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Sotsu, Masutake
Journal of Energy and Power Engineering, 14(8), p.251 - 258, 2020/08
Thermal expansion behavior was investigated for evaluation of the core support plate expansion reactivity in the Unprotected Loss of Heat Sink reactor trip failure event. A possibility of mechanical restraint was investigated in thermal expansion of the core structure for the prototype fast breeder reactor Monju. The reactor core expansion was simulated in a three-dimensional finite element analysis model of the reactor vessel considering detailed temperature distribution of the sodium coolant based on the thermal-hydraulic analysis result of the whole core model. It was found that the thermal expansion of the core was not restrained in the ULOHS evert, although part of the core structure is mechanically restrained.
Sotsu, Masutake; Hazama, Taira
Journal of Energy and Power Engineering, 13(11), p.393 - 403, 2019/11
This paper describes methods and results of the uncertainty evaluation of the significant plant response analysis of the reactor trip failure event, i.e. anticipated transient without scram of the Japanese prototype fast breeder reactor Monju. Unprotected loss of heat sink has a relatively large contribution to the core damage frequency due to reactor trip failure. The uncertainty of the allowable time to core damage in this event by plant transient response analysis, so far, has been estimated with considering the range of reactivity coefficients. There are some cases where core damage is considered to be avoided. Specifically, it is assumed that the core damage due to the ULOHS event would be avoided if the sodium temperature at the pump inlet stays below 650
C for 1 h; otherwise the possibility of cavitation occurring in the hydrostatic bearing increases. In this study, a method is developed to search for a solution as an inverse problem of multiple input variables that satisfy the temperature condition. This paper, as a first step, describes input conditions and probability to satisfy the temperature are evaluated through analyses treating input parameters, reactivity coefficients and kinetic parameters, as variables within the design range.
Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke
Journal of Energy and Power Engineering, 9(7), p.599 - 607, 2015/07
The analysis on a density stratification layer consisting of multiple gases in the reactor containment vessel is important for the safety assessment of sever accidents. The Japan Atomic Energy Agency (JAEA) has started the project on the containment thermal hydraulics. We carried out Computational Fluid Dynamics (CFD) analyses in order to investigate the erosion of the density stratification layer by a vertical buoyant jet under this project. We used the Reynolds averaged numerical simulation (RANS) and Large eddy simulation (LES) models to analyze the erosion of a density stratification layer by a vertical buoyant jet in a small vessel which represents a containment vessel. This numerical study calculates the turbulent mixing of a two-component (air and helium) gas mixture. The turbulence models used for the RANS analyses are two types of k-
models. The first model is the low Reynolds number k- model developed by Launder and Sharma. The second model is modified from the first model in order to accurately consider the turbulent production and damping in a stratification layer. The results indicated while the erosion rate calculated by the low-Re k- model was much faster than that of the LES model, the modified k- model could calculate the erosion rate similar to the LES result.
Takeda, Takeshi; Onuki, Akira*; Nishi, Hiroaki*
Journal of Energy and Power Engineering, 9(5), p.426 - 442, 2015/05
Sotsu, Masutake
Journal of Energy and Power Engineering, 8(7), p.1286 - 1292, 2014/07
In order to evaluate the operational reliability of Japanese fast breeder reactor MONJU, frequencies of important intermediate events and equipment failures resulting during reactor automatic trip are predicted using fault tree analysis technique for the plant system model. The targeted devices are the following: primary heat transport system (PHTS), secondary heat transport system (SHTS), water and steam system (WS), plant protection system (PPS) and plant control system (PCS). In this paper was estimated the frequency of automatic reactor trips by extracting and analyzing the important intermediate events and equipment failures covering all the derived fault trees of these systems. The analyses predicted 1.2/reactor year (RY) the value of unplanned shut down frequency by the internal factor of the system.
Kamiyama, Kenji; Tobita, Yoshiharu; Suzuki, Toru; Matsuba, Kenichi
Journal of Energy and Power Engineering, 8(5), p.785 - 793, 2014/05
Kato, Masato; Abe, Tomoyuki
Journal of Energy and Power Engineering, 7(10), p.1865 - 1870, 2013/10
Mori, Takero; Sotsu, Masutake; Honda, Kei; Suzuki, Satoshi*; Ohira, Hiroaki
Journal of Energy and Power Engineering, 7(9), p.1644 - 1655, 2013/09
The prototype fast breeder reactor "MONJU" has an ex-vessel fuel storage system which consists mainly of an ex-vessel fuel storage tank (EVST) and an EVST sodium cooling system. EVST sodium cooling system consists of three independent loops. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an station blackout (SBO) were performed. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450C. However, the structural integrity of the EVSS was maintained. The analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.
Takeda, Takeshi; Otsu, Iwao; Nakamura, Hideo
Journal of Energy and Power Engineering, 7(6), p.1009 - 1022, 2013/06
Takeuchi, Masayuki; Sano, Yuichi; Nakajima, Yasuo; Uchiyama, Gunzo; Nojima, Yasuo*; Fujine, Sachio*
Journal of Energy and Power Engineering, 7(6), p.1090 - 1096, 2013/06
The corrosion behavior of a titanium-5% tantalum alloy (Ti-5Ta) in hot nitric acid condensate was investigated to understand aging behavior of reprocessing equipments. On the basis of long-term immersion tests, it was determined that the corrosion of Ti-5Ta in nitric acid condensate is accelerated with an increase in the concentration. The corrosion rate was nearly constant during the immersion test and the coupons suffered from uniform corrosion. In addition, it is important to note that the nitric acid concentration in the condensate increased on addition of metal salts to the heated nitric acid solution. The larger valence of metal ions was contributed to the increase in the concentration of nitric acid condensate. Consequently, the metal salt in the heated nitric acid solution accelerates the corrosion of Ti-5Ta in the condensate. Therefore, the nitric acid condensate condition should be carefully considered for the corrosion environment of titanium and its alloys.
Ohira, Hiroaki; Honda, Kei; Sotsu, Masutake
Journal of Energy and Power Engineering, 7(4), p.679 - 688, 2013/04
Fujita, Tomoo; Shinkai, Fumiaki*; Nobuto, Jun*
Journal of Energy and Power Engineering, 6(8), p.1191 - 1203, 2012/08
This study aims at validating applicability of the theoretical formula by Gustafson and Still to analysis of permeation behavior of Bingham fluid into a one-dimensional micro-fissure by comparing analysis results with laboratory tests data conducted using bentonite slurry as a Bingham fluid. The results showed that the calculation values by Gustafson and Stille model agreed well with values of laboratory penetration test.
Fujita, Tomoo; Sugita, Yutaka; Toida, Masaru*
Journal of Energy and Power Engineering, 5(5), p.419 - 427, 2011/05
For the geological disposal of high level radioactive wastes, an excavation damaged zone (EDZ) having high hydraulic conductivity resulting from the development of fractures in the rock adjacent to the tunnels will be one of the potential pathways for radioactive contaminant transport,. The potential pathways will be sealed by closure components, that is, a combination of tunnel plug, backfill and grout, the latter material being a clay-based mixture in consideration of the need for long-term stability of the seals. Clay-based grout is one of the effective candidate materials that can be used to interrupt the migration of radionuclides through an EDZ. Laboratory testing of clay-based grout using pulverized bentonite, with the objective of improvement in grout penetration into a rockmass, was conducted. The results showed that the pulverization of clay-based grout had a positive effect on filtration.
