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Futagami, Satoshi; Kubo, Shigenobu; Sofu, T.*; Ammirabile, L.*; Gauthe, P.*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10
Yamano, Hidemasa; Kubo, Shigenobu; Tokizaki, Minako*; Nakamura, Hironori*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 12 Pages, 2022/10
Specific design features of advanced sodium-cooled fast reactors (SFRs) designed in Japan are a passive reactor shutdown system, a passive decay heat removal system (DHRS), and an in-vessel retention (IVR) concept against an anticipated transients without scram (ATWS) in design extension condition (DECs). The present paper describes numerical analysis methodologies for event sequences studied in Japan and some numerical analyses of DECs to show the effectiveness of the passive shutdown system against a typical ATWS and severe accident mitigation measures for the IVR of molten core. For the passive shutdown capability, the numerical analysis has demonstrated the effectiveness of a self-actuated shutdown system against a severe ATWS event, for which the temperature response time was separately evaluated by a computational fluid dynamics (CFD) code. A recently developed debris-bed cooling analysis methodology coupled with a CFD code and a debris-bed module has successfully simulated a three-dimensional coolant flow field near the debris bed with the passive DHRS capability in order to demonstrate the debris-bed coolability on a core catcher.
Nakamura, Hideo; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*
Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10
Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*
Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-6) (Internet), 12 Pages, 2016/09
A numerical simulation code named MUGTHES which can deal with conjugate heat transfer problem between the fluid and the structure parts has been developed in order to predict the thermal response in the structure for estimation of the thermal fatigue issue. To perform fundamental validation of the MUGTHES, the benchmark simulation was considered using the experiment of planar triple parallel jets mixing sodium test (PLAJEST). Since it was known by literatures that three representative flow mixing patterns were shown in accordance with the velocity rate of the side jets to the center jet, three typical experimental conditions in the PLAJEST were employed as boundary conditions for the benchmark. Through the numerical simulations, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model to simulate thermal striping phenomena was confirmed.
Uchibori, Akihiro; Watanabe, Akira*; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki
Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-6) (Internet), 11 Pages, 2016/09
For assessment of the wastage environment under tube failure accident in a steam generator of sodium-cooled fast reactors, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. The original SERAPHIM code is based on the finite difference method. In this study, unstructured mesh-based numerical method was developed and introduced into the SERAPHIM code to advance a numerical accuracy for the complex-shaped domain including multiple heat transfer tubes. Validity of the unstructured mesh-based SERAPHIM code was investigated through the analysis of an under expanded jet experiment. The calculated pressure profile showed good agreement with the experimental data. Numerical analysis of water vapor discharging into liquid sodium was also performed. It was demonstrated that the proposed numerical method could be applicable to evaluation of the sodium-water reaction phenomenon.
Takase, Kazuyuki
Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-5) (Internet), 11 Pages, 2014/09
When devising the thermal design of supercritical water reactors, it is necessary to develop an analysis method that correctly predicts the turbulent heat transfer characteristics in subchannels of fuel bundles. Spacers are set into the subchannels to maintain the distances between adjacent fuel rods. The turbulent heat transfer is generally enhanced by the spacers' reduction of the cross-sectional area in the subchannels. However, since the thermophysical properties of supercritical fluids drastically change in the vicinity of a pseudocritical point, the enhancement of the turbulent heat transfer depends on the thermal design. To this end, the Japan Atomic Energy Agency is developing an analysis method that will predict the thermal-hydraulic characteristics of supercritical fluids. The heat transfer calculations were performed using a newly developed code under conditions of a subchannel with a spacer. The enhancement of the turbulent heat transfer coefficient in the subchannels with spacers was analyzed numerically.
Ishigaki, Masahiro; Abe, Satoshi; Shibamoto, Yasuteru; Yonomoto, Taisuke
Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-5) (Internet), 11 Pages, 2014/09
Tanaka, Masaaki; Ohno, Shuji; Ohshima, Hiroyuki
Proceedings of OECD/NEA & IAEA Workshop on Application of CFD/CMFD Codes to Nuclear Reactor Safety and Design and their Experimental Validation (CFD4NRS-5) (Internet), 14 Pages, 2014/09
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.
Okumura, Keisuke; Oka, Yoshiaki*; Ishiwatari, Yuki*
Nuclear Reactor Design, p.49 - 126, 2014/00
As necessary knowledge for nuclear reactor design, fundamental methods of nuclear reactor calculations are described. First, we explain, in an easy-to-understand manner, what kinds of numerical methods and procedures are employed in the codes which treat nuclear data processing, lattice calculation, lattice burnup calculation, core calculation, nuclear and thermal-hydraulic coupled calculation, core burnup calculation and space-dependent core calculation. Next we show the calculated examples of the optimizations for a fuel loading pattern and control rod insertion pattern in the thermal-hydraulic coupled core burnup calculation. Furthermore, we describe the methods of the plant dynamics analyses using a simple node junction model for the heat transfer calculation. The plant dynamics cover the analyses of control and startup properties, reactor stability, and safety analysis.
