Ina, Takuya*; Idomura, Yasuhiro; Imamura, Toshiyuki*; Yamashita, Susumu; Onodera, Naoyuki
Proceedings of 12th Workshop on Latest Advances in Scalable Algorithms for Large-Scale Systems ScalA21) (Internet) , 8 Pages, 2021/11
A new mixed-precision preconditioner based on the iterative refinement (IR) method is developed for preconditioned conjugate gradient (P-CG) and multigrid preconditioned conjugate gradient (MGCG) solvers in a multi-phase thermal-hydraulic CFD code JUPITER. In the IR preconditioner, all data is stored in FP16 to reduce memory access, while all computation is performed in FP32. The hybrid FP16/32 implementation keeps the similar convergence property as FP32, while the computational performance is close to FP16. The developed solvers are optimized on Fugaku (A64FX), and applied to ill-conditioned matrices in JUPITER. The P-CG and MGCG solvers with the new IR preconditioner show excellent strong scaling up to 8,000 nodes, and at 8,000 nodes, they are respectively accelerated up to 4.86 and 2.39 from the conventional ones on Oakforest-PACS (KNL).
Yoshida, Kazuo; Tamaki, Hitoshi; Hiyama, Mina*
JAEA-Research 2021-005, 25 Pages, 2021/08
An accident of evaporation to dryness by boiling of high level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru) are released from the tanks with water and nitric-acid mixed vapor into atmosphere. Accurate quantitative estimation of released Ru is one of the important issues for risk assessment of those facilities. To resolve this issue, an empirical correlation equation of Ru mass transfer coefficient across the vapor-liquid surface, which can be useful for quantitative simulation of Ru mitigating behavior, has been obtained from data analyses of small-scale experiments conducted to clarify gaseous Ru migrating behavior under steam-condensing condition. A simulation study has been also carried out with a hypothetical typical facility building successfully to demonstrate the feasibility of quantitative estimation of amount of Ru migrating in the facility using the obtained correlation equation implemented in SCHERN computer code which simulates chemical behaviors of nitrogen oxide based on the condition also simulated thermal-hydraulic computer code.
Maruyama, Yu; Yoshida, Kazuo
Nihon Genshiryoku Gakkai-Shi ATOMO, 63(7), p.517 - 522, 2021/07
no abstracts in English
Wang, Z.; Duan, G.*; Koshizuka, Seiichi*; Yamaji, Akifumi*
Nuclear Power Plant Design and Analysis Codes, p.439 - 461, 2021/00
Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki
Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06
Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.
Wan, T.; Obayashi, Hironari; Sasa, Toshinobu
Nuclear Technology, 205(1-2), p.188 - 199, 2019/01
Itoi, Tatsuya*; Iwaki, Chikako*; Onuki, Akira*; Kito, Kazuaki*; Nakamura, Hideo; Nishida, Akemi; Nishi, Yoshihisa*
Nihon Genshiryoku Gakkai-Shi ATOMO, 60(4), p.221 - 225, 2018/04
no abstracts in English
Proceedings of ANS International Conference on Best Estimate Plus Uncertainties Methods (BEPU 2018) (USB Flash Drive), 8 Pages, 2018/00
no abstracts in English
Wan, T.; Obayashi, Hironari; Sasa, Toshinobu
Proceedings of 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-17) (USB Flash Drive), 13 Pages, 2017/09
Takata, Takashi; Ohno, Shuji; Tajima, Yuji*
Mechanical Engineering Journal (Internet), 4(3), p.16-00577_1 - 16-00577_11, 2017/06
Evaluation of accidental sodium leak, combustion, and its thermal consequence is one of the important issues to be assessed in the field of sodium-cooled fast reactor (SFR). The present paper deals with the sodium pool fire and subsequent heat transfer behavior in air atmosphere two-cell geometry both experimentally and analytically because such two-cell configuration is considered as a typical one to possess important characteristic of multi-compartment system seen in an actual plant. As a result of the numerical analysis using a lumped-parameter based zonal model safety analysis code SPHINCS, the applicability of the ventilation model implemented in SPHINCS has been demonstrated. It is also investigated that the buoyancy- driven ventilation is dominant in the experiment.
Inaba, Yoshitomo; Isaka, Kazuyoshi; Shibata, Taiju
JAEA-Data/Code 2017-002, 74 Pages, 2017/03
In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as core geometry and specifications, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code, named FTCC, which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This report describes the calculation objects and models, the basic equations, the strong points (improvement points from the HTTR design code), the code structure, the using method of FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for high temperature gas-cooled reactors. In addition, the effect of hot spot factors and fuel cooling forms on reducing the maximum fuel temperature is investigated with FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
Inaba, Yoshitomo; Nishihara, Tetsuo
Annals of Nuclear Energy, 101, p.383 - 389, 2017/03
In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as thermal power, core geometry, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code named FTCC which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This paper describes calculation objects and models, basic equations, improvement points from the HTTR design code in FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for HTGRs. In addition, the effect of cooling forms on the maximum fuel temperature is investigated by using FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
Akimoto, Hajime; Sugawara, Takanori
JAEA-Data/Code 2016-008, 87 Pages, 2016/09
Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.
Tochio, Daisuke; Fujimoto, Nozomu
Journal of Nuclear Science and Technology, 53(3), p.425 - 431, 2016/03
The future HTGR is now designed in JAEA. The reactor has many merging points of helium gas with different temperature. It is needed to clear the mixing characteristics of helium gas at the pipe in the HTGR from the viewpoint of structure integrity and temperature control. Previously, the reactor inlet coolant temperature was controlled lower than specific one in the HTTR due to lack of mixing of helium gas in the primary cooling system. Now the control system is improved to use the calculated bulk temperature of reactor inlet helium gas. In this paper, thermal-hydraulic analysis on the primary cooling system of the HTTR was conducted to clarify the mixing behavior of helium gas. As the result, it was confirmed that the mixing behavior of helium gas in the primary cooling system is mainly affected by the aspect ratio of annular flow path, and it is needed to consider the mixing characteristics of helium gas at the piping design of the HTGR.
Arai, Kenji*; Umezawa, Shigemitsu*; Oikawa, Hirohide*; Onuki, Akira*; Nakamura, Hideo; Nishi, Yoshihisa*; Fujii, Tadashi*
Nihon Genshiryoku Gakkai-Shi ATOMO, 58(3), p.161 - 166, 2016/03
no abstracts in English
Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 11 Pages, 2015/11
Yoshida, Hiroyuki; Uesawa, Shinichiro; Nagatake, Taku; Jiao, L.; Liu, W.; Takase, Kazuyuki
Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 9 Pages, 2015/11
Watanabe, Osamu*; Oyama, Kazuhiro*; Endo, Junji*; Doda, Norihiro; Ono, Ayako; Kamide, Hideki; Murakami, Takahiro*; Eguchi, Yuzuru*
Journal of Nuclear Science and Technology, 52(9), p.1102 - 1121, 2015/09
A natural circulation (NC) evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500MW adopting the NC decay heat removal system (DHRS). The methodology consists of a 1D safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a 3D fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method. The safety analysis method and the 3D analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a 1/7 scaled sodium test simulating the primary system and the DHRS, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the 3D analysis. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.
Nakamura, Hideo; Arai, Kenji*; Oikawa, Hirohide*; Fujii, Tadashi*; Umezawa, Shigemitsu*; Abe, Yutaka*; Sugimoto, Jun*; Koshizuka, Seiichi*; Yamaguchi, Akira*
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5353 - 5366, 2015/08
Journal of Thermal Science, 24(3), p.295 - 301, 2015/06
Before rise-to-power tests, the actual measured value of heat released from the Reactor Pressure Vessel (RPV) or removed by the Vessel Cooling System (VCS) cannot be obtained. It is difficult for operators to evaluate the reactor outlet coolant temperature supplied from the High Temperature Engineering Test Reactor (HTTR) before rise-to-power tests. Therefore, when the actual measured value of heat released from the RPV or removed by the VCS are changed during rise-to-power tests, operators need to evaluate quickly, within a few minutes, the heat removed by the VCS and the reactor outlet coolant temperature of 30 (MW), at the 100% of the reactor power, before the temperature achieves to 967 (C) which is the maximum temperature limit generating the reactor scram. In this paper, a rapid evaluation method for use by operators is presented.