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variations by nonuniform core configurationsGunji, Satoshi; Araki, Shohei; Suyama, Kenya
Nuclear Science and Engineering, 197(8), p.2017 - 2029, 2023/08
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)The fuel debris generated by the accident at the Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Plant is expected to have not only heterogeneous but also nonuniform compositions. Similarly, damaged fuel assemblies remaining in the reactor vessels also have nonuniform configurations due to some missing fuel rods. This non-uniformity may cause changing neutron multiplication factors. The effect of non-uniformity on the neutron multiplication factor is clarified by computations, and the possibility of experimentally validating the computations used for criticality management is being investigated. For this purpose, in this study the criticality effects of several core configurations of a new critical assembly, STACY, of the Japan Atomic Energy Agency with nonuniform arrangements of uranium oxide fuel rods, concrete rods, and stainless-steel rods were studied to confirm benchmarking potential. The difference in these arrangements changed the neutron multiplication factor by more than 1 $. We confirmed that changes in local neutron moderation conditions and the clustering of specific components caused this effect. In addition, the feasibility of benchmark experimental cores with nonuniform arrangements is evaluated. If benchmarking of such experiments could be realized, it would help to validate calculation codes and to develop criticality management methods by machine learning.
Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Removal of core decay heat by utilizing natural circulation is expected as a significant measure to enhance the safety of sodium-cooled fast reactors (SFRs). Accurate evaluation of the temperature distribution in the fuel assembly (FA) at the low Re regime in natural circulation operation is demanded. A detailed thermal-hydraulics analysis code named SPIRAL has been developed to clarify thermal-hydraulic phenomena in the FA at various operation conditions. In this study, SPIRAL with the hybrid turbulence model was applied to analyze a large-scale fuel assembly experiment of a 91-pin bundle for two cases at the mixed and the natural convection conditions respectively in low Re regime with heat transfer from outside of the FA. The applicability of the SPIRAL to the thermal-hydraulics evaluation of FA at mixed and natural convection conditions was confirmed by the comparisons of temperatures predicted by SPIRAL with those measured in the experiment.
Hiyama, Tomoyuki; Aizawa, Kosuke; Nishimura, Masahiro; Kurihara, Akikazu
JAEA-Research 2021-009, 29 Pages, 2021/11
JAEA-Research-2021-009.pdf:2.25MB
In sodium-cooled fast reactors, high burnup of fuel is required for practical use. It is important to predict and evaluate the flow behavior in a fuel assembly because there is a concern that the heat removal capacity of the fuel assembly with high burnup will be locally reduced due to swirling and thermal deformation of the fuel rods. In this study, flow field measurement tests were conducted using a 3-pin bundle system test specimen for the purpose of elucidating the phenomenon and constructing a verification database for thermal hydraulics analysis code. The viewpoints of the experiment for elucidating the phenomenon are as follows; (1) Overall flow behavior in the subchannel including near the wrapping wire, (2) Relationship between Reynolds number including laminar flow region and flow field, and (3) Evaluation of the effect of the presence or absence of wrapping wire on the flow field. As a result, detailed flow field data in the subchannel was obtained by PIV measurement. It was found that when the wrapping wire crossed the subchannel, the flow occurred toward adjacent subchannel and the flow occurred that follows the winding direction of the wrapping wire. It was confirmed that the tendency of the flow velocity distribution of the Reynolds number in the laminar flow region is significantly different from that of the transition region and the turbulent region under the condition. The test was conducted using a same 3-pin bundle system without the wrapping wire, and it was confirmed that mixing by the wrapping wire occurred even in the laminar flow region.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 8 Pages, 2021/08
In the design study of an advanced sodium-cooled fast reactor (Advanced-SFR) in JAEA, the use of a specific fuel assembly (FA) with an inner duct structure called FAIDUS has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the temperature distribution to confirm feasibility of FAIDUS. For the FAIDUS, confirmation of validity of the numerical results by a subchannel analysis code named ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mock-up experiment, yet. Therefore, the code-to-code comparisons with numerical results of ASFRE and those of a CFD code named SPIRAL was conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism of the specific temperature and velocity distributions appearing around the inner duct between the numerical results by ASFRE and those by SPIRAL.
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.
Kikuchi, Norihiro; Doda, Norihiro; Hashimoto, Akihiko*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2017 Koen Rombunshu (CD-ROM), 4 Pages, 2017/08
A specific fuel assembly named FAIDUS (Fuel Assembly with Inner Duct Structure) has been developed as one of the measures to enhance safety of the reactor in the core disruptive accident (CDA) in JAEA. Thermal-hydraulics evaluations in FAIDUS under various operation conditions including the CDA are required to confirm its design feasibility. Therefore, numerical simulations by using thermal-hydraulics analysis program named SPIRAL developed in JAEA are conducted to analyze the thermal-hydraulics in the FAIDUS. Through the numerical simulation in the FAIDUS under tentative rated operation condition of an Advanced SFR, it was indicated that the flow and temperature distribution in the FAIDUS showed the same tendency as that in ordinary FA and specific characteristics was not observed.
Kikuchi, Norihiro; Ohshima, Hiroyuki; Imai, Yasutomo*; Hiyama, Tomoyuki; Nishimura, Masahiro; Tanaka, Masaaki
Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2015 Koen Rombunshu, p.179 - 180, 2015/08
In an economically improved sodium-cooled fast reactor, a narrower gap is considered among the fuel pins so as to achieve a high burn-up. Therefore, it is needed to evaluate thermal-hydraulic characteristics in case of a change of the gap geometry due to deformation of fuel pin caused by such as a swelling and a thermal bowing. For this purpose, a FEM analysis code, SPIRAL has been being developed in JAEA and the code validations using water or sodium experimental results have also being performed. In this study, a numerical analysis of a flow field around wire-wrapped fuel pins based on a 3-pin bundle water experiment was carried out as a validation study of SPIRAL. As a result, it was demonstrated that the hybrid-type turbulent model incorporated in SPIRAL has a high applicability to investigate the flow structure of the narrow gap in the fuel assembly.
Department of Hot Laboratories
JAERI-Review 2005-047, 95 Pages, 2005/09
JAERI-Review-2005-047.pdf:6.27MB
This is an annual report in 2004 fiscal year that describes activities of the Reactor Fuel Examination Facility (RFEF), the Waste Safety Testing Facility (WASTEF), and the Research Hot Laboratory (RHL) in the Department of Hot laboratories. In RFEF, BWR fuel rods were withdrawn from a fuel assembly irradiated for 5 cycles in the Fukushima-2 Nuclear Power Station Unit-1 and PIEs including nondestructive examination of those rods were carried out. In WASTEF, Slow Strain Rate Tests for detecting the susceptibility to IASCC, the corrosion test of reprocessing plant materials, tests for evaluating barrier performance in terms of waste disposal were performed. A secondary system pipe from the Mihama Nuclear Power Station Unit-3 was accepted to inspect the ageing fracture of it. In RHL, 15 lead cells are dismantled under the decommissioning plan at JAERI Tokai. And an arrangement of the RHL facility was started to use the storage of unirradiated nuclear materials.
Takase, Kazuyuki; Ose, Yasuo*; Yoshida, Hiroyuki; Akimoto, Hajime; Aoki, Takayuki*
Dai-24-Kai Nihon Shimyureshon Gakkai Taikai Happyo Rombunshu, p.161 - 164, 2005/07
no abstracts in English
Department of Hot Laboratories
JAERI-Review 2005-003, 105 Pages, 2005/02
JAERI-Review-2005-003.pdf:21.21MB
no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*
Hai Pafomansu Komputingu To Keisan Kagaku Shimpojium (HPCS 2005) Rombunshu, P. 16, 2005/01
no abstracts in English
Ose, Yasuo*; Takase, Kazuyuki; Yoshida, Hiroyuki; Kano, Takuma; Akimoto, Hajime
Dai-18-Kai Suchi Ryutai Rikigaku Shimpojiumu Koen Yoshishu (CD-ROM), 6 Pages, 2004/12
no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*
Dai-23-Kai Nihon Shimyureshon Gakkai Taikai Happyo Rombunshu, p.121 - 124, 2004/06
no abstracts in English
Ose, Yasuo*; Takase, Kazuyuki; Yoshida, Hiroyuki; Kano, Takuma; Kureta, Masatoshi; Akimoto, Hajime
Dai-41-Kai Nihon Dennetsu Shimpojiumu Koen Rombunshu, 2 Pages, 2004/05
no abstracts in English
Takase, Kazuyuki; Yoshida, Hiroyuki; Ose, Yasuo*; Tamai, Hidesada; Akimoto, Hajime
Transactions of the American Nuclear Society, 89, p.88 - 89, 2003/11
no abstracts in English
Yoshida, Hiroyuki; Takase, Kazuyuki; Ose, Yasuo*; Tamai, Hidesada; Akimoto, Hajime
Proceedings of International Conference on Global Environment and Advanced Nuclear Power Plants (GENES4/ANP 2003) (CD-ROM), 8 Pages, 2003/09
no abstracts in English
Takase, Kazuyuki; Masuko, Kenji*; Ose, Yasuo*; Tamai, Hidesada; Kume, Etsuo
Kashika Joho Gakkai-Shi, 23(Suppl.1), p.363 - 364, 2003/07
no abstracts in English
Department of Hot Laboratories
JAERI-Review 2002-039, 106 Pages, 2003/01
JAERI-Review-2002-039.pdf:9.46MB
no abstracts in English
JNC TN1400 2000-012, 250 Pages, 2000/11
JNC-TN1400-2000-012.pdf:10.18MB
no abstracts in English
