doi.org/10.1115/1.4063874","url":null,"abstract":"Abstract A Level 3 Probabilistic Safety Assessment (L3 PSA) is required in UK Generic Design Assessment (GDA) to demonstrate that a new nuclear power plant is suitable to be built in UK. L3 PSA is used to assess the individual and societal risk and compare the results against the offsite Radiation Protection Targets (RPTs) for fault and accident conditions. There is little relevant good practice and mature standard for L3 PSAs that have recently been implemented worldwide. In this study, a pilot L3 PSA is performed for UK HPR1000 to reflect the UK context and relevant good practices. It introduces the methodology and the processes to be followed to perform conditional consequences calculations for the faults and accident scenarios. All radiation sources are considered and analyzed. The radiological risks to a potential UK site are analyzed and compared against RPTs. A widely used code - PC COSYMA, is selected for quantification. The strengths and limitations of the code are justified based on the project situation, and either qualitative arguments or supplementary analysis is subsequently proposed to overcome the limitations. The final L3 PSA results are derived to support the demonstration that the offsite radiological risks for UK HPR1000 have been achieved as low as reasonably practicable (ALARP) and has met the UK regulatory expectation.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-10-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135513156","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

Advancement Towards the Experimental Measurement of the Lbe Thermal Properties Using DSC Technique 用DSC技术测定Lbe热性能的实验研究进展

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-20 DOI: 10.1115/1.4063572

Satya Saraswat, Nicola Forgione, Massimo Emilio Angiolini

{"title":"Advancement Towards the Experimental Measurement of the Lbe Thermal Properties Using DSC Technique","authors":"Satya Saraswat, Nicola Forgione, Massimo Emilio Angiolini","doi":"10.1115/1.4063572","DOIUrl":"https://doi.org/10.1115/1.4063572","url":null,"abstract":"Abstract The differential scanning calorimetry (DSC) method has recently emerged as a sophisticated and precise technique for promising contributions to the thermal analysis of various materials, including heavy liquid metal (HLM) coolants. However, there is a lack of experimental studies on the thermal properties of lead-based fluids, such as lead–bismuth eutectic (LBE) and lead–lithium eutectic, which are potential candidates for use as coolants, breeders, and neutron multipliers in advanced nuclear systems like the fourth-generation lead-cooled fast reactor. The available experimental data on the thermal properties of LBE and other lead-based fluids is limited, and the measurements have significant uncertainty. In addition, the composition of components used in the previous studies is inconsistent, and the environmental conditions were often unknown. Therefore, to fill these gaps and advance the thermal properties measurement technique for heavy liquid metal coolants, ENEA Brasimone, in collaboration with DICI-UNIPI, has installed a DSC instrument setup. The experiments performed at the installed DSC setup are focused on measuring some essential thermal properties of LBE using DSC. The experience gained from this work will facilitate the measurement of other fluids based on lead alloy, especially lead–lithium eutectic, a potential candidate for breeder, coolant, and neutron multiplier in demonstration power plant fusion reactors. This study represents the first effort to advance the DSC approach for accurately measuring the thermal characteristics of heavy liquid metals that are highly reactive, such as lead–lithium, which has significant potential in advanced nuclear systems.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-10-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135513941","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

Technical Evaluation Requirements on Welding-Related Nonconformance Reports 焊接不合格报告技术评价要求

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-20 DOI: 10.1115/1.4063630

Chongzhi Wu, Mingkai Li, Zongkui Wang, Zhenguo Peng

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Ned Chair's Message Ned Chair的话

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-19 DOI: 10.1115/1.4063844

Asif Arastu

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Neutronic Analysis and Fuel Cycle Parameters of Transuranic-UO2 Fueled Dual Cooled VVER-1000 Assembly 超铀- uo2燃料双冷VVER-1000组件的中子分析和燃料循环参数

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-14 DOI: 10.1115/1.4063757

Md. Tanvir Ahmed, Afroza Shelley, Md. Nobir Hosen

{"title":"Neutronic Analysis and Fuel Cycle Parameters of Transuranic-UO2 Fueled Dual Cooled VVER-1000 Assembly","authors":"Md. Tanvir Ahmed, Afroza Shelley, Md. Nobir Hosen","doi":"10.1115/1.4063757","DOIUrl":"https://doi.org/10.1115/1.4063757","url":null,"abstract":"Abstract The feasibility of transuranic (TRU) fuel rods has been studied for a dual-cooled VVER-1000 assembly along with the conventional UO2 fuel rods. It has been found that the heterogeneous arrangements of TRU and UO2 fuel rods help to improve the multiplication factor, enhance the fuel cycle parameters and maintain the negative Doppler coefficient of the reference VVER-1000 assembly. Within the different heterogeneous combinations of TRU and UO2 fuel rods, the equal number of UO2 and TRU fuel rods, which is referred to as Model 6 increased the multiplication factor of the reference assembly by 7.66% at the beginning of the cycle. Furthermore, the fuel cycle parameter becomes almost double for this Model in comparison with the reference assembly. TRU rods make the neutron spectrum of reference assembly slightly harder and increase the fast fission rate. However, the Doppler coefficient becomes less negative, and flux level decreases with increasing the number of TRU rods. By considering the safety parameters and neutronic behavior, Model 6 (equal number of UO2 and TRU fuel rods) is to be effective for the annular fueled VVER-1000 reactor.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-10-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135803736","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

Editorial Message to Journal Readers 给期刊读者的社论信息

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-14 DOI: 10.1115/1.4063758

Igor Pioro

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Feasibility Study on the Application of Boron Carbide for Long Term Reactivity Control in the LOTUS Small Fast Reactor 碳化硼用于LOTUS小快堆反应性长期控制的可行性研究

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-13 DOI: 10.1115/1.4063739

Thanh Mai Vu, Thi Hong Bui, Le Quang Linh Tran

{"title":"Feasibility Study on the Application of Boron Carbide for Long Term Reactivity Control in the LOTUS Small Fast Reactor","authors":"Thanh Mai Vu, Thi Hong Bui, Le Quang Linh Tran","doi":"10.1115/1.4063739","DOIUrl":"https://doi.org/10.1115/1.4063739","url":null,"abstract":"Abstract LOTUS reactor core is a small modular lead-cooled fast reactor with designed power of 200 MWth under development at VNU University of Science, Hanoi for a floating nuclear power plant application. For that purpose, advanced passive safety features and no refuelling requirement are the priorities in the core design process. To endure the continuous operation over a long lifetime, the start-up core exhibits excess reactivity to cover the reactivity loss due to burnup. The reactivity control system includes burnable poison and absorber rods and layers made of B4C which are employed in the reactor to minimize the excess reactivity of the core to about 1 $ to enhance the safety features of the core. The burnable poison is fixed inside the reactor while absorber rods/absorber layers were withdrawn or inserted in sequence to achieve the required excess reactivity of about 700 pcm. The reactivity control was arranged into ten steps to achieve the operating time of 15 effective full-power years without refuelling. Good neutronics behaviour of the core was observed with negative fuel temperature coefficient and coolant void reactivity and maximum radial power peaking factor of 1.32. However, a quite large residual absorption caused by fixed burnable poison inside fuel assemblies was revealed. In further study, to increase the neutron absorption efficiency of burnable poison in the fast spectrum as well as the reactor lifetime, a neutron moderator will be considered to add into the burnable poison rods.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-10-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135918087","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

Cold Crucible Facility for Severe Accident Research at CNL 中国科学院重大事故研究用冷坩埚装置

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-07 DOI: 10.1115/1.4063673

Robert David, Blake Mitchell, Justin Spencer, Norm Lair, Sergei Petoukhov

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A Consistent One-Dimensional Multigroup Diffusion Model for Molten Salt Reactor Neutronics Calculations 熔盐堆中子计算的一维一致多群扩散模型

Journal of Nuclear Engineering and Radiation Science Pub Date : 2023-10-06 DOI: 10.3390/jne4040041

Mohamed Elhareef, Zeyun Wu, Massimiliano Fratoni

{"title":"A Consistent One-Dimensional Multigroup Diffusion Model for Molten Salt Reactor Neutronics Calculations","authors":"Mohamed Elhareef, Zeyun Wu, Massimiliano Fratoni","doi":"10.3390/jne4040041","DOIUrl":"https://doi.org/10.3390/jne4040041","url":null,"abstract":"Molten Salt Reactors (MSRs) have recently gained resurged research and development interest in the advanced reactor community. Several computational tools are being developed to capture the strong neutronics/thermal-hydraulics coupling effect in this special reactor configuration. This paper presents a consistent one-dimensional (1D) multigroup neutron diffusion model for MSR analysis, with the primary aim for fast and accurate calculations for long transients, as well as sensitivity and uncertainty analysis of the reactor. A fictitious radial leakage cross section is introduced in the model to properly account for the radial leakage effects of the reactor. The leakage cross section and other consistent neutronics parameters are generated with the Monte Carlo code Serpent using high-fidelity three-dimensional (3D) models. The accuracy of the 1D consistent model is verified by the reference solution from the Monte Carlo model on the Molten Salt Reactor Experiment (MSRE) configuration. The 1D consistent model successfully reproduced the integrated flux from the 3D model and the reactor multiplication factor keff with the error in the range of 95 to 397 pcm (per cent mille), depending on discretized energy group structures. The developed model is also extended to estimate the reactivity loss due to fuel circulation in MSRE. The estimate of reactivity loss in dynamics analysis is in great agreement with the experimental data. This model functions as the first step in the development of a 1D fully neutronics/thermal-hydraulics coupled model for short- and long-term MSRE transient analysis.","PeriodicalId":16756,"journal":{"name":"Journal of Nuclear Engineering and Radiation Science","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2023-10-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"135351347","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