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Tanaka, Masaaki; Uchibori, Akihiro; Okano, Yasushi; Yokoyama, Kenji; Uwaba, Tomoyuki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai
Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09
The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. This introductory paper also provides an overview of an integrated evaluation system named ARKADIA to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.
Kikuchi, Norihiro; Mori, Takero; Okajima, Satoshi; Tanaka, Masaaki; Miyazaki, Masashi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
In JAEA, the design optimization method for plant structure has been developed on the process to output optimal solution of the thickness of reactor vessel wall against thermal transient and seismic loads in a SFR as a representative problem. Resistance characteristic of the wall on the load derived from thermal transient is one of the most important factors for safety estimation on the structural integrity. Failure probability of components against thermal transient was set to one of variables in the objective function for a common scale to compare with other variables in different failure mechanisms. In the iterative process to achieve the optimal solution, a number of evaluations to measure the influence on the load derived from thermal transient was necessarily conducted. More reduction of required time for evaluations is desired. To perform the iterative evaluation process efficiently, the automatization of parametric analyses was implemented in the optimization process.
Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*
Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.73 - 80, 2020/10
A finite element thermal-hydraulics simulation code SPIRAL has been developed in Japan Atomic Energy Agency (JAEA) to analyze the detailed thermal-hydraulics phenomena in a fuel assembly (FA) of Sodium-cooled Fast Reactors (SFRs). The numerical simulation of a large-scale sodium test for 91-pin bundle (GR91) at low flow rate condition was performed for the validation of SPIRAL with the hybrid k-e turbulence model to take into account the low Re number effect near the wall in the flow and temperature fields. Through the numerical simulation, specific velocity distribution affected by the buoyancy force was shown on the top of the heated region and the temperature distribution predicted by SPIRAL agreed with that measured in the experiment and the applicability of the SPIRAL to thermal-hydraulic evaluation of large-scale fuel assembly at low flow rate condition was indicated.
Aoki, Takeshi; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
The flow distribution analysis, which is a part of thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR) considering unintended flows between graphite blocks, has been performed for steady and conservative conditions. On the other hand, the transient analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR. The present study aims to develop the transient flow distribution analysis code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated for analysis of the unintended flows in the core and the molecular diffusion dominant in beginning air ingress behavior in an air ingress accident.
Aoki, Takeshi; Isaka, Kazuyoshi; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
The flow distribution analysis performed in the HTGR design has to take into account the interaction thermal and radiation deformations of the graphite structure, and the gaps between the graphite structures forming unintended flow. In the present study, a user-friendly flow network calculation code (FNCC) has been developed on the basis of experiences of High Temperature engineering Test Reactor (HTTR) design for HTGR design with enhanced compatibility with other HTGR design codes and with considering graphite block deformation in iteration process without manual control. The validation of FNCC was performed for the one-column flow distribution test. The analytical results using FNCC showed good agreement with the experimental results. It is concluded that FNCC was validate for the analysis of distributions of flowrate and pressure for the flow network model including the unintended flow paths in prismatic-type HTGRs.
Muramatsu, Toshiharu; Sato, Yuji; Kamei, Naomitsu; Aoyagi, Yuji*; Shobu, Takahisa
Nihon Kikai Gakkai Dai-13-Kai Seisan Kako, Kosaku Kikai Bumon Koenkai Koen Rombunshu (No.19-307) (Internet), p.157 - 160, 2019/10
no abstracts in English
Nakamura, Hideo
Nihon Genshiryoku Gakkai-Shi ATOMO
, 61(4), p.270 - 272, 2019/04
no abstracts in English
Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08
Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish an appropriate numerical models for the direct heat exchanger (DHX).
Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00431_1 - 16-00431_11, 2017/04
A plant dynamics analysis code Super-COPD is being developed in JAEA for the design and safety assessments of sodium-cooled fast reactors (SFRs). In this study, the friction loss coefficients in the whole core thermal-hydraulic model was modified to improve the prediction accuracy of the sodium temperature distribution in a fuel subassembly under the natural circulation conditions. The modified whole core model was applied to analyses of experiments that were performed by using JAEA's test facility PLANDTL as a part of the code validation study. The obtained numerical results of sodium temperature distributions in the core showed good agreement with the measured data. It implies that the modified whole core model can properly reproduce dominant thermal-hydraulic phenomena in the core region under natural circulation conditions, i.e., flow redistribution among fuel subassemblies as well as in a fuel subassembly and inter-subassembly heat transfer.
Kaji, Yoshiyuki; Katsuyama, Jinya; Yamaguchi, Yoshihito; Nemoto, Yoshiyuki; Osaka, Masahiko
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
To investigate the inhomogeneous temperature and stress distribution by geometrical complex of BWR lower head, the detailed 3D model of RPV lower head with control rod guide tubes and shroud supports are constructed and the 3D thermal hydraulic analysis of simulated molten pool and creep deformation analysis of lower head are performed using ANSYS Fluent / Mechanical finite element code. It is found that failure for BWR lower head might be caused by combination between melting failure in inner surface of lower head and creep failure in outer surface of lower head.
Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki
Nuclear Engineering and Design, 312, p.30 - 41, 2017/02
Times Cited Count：6 Percentile：51.46(Nuclear Science & Technology)In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related Research and Development results on innovative technologies and lessons learned from Fukushima Dai-ichi Nuclear Power Plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V and V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.
Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki
Nuclear Engineering and Design, 299, p.174 - 183, 2016/04
Times Cited Count：4 Percentile：36.53(Nuclear Science & Technology)A procedure called as V2UP (Verification and Validation plus Uncertainty quantification and Prediction) was made by referring to the existing guidelines on V&V and the methodologies of the safety assessment (CSAU, ISTIR, EMDAP). The V2UP consisted of five components as follows: (1) phenomena analysis with the Phenomena Identification and Ranking Table (PIRT) method, (2) implementation of the V&V, (3) design and rearrangement of experiments for the V&V, (4) uncertainty quantification in each problem and integration of uncertainties and (5) numerical prediction (estimation) for the target issue. Although the complete application of the procedure has not been performed at this moment, a flow chart of the V2UP procedure was described in this paper with recent results of the examinations.
Sato, Hiroshi; Aso, Tomokazu; Kogawa, Hiroyuki; Teshigawara, Makoto; Hino, Ryutaro
JAERI-Tech 2004-018, 23 Pages, 2004/03
JAERI-Tech-2004-018.pdf:2.42MB
The Japan Atomic Energy Research Institute is constructing a mega-watt class spallation neutron source in cooperation with the High Energy Accelerator Research Organization. A cold moderator using liquid hydrogen is one of the key components in the system, which directly affects the neutronic performance both in intensity and pulse time structure. Since a hydrogen temperature rise in the moderator vessel affects the neutronic performance, it is necessary to suppress the recirculation and stagnant regions which would cause hot spots. A cold moderator with a poison plate (poisoned decoupled moderator) has a high possibility to generate the stagnant region on and near the poison plate. Thermal-hydraulic analyses were carried out with proposed inner structure of the poisoned cold moderator. The stagnant and recirculation regions could be reduced by making a gap between the poison plate end and the vessel bottom surface, and the local temperature rise also could be kept under the required design value.
Takase, Kazuyuki
Konsoryu, 17(3), p.267 - 276, 2003/09
no abstracts in English
Ida, Mizuho*; Nakamura, Hideo; Nakamura, Hiroshi*; Nakamura, Hiroo; Ezato, Koichiro; Takeuchi, Hiroshi
Fusion Engineering and Design, 63-64, p.333 - 342, 2002/12
Times Cited Count：43 Percentile：91.34(Nuclear Science & Technology)no abstracts in English
Ose, Yasuo*; Takase, Kazuyuki; Yoshida, Hiroyuki; Akimoto, Hajime
Proceedings of 10th International Conference on Nuclear Engineering (ICONE 10) (CD-ROM), 8 Pages, 2002/00
no abstracts in English
Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro
Nihon Genshiryoku Gakkai-Shi, 43(11), p.1149 - 1158, 2001/11
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)no abstracts in English
Takase, Kazuyuki
Proceedings of the 1st International Symposium on Advanced Fluid Information (AFI-2001), p.376 - 381, 2001/10
no abstracts in English
Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro
JAERI-Tech 2001-051, 22 Pages, 2001/08
JAERI-Tech-2001-051.pdf:4.51MB
no abstracts in English
Takase, Kazuyuki
Fusion Engineering and Design, 54(3-4), p.605 - 615, 2001/04
Times Cited Count：12 Percentile：64.73(Nuclear Science & Technology)no abstracts in English
