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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.
Onoe, Hironori; Kosaka, Hiroshi*; Matsuoka, Toshiyuki; Komatsu, Tetsuya; Takeuchi, Ryuji; Iwatsuki, Teruki; Yasue, Kenichi
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 26(1), p.3 - 14, 2019/06
In this study, it is focused on topographic changes due to uplift and denudation, also climate perturbations, a method which is able to assess the long-term variability of groundwater flow conditions using the coefficient variation based on some steady-state groundwater flow simulation results was developed. Spatial distribution of long residence time area which is not much influenced due to long-term topographic change and recharge rate change during the past one million years was able to estimate through the case study of the Tono area, Central Japan. By applying this evaluation method, it is possible to identify the local area that has low variability of groundwater flow conditions due to topographic changes and climate perturbations from the regional area quantitatively and spatially.
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).
Sakai, Akihiro; Kurosawa, Ryohei*; Totsuka, Masayoshi; Nakata, Hisakazu; Amazawa, Hiroya
JAEA-Technology 2016-032, 117 Pages, 2017/02
JAEA-Technology-2016-032.pdf:12.84MB
JAEA has been planning to implement near surface disposal of low level waste generated from research, medical, and industrial facilities. JAEA plans to carry out 3d analysis of groundwater flow in geological model around the disposal site because of development of migration assessment modeling of radioactivity materials in the site. In the safety demonstration test in JAEA, 3d analysis of groundwater flow was carried out on 1999. The analysis was calculated by using the code "3D-SEEP". But it is necessary to improve the conditions of the model in the analysis. Therefore, we improved the geological model which had been developed carried out 3d analysis of groundwater flow by using the current 3D-SEEP for the specified disposal site in the future. From the result, we expect that 3d analysis of groundwater flow in the environment around the specified near surface disposal site will be able to be sufficiently conducted by developing an appropriate model for the disposal site.
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.
Onoe, Hironori; Saegusa, Hiromitsu; Takeuchi, Ryuji
Doboku Gakkai Rombunshu, C (Chiken Kogaku) (Internet), 72(1), p.13 - 26, 2016/01
The Japan Atomic Energy Agency is conducting the Mizunami Underground Research Laboratory (URL) project in Mizunami, Gifu, in order to establish scientific and technical basis for geological disposal of high-level radioactive waste. This paper comprehensively describes the result of groundwater flow modeling using data of hydraulic responses and hydrochemical changes due to URL construction. Technical know-how and methodology of hydrogeological monitoring and groundwater flow modeling were presented for characterization of hydraulic heterogeneities in fractured crystalline rock. Furthermore, effectivity of data acquisition of hydrochemical changes in groundwater for validation of result of groundwater flow modeling was indicated.
Saegusa, Hiromitsu; Onoe, Hironori; Ishibashi, Masayuki; Tanaka, Tatsuya*; Abumi, Kensho*; Hashimoto, Shuji*; Bruines, P.*
JAEA-Research 2015-011, 59 Pages, 2015/10
JAEA-Research-2015-011.pdf:49.44MB
It is important to evaluate groundwater flow characteristics on several spatial scales for assessment of long-term safety on geological disposal of high-level radioactive wastes. An estimation of hydraulic heterogeneity caused by fracture network is significant for evaluation of the groundwater flow characteristics in the region of tens of meters square. Heterogeneity of equivalent hydraulic properties is needed to estimate for evaluation of the groundwater flow characteristics in the region of several km square. In order to develop the methodology for multi-scale hydrogeological modeling taking into account the hydraulic heterogeneity, spatial distribution of fractures and their hydraulic properties have been modeled using discrete fracture network (DFN) model. Then, hydrogeological continuum model taking into account the hydraulic heterogeneity has been estimated based on the DFN model. Through this study, the methodology for multi-scale hydrogeological modeling according to type of investigation data has been proposed.
Onoe, Hironori; Kosaka, Hiroshi*; Takeuchi, Ryuji; Saegusa, Hiromitsu
JAEA-Research 2015-008, 146 Pages, 2015/08
JAEA-Research-2015-008.pdf:76.46MB
Mizunami Underground Research Laboratory (MIU) Project is being carried out by Japan Atomic Energy Agency (JAEA) in the Cretaceous Toki granite in the Tono area, central Japan. The MIU Project has three overlapping phases: Surface-based Investigation (Phase I), Construction (Phase II) and Operation (Phase III). In this study, calibration of hydrogeological model and groundwater flow simulation using the data obtained by the Phase I and Phase II were carried out in order to develop the methodology for construction and update of hydrogeological model on Site Scale. As a result, hydrogeological model on Site Scale, which is able to simulate comprehensively the obtained data regarding groundwater pressure distribution before excavation of the MIU facilities, hydraulic responses and inflow volume during excavation of the MIU facilities, was constructed.
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.
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
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
*
JNC TN1400 2001-013, 70 Pages, 2001/08
JNC-TN1400-2001-013.pdf:5.13MB
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
Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko*; Hino, Ryutaro
JAERI-Conf 2001-002, p.893 - 903, 2001/03
no abstracts in English
Kaminaga, Masanori; Terada, Atsuhiko*; Haga, Katsuhiro; Kinoshita, Hidetaka; Ishikura, Shuichi*; Hino, Ryutaro
JAERI-Tech 2000-076, 70 Pages, 2001/01
JAERI-Tech-2000-076.pdf:4.01MB
no abstracts in English
Research Group for Advanced Reactor System; Research Group for Reactor Physics; Research Group for Thermal and Fluid Engineering
JAERI-Research 2000-035, 316 Pages, 2000/09
JAERI-Research-2000-035.pdf:19.81MB
no abstracts in English
*; Saegusa, Hiromitsu
JNC TY7400 2000-004, 62 Pages, 2000/05
JNC-TY7400-2000-004.pdf:1.45MB
None
Koshizuka, Seiichi*; *; Okano, Yasushi; *; Yamaguchi, Akira
JNC TY9400 2000-012, 91 Pages, 2000/03
JNC-TY9400-2000-012.pdf:2.82MB
no abstracts in English
