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Ohira, Hiroaki; Doda, Norihiro; Kamide, Hideki; Iwasaki, Takashi*; Minami, Masaki*
Proceedings of 2015 International Congress on Advances in Nuclear Power Plants (ICAPP 2015) (CD-ROM), p.2585 - 2592, 2015/05
IAEA's Coordinated Research Project on Benchmark Analyses of Shutdown Heat Removal Test (SHRT) performed at the Experimental Breeder Reactor-II (EBR-II) has been carried out since 2012. The benchmark specifications were provided by the Argonne National Laboratory (ANL) and the model development for thermal-hydraulics codes and/or plant dynamics codes has been conducted by participating organizations. The experimental data were also provided by the ANL after the calculations have been performed as the blind simulation. JAEA participated in this benchmark analyses, and the plant dynamics analysis code; Super-COPD was applied to the SHRT-17 simulation. The calculated inlet temperature of the high pressure plenum agreed well with the test data in all simulation time. Although the Z-pipe inlet temperature and the IHX intermediate outlet temperature had some discrepancy in the first 400 sec. caused by larger mass flow rate of the primary pump and the perfect mixing model of upper plenum, these temperatures and the flow rate agreed well with the measured data after 400 sec. Hence it was concluded the present analytical model could predict the natural circulation in good accuracy.
Yoshikawa, Shinji; Minami, Masaki*; Takahashi, Tadao*
Journal of Nuclear Science and Technology, 48(4), p.709 - 714, 2011/04
In design problems of large-scale systems like fast breeder reactors, inter-relations among design specifications are very important where a selected specification option is transferred to other specification selections as a premise to be taken account in engineering judgments. These inter-relations are also important in design case studies with hypothetical adoption of rejected design options for evaluation of deviation propagations among design specifications. Some of these rejected options have potential worth for future reconsideration by some circumstance changes (e.g., advanced simulations to exclude needs for mock-up tests, etc), to contribute for flexibility in system designs. In this study, a computer software is built to visualize design problem structure by representing engineering knowledge nodes on individual specification selections along with inter-relations of design specifications, to validate the knowledge representation method and to derive open problems.
Yamada, Fumiaki; Minami, Masaki*
JAEA-Data/Code 2010-023, 79 Pages, 2010/12
JAEA-Data-Code-2010-023.pdf:3.27MB
Japan Atomic Energy Agency constructed a computational model for safety analysis of Monju reactor core to be built into a modularized plant dynamics analysis code Super-COPD code, for the purpose of heat removal capability evaluation at the in total 21 defined transients in the annex to the construction permit application. The applicability of this model to core heat removal capability evaluation has been estimated by back to back result comparisons of the constituent models with conventionally applied codes and by application of the unified model.
Yoshikawa, Shinji; Minami, Masaki*
JAEA-Data/Code 2008-024, 28 Pages, 2009/01
JAEA-Data-Code-2008-024.pdf:5.83MB
A series of information required for numerical simulation of sodium thermal stratification observed at the plant trip test of "Monju" conducted in 1995 is provided, which consists of the test outline, geometry data of the reactor vessel upper plenum between the reactor core top and reactor outlet nozzles, and flow inlet boundary conditions at the reactor core top surface.
Isobe, Mitsutaka*; Toi, Kazuo*; Matsushita, Hiroyuki*; Goto, Kazuyuki*; Suzuki, Chihiro*; Nagaoka, Kenichi*; Nakajima, Noriyoshi*; Yamamoto, Satoshi*; Murakami, Sadayoshi*; Shimizu, Akihiro*; et al.
Nuclear Fusion, 46(10), p.S918 - S925, 2006/10
Times Cited Count：30 Percentile：69.47(Physics, Fluids & Plasmas)no abstracts in English
Minami, Masaki*; Sakata, Hideaki; Yoshikawa, Shinji; Yamada, Fumiaki
JNC TN4410 2005-001, 123 Pages, 2005/03
JNC-TN4410-2005-001.pdf:6.81MB
A software system for straightforward and quick conceptual studies and technical evaluations of fast breeder reactor plants has been developed, mainly targeting the Japanese Demonstration Fast Breeder Reactor Monju. The studies and evaluations by this system used to be limited within steady and nominal conditions, excluding influences by changing specification values in accidental conditions. In this fiscal year, a new software component has been included in the system for simplified evaluation of transient characteristics, which is an essential for complete design of an FBR plant. This new evaluation function enabled to detect specifications to vary over acceptable range in transient conditions, and to notify necessity for re-adjustment of steady state design specification values. This system was also utilized to generate a sensitivity survey program in order to evaluate appropriateness of Monju design. The appropriateness of Monju design was evaluated in two ways. The first one is to follow specification selecting sequence as the as-built Monju has actually been designed. The second one is hypothetical deviations of major specifications of Monju and observation of the influences on other specifications. As a result, Monju design was confirmed to be adequate from the view point of need to meet design limitations at the design stage of Monju. This system is expected to be systematically re-arranged, because the system has now considerably complicated configuration after additions of various programs for many objects. This arrangement will facilitate future contribution of this system to technical studies in order for development of FBRs.
*; *; *
PNC TN941 85-02, 57 Pages, 1985/01
The validation study of the PISCES-2DELK code and a crushable model developed to analyze to the effects of the thermal shield layer under the plug head on the structural response of the prototype reactor vessel were performed by using the shock structural experimental results of 1/33- and 1/15-scale models of prototype reactor vessel. Because the hardening effect of the structures due to higher strain rate is important in the analysis of the scale model experiments, two hardening correlations based on different experimental data were used for the present study. As the results of the present study, it is concluded that (1)PISCES-2DELK code is capable of predicting well the results of the shock structural scale-model experiments, and (2)the cushable model developed to analyze the effects of the thermal shield layer under the plug head is reasonable. By incorporating the crushable madel into PISCES-2DEIK, the overall shock structural response of the reactor vessel is able to be analyzed by the PISCES-2DELK code, including the crushable effects of the thermal shield layer under the plug head.
*; *; *
PNC TN941 84-168, 228 Pages, 1984/12
Two groups of the parametric studies were performed using the PISCES-2DELK Code to examine the problems in the detailed assessment of the structural integrity of the reactor primary system under hypothetical core disruptive accidents (HCDAs) in the prototype fast reactor "Monju". The results were integrated and discussed in the present report. The first group of the studies was performed to examine the sensitivities of uncertainties in input data and modeling employed in the analyses. In the second group of the studies, several important structures were selected to discuss their effects on the structural response of the reactor vessel to HCDA loads. The results of the studies were discussed based on the point of view of an energy flow from an expanding core. It was found that there were general relationships in this energy flow. The relationships were then integrated into simple correlations among: the expansion work of the HCDA bubble; the kinetic energy of the coolant slug upon impact with the plug head; the energy absorbed by the crushable structures under the plug head; the internal energy increase due to the deformation of the reactor vessel upon coolant slug impact; the maximum hoop strain of the reactor vessel. The correlations thus developed have been very useful not only to interprete the result of each parametric case in detail, but also to extrapolate or interpolate the results to cases with different initial and/or boundary conditions without re-running PISCES-2DELK Code.
*; *; *
PNC TN941 84-16, 389 Pages, 1984/01
The objective of the present study is to develop the analytical methods of the structural response of the Monju reactor vessel under Hypothetical Core Disruptive Accidents (HCDAs) by the PISCES-2DELK code which was introduced to the Power Reactor and Nuclear Fuel Development Corporation (PNC) in 1981. Most of the activities in the present study focused on the generic shock structural problems in fast breeder reactors rather than on the problems specific to the Monju reactor. In the first stage of this study, the general hydrodynamic and/or structural behaviors were analysed and discussed for verifying the overall functions of the PISCES-2DELK code. In the second stage, some special functions of the code were examined in detail for future use in shock structural analyses of reactor cases. The many sensitivity studies on these functions were performed based on the model of the 1/30-scale model experiment of the Clinch River Breedar Reactor (CRBR) which was also taken as one of the benchmark problems in the Analysis of Primary Containment Transients (APRICOT) program. Finally, some important problems in the shock structural analyses of the prototype reactor vessel were studied based on a simplified model of the Monju reactor vessel. The surveyed items include the mesh size effect of Euler and shell processors, modeling of energy source and porous material, energy transport model. cavitation model etc. Many of the conclusions derived here are extensively applicable to many shock structural problems. The results from the present study will be effectively reflected to the subsequent in-depth analyses of the Monju reactor vessel under HCDAs.
Yoshikawa, Shinji; Minami, Masaki*; Takahashi, Tadao*
no journal, ,
Aiming at profound understanding of fast breeder reactor plants, a trial software has been built to correlate various technical information items considered in the design stage of the Japanese prototype fast breeder reactor Monju, along with some decision making consequences as of plant characteristics evaluation.
Yoshikawa, Shinji; Minami, Masaki*
no journal, ,
This document provides a set of technical data for the thermal hydraulic analysis of liquid sodium in the upper plenum of Monju reactor vessel, consisting of the additional information which JAEA stated to be offered at first RCM held in September 2008 in Vienna, and the table of temperature versus time measured by the vertical array of thermocouples inserted in the reactor upper plenum during the turbine trip test of the system start-up tests conducted in December 1995.
Yamada, Fumiaki; Minami, Masaki*
no journal, ,
A computational model to simulate integrated behaviors of a reactor core and the connected heat transport system was built and verified by trial calculations on a set of representative anomaly transient event, in order for application of the plant dynamics analysis code Super-COPD to evaluation of the cooling capability of Monju reactor core.
Doda, Norihiro; Igawa, Kenichi*; Minami, Masaki*; Iwasaki, Takashi*; Ohira, Hiroaki
no journal, ,
Sodium-cooled fast reactors have been developed aiming at introducing natural circulation decay heat removal systems by utilizing the characteristic of having a large coolant temperature difference between at the inlet and at the outlet of reactor vessel. In this study, as part of validation for core hot spot evaluation method, which is required for adoption of natural circulation decay heat removal systems, EBR-II (Experimental Breeder Reactor II) shutdown heat removal test was simulated. The simulation results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly in natural circulation decay heat removal.
