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Morita, Koji*; Zhang, S.*; Koshizuka, Seiichi*; Tobita, Yoshiharu; Yamano, Hidemasa; Shirakawa, Noriyuki*; Inoue, Fusao*; Yugo, Hiroaki*; Naito, Masanori*; Okada, Hidetoshi*; et al.
Nuclear Engineering and Design, 241(12), p.4672 - 4681, 2011/12
Times Cited Count：15 Percentile：73.97(Nuclear Science & Technology)A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of key phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The key phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, (7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several key phenomena are summarized. The present results demonstrate COMPASS will be useful to understand and clarify the key phenomena of CDAs in SFRs in details.
Koshizuka, Seiichi*; Morita, Koji*; Arima, Tatsumi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Naito, Masanori*; Shirakawa, Noriyuki*; Okada, Hidetoshi*; Uehara, Yasushi*; et al.
Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 11 Pages, 2010/10
In this paper, FY2009 results of the COMPASS code development are reported. Validation calculations for melt freezing and blockage formation, eutectic reaction of metal fuel, duct wall failure (thermal-hydraulic analysis), fuel pin failure and disruption and duct wall failure (structural analysis) are shown. Phase diagram calculations, classical and first-principles molecular dynamics were used to investigate physical properties of eutectic reactions: metallic fuel/steel and control rod material/steel. Basic studies for the particle method and SIMMER code calculations supported the COMPASS code development. COMPASS is expected to clarify the basis of experimentally-obtained correlations used in SIMMER. Combination of SIMMER and COMPASS will be useful for safety assessment of CDAs as well as optimization of the core design.
Morita, Koji*; Zhang, S.*; Arima, Tatsumi*; Koshizuka, Seiichi*; Tobita, Yoshiharu; Yamano, Hidemasa; Ito, Takahiro*; Shirakawa, Noriyuki*; Inoue, Fusao*; Yugo, Hiroaki*; et al.
Proceedings of 18th International Conference on Nuclear Engineering (ICONE-18) (CD-ROM), 9 Pages, 2010/05
A five-year research project has been initiated in 2005 to develop a code based on the MPS (Moving Particle Semi-implicit) method for detailed analysis of specific phenomena in core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The code is named COMPASS (Computer Code with Moving Particle Semi-implicit for Reactor Safety Analysis). The specific phenomena include (1) fuel pin failure and disruption, (2) molten pool boiling, (3) melt freezing and blockage formation, (4) duct wall failure, (5) low-energy disruptive core motion, (6) debris-bed coolability, and (7) metal-fuel pin failure. Validation study of COMPASS is progressing for these key phenomena. In this paper, recent COMPASS results of detailed analyses for the several specific phenomena are summarized.
Shirakawa, Noriyuki*; *; *; *
JNC TJ9440 99-009, 195 Pages, 1999/03
It is necessary for the evaluation of the design base flow rate of water leakage out of the steam generator heat transfer tubes to evaluate the possibility of a tube-failure propagation quantitatively, which needs development of the followings, (1)Blow down model of the steam generator heat transfer tube, (2)Rupture model at high temperature of the steam generator heat transfer tube, and (3)Sodium-water chemical reaction model to analyze the temperature distribution around a leakage point. In this study, development effort is focused on the item (3)that is most important to the evaluation of failure propagation. The FBR safety aalysis code SIMMER-III is used as a reference to realize the analysis and is named SIMMER-SW. The followings were done in this work: (a)Coding of the chemical reaction model, and verification of the model functions, (b)Coding of the rod bundle pressure drop model, and verification of the model functions, (3)Coding of the preprocessor to prepare the input for a rod bundle, (4)Data production of the EOS(Equation-of-State) and TPP(Thermo Physical Properties), (5)Investigation of the interaction between sodium and jet out of a leak hole, (6)Modeling of the thermocouple, (7)SWAT1/P06 test analysis, and (8)SWAT3/Run19 test analysis. The following knowledge was obtained: (a)The results of the SWAT1/P06 test analysis which can be done in two-dimension agree well with that of the experiment at least qualitatively, (b)Thermocouple model is very useful in the analysis such that various kinds of components contact a thermocouple, (c)The constant K in the chemical reaction model is likely to be 0.01
K0.1, and (d)The results of the SWAT3/Run19 test analysis which should be done in three-dimension do not agree with that of the experiment.
