Transactions of the Atomic Energy Society of Japan最新文献_第7页

Measurement of 137Cs Activity with Pinhole Gamma Camera 用针孔伽马照相机测量137Cs活度

Transactions of the Atomic Energy Society of Japan Pub Date : 2020-01-01 DOI: 10.3327/taesj.j19.026

H. Hirayama, Katsumi Hayashi, K. Iwanaga, K. Kondo, Seishiro Suzuki

引用次数: 5

Pressure Resistance Thickness of Disposal Containers for Spent Fuel Direct Disposal 乏燃料直接处置容器的耐压厚度

Transactions of the Atomic Energy Society of Japan Pub Date : 2020-01-01 DOI: 10.3327/taesj.j19.007

Y. Sugita, N. Taniguchi, Hitoshi Makino, Shinichiro Kanamaru, T. Okumura

引用次数: 0

Evaluation of Impact with High Temperature Engineering Test Reactor Using Realistic Model of Stack and Reactor Building 高温工程试验堆的冲击评价——基于堆堆结构的真实模型

Transactions of the Atomic Energy Society of Japan Pub Date : 2020-01-01 DOI: 10.3327/taesj.j19.005

M. Ono, Y. Fujiwara, Tetsuro Matsumoto, K. Iigaki

引用次数: 0

Damage Assessment in Case of Cable Fire in Nuclear Facilities Considering Damage Criteria for Each Material Type 考虑各类材料损伤标准的核设施电缆火灾损伤评估

Transactions of the Atomic Energy Society of Japan Pub Date : 2020-01-01 DOI: 10.3327/taesj.j19.021

Susumu Tsuchino, A. Matsuda

{"title":"Damage Assessment in Case of Cable Fire in Nuclear Facilities Considering Damage Criteria for Each Material Type","authors":"Susumu Tsuchino, A. Matsuda","doi":"10.3327/taesj.j19.021","DOIUrl":"https://doi.org/10.3327/taesj.j19.021","url":null,"abstract":"Various types of ﬂame retardant cable are used in nuclear power plants. The cable insulation and sheath materials are generally categorized into two types: thermosetting （ TS ） and thermoplastic （ TP ） materials. Domestic cable ﬁre damage criteria are set for each material type on the basis of the US guidelines, although they do not clearly deﬁne which type a TS / TP composite cable falls into. In this study, cable burning tests were conducted for TP and TS / TP composite cables used in domestic nuclear power plants to evaluate the validity of applying the US cable damage criteria to domestic cables. For varions cable damage scenarios, the tests measured the temperature / time dependence of insulation resistance between conductors and between cable and trays. The results of these tests justiﬁed the applicability of the US damage criteria. In addition, it was conﬁrmed that the relationship be tween cable temperature and insulation resistance can be expressed by an Arrhenius plot. These ﬁnd ings are expected to be utilized for the Safety Improvement Evaluation periodically conducted by nu clear facility operators.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2020-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437320","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 Fabrication Technology for Oxidation-Resistant Fuel Elements for High-Temperature Gas-Cooled Reactors 高温气冷堆抗氧化燃料元件制造技术的发展

Transactions of the Atomic Energy Society of Japan Pub Date : 2019-03-01 DOI: 10.3327/TAESJ.J17.027

J. Aihara, Masaki Honda, S. Ueta, H. Ogawa, Koichiro Ohira, Y. Tachibana

引用次数: 1

Thermal and Mechanical Properties of Fe2Zr Fe2Zr的热性能和力学性能

Transactions of the Atomic Energy Society of Japan Pub Date : 2019-03-01 DOI: 10.3327/TAESJ.J18.011

Daisuke Okada

引用次数: 4

Wettability of Liquid Cesium Halides on Oxide Single Crystals 液态卤化铯在氧化物单晶上的润湿性

Transactions of the Atomic Energy Society of Japan Pub Date : 2019-03-01 DOI: 10.3327/TAESJ.J18.002

H. Ishii, Y. Murakami, Y. Ohishi, H. Muta, M. Uno, K. Kurosaki

{"title":"Wettability of Liquid Cesium Halides on Oxide Single Crystals","authors":"H. Ishii, Y. Murakami, Y. Ohishi, H. Muta, M. Uno, K. Kurosaki","doi":"10.3327/TAESJ.J18.002","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.002","url":null,"abstract":"The wettability of liquid-state volatile ﬁssion products （ FPs ） for solid-state fuels is important for the evaluation of FP release behavior during severe accidents. Previously, we have reported that liquid CsI exhibits extremely high wettability with a contact angle of virtually 0 ° on a polycrystalline UO 2 surface ［ K. Kurosaki et al., Sci. Rep. 7, Article number: 11449 （ 2017 ）］ and a single-crystalline yttria-stabilized zirconia （ YSZ ） surface ［ H. Ishii et al., J. Nucl. Sci. Technol. 55, No. 8, 838 – 842 （ 2018 ）］ . Here, based on the previous studies, we investigate the wettability of liquid cesium halides （ CsCl, CsBr ） on single-crystalline YSZ, TiO 2 , and MgO. We observe the high wettability of liquid cesium halides on single-crystalline YSZ and TiO 2 with contact angles of nearly 0 ° in all cases. However, liquid cesium halides exhibit completely different wettability on single-crystalline MgO, where the contact angles are measured to be 50 ° , 44 ° , and 25 ° for CsCl, CsBr, and CsI, respectively. Similarly to UO 2 , YSZ and TiO 2 are non-stoichiometric compounds containing oxygen defects, while MgO is a stoichiometric line compound. Thus, it is assumed that the oxygen defects play a role in the high wettability of liquid CsI on solid UO 2 .","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":" ","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"47029184","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

Examination Method for Flow Fields in Piping Systems of Nuclear Power Plants Using Computational Fluid Dynamics 核电站管道系统流场的计算流体力学检测方法

Transactions of the Atomic Energy Society of Japan Pub Date : 2019-03-01 DOI: 10.3327/TAESJ.J18.004

K. Yoshimura, T. Sugii, T. Sano, E. Ishii, Takumi Kitagawa

{"title":"Examination Method for Flow Fields in Piping Systems of Nuclear Power Plants Using Computational Fluid Dynamics","authors":"K. Yoshimura, T. Sugii, T. Sano, E. Ishii, Takumi Kitagawa","doi":"10.3327/TAESJ.J18.004","DOIUrl":"https://doi.org/10.3327/TAESJ.J18.004","url":null,"abstract":"We developed an examination method that uses computational fluid dynamics (CFD) to investigate the effects of a complex pipe geometry on flow fields. Two kinds of pipe model with different geometries are simulated to test the developed method. The simulation models were split into several computational regions to reduce the computation time. The simulation results showed that the fluctuation of the flow rate depended on the pipe geometry, which qualitatively agreed well with the experimental results. The simulation results of one of the two models showed a swirling flow around the orifice with large fluctuations of the flow rate. It was found that the swirling flow caused velocity fluctuations in the recirculation zone around the tap positions, which resulted in the large fluctuations of the flow rate. We also investigated the mechanisms generating the swirling flow. The simulation results showed that the high velocity of the flow along the wall was caused by the valve and the bend pipe. The high-velocity flow then moves along the pipe wall of the tee, which causes the flow to swirl. These results show that the developed method can be used to evaluate the flow fields in piping systems.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"42100306","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

Study on Practical Application of Optimization Method for Radioactive Waste Disposal Facility Design Based on Probabilistic Approach: Discussion on Procedure for Setting Radionuclide Migration Parameters 基于概率法的放射性废物处理设施设计优化方法的实际应用研究——关于放射性核素迁移参数设置程序的探讨

Transactions of the Atomic Energy Society of Japan Pub Date : 2019-03-01 DOI: 10.3327/TAESJ.J18.003

R. Nakabayashi, Yasutaka Watanabe, D. Minato, D. Sugiyama

引用次数: 1

Generation of MCNP Whole-Core Reference Solutions for Advanced BWR MOX Core and Its Application to Neutronics Calculation Performance V&V of the BWR Nuclear Design Code 先进沸水堆MOX堆核MCNP全芯参考解的生成及其在沸水堆核设计规范中子计算性能V&V中的应用

Transactions of the Atomic Energy Society of Japan Pub Date : 2019-01-01 DOI: 10.3327/TAESJ.J17.015

Tadashi Ikehara, M. Sasagawa, Sho Takano, Teppei Yamana, Naoki Yanagisawa

{"title":"Generation of MCNP Whole-Core Reference Solutions for Advanced BWR MOX Core and Its Application to Neutronics Calculation Performance V&V of the BWR Nuclear Design Code","authors":"Tadashi Ikehara, M. Sasagawa, Sho Takano, Teppei Yamana, Naoki Yanagisawa","doi":"10.3327/TAESJ.J17.015","DOIUrl":"https://doi.org/10.3327/TAESJ.J17.015","url":null,"abstract":"MCNP whole-core reference solutions with pin-by-pin resolution were generated to numerically complement the operating and/or experimentally measured data of an Advanced BWR MOX core loaded with UO2 to 100% MOX fuels for the verification and validation (V&V) of BWR core design codes. To this end, (1) the degree of preciseness in geometrical and material modeling under the allowable storage limitations of MCNP was explored, (2) the correctness of predicted neutronics phenomena inside BWR cores by MCNP physics was investigated and the necessary enhancement was made for MCNP, (3) the smallness of uncertainties of results in MCNP due to its stochastic treatment and the nuclear data employed was ensured, (4) the representativeness of BWR core characteristics obtained from MCNP reference solutions was qualified. According to the results, the MCNP-generated reference solutions are applicable to validating the neutronics calculation performance of BWR core design codes as an alternative to measured data. In addition, extraction of the detailed neutronics quantities such as fuel-pin-wise flux enables us to verify the physics modules of the design codes such as the neutron flux solver. The MCNP whole-core reference solutions proved to be applicable in performing MCNP-based BWR design code V&V.","PeriodicalId":55893,"journal":{"name":"Transactions of the Atomic Energy Society of Japan","volume":"1 1","pages":""},"PeriodicalIF":0.0,"publicationDate":"2019-01-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"69437173","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