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Journal Articles

Validation and applicability of reactor core modeling in a plant dynamics code during station blackout

Mori, Takero; Ohira, Hiroaki; Sotsu, Masutake; Fukano, Yoshitaka

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04

Since safety measures against severe accidents (SAs) such as a long-term station blackout (SBO) are required for Japanese prototype fast breeder reactor Monju, a validation is necessary for the plant dynamics code during SBO. In order to take into account the phenomena in natural circulation: a heat transfer among subassemblies and a flow redistribution, a whole core model has been developed for the plant dynamics code, Super-COPD. This model has been validated by test results of natural circulation in actual facility. In this study, this whole core model was applied to Monju core to evaluate safety measures against SBO, and the pressure loss model of Monju was validated by comparing with results of the plant trip test from the power of 40%. In addition, an analysis was conducted for SBO to investigate the applicability of this model to Monju. The applicability of this model was confirmed by comparing with analytical results using the model without heat transfer between assemblies.

Journal Articles

Analysis of natural circulation tests in the experimental fast reactor JOYO

Nabeshima, Kunihiko; Doda, Norihiro; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki; Iwasaki, Takashi*

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.1041 - 1049, 2015/08

Natural circulation is one of the most important mechanisms to remove decay heat in the sodium cooled fast reactors from the viewpoint of passive safety. On the other hand, it is difficult to evaluate plant dynamics accurately under low flow natural circulation condition. In this study, Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. Almost all plant components in JOYO including four air-coolers were modeled in Super COPD. Furthermore, the full scale modeling of fuel subassembly was also adopted in this analysis. The natural circulation test after reactor scram from 100 MW full power at JOYO was selected and simulated by Super-COPD. The transient behaviors predicted by Super-COPD showed good agreement with the experimental data.

Journal Articles

An Investigation of thermal-hydraulics behavior of MONJU reactor upper plenum under 40%-rated steady state

Honda, Kei; Ohira, Hiroaki; Mori, Takero

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 12 Pages, 2014/12

Thermal-hydraulics analyses of the reactor upper plenum had been performed in IAEA/Monju-CRP from 2008 to 2012. However, all of the participants got a temperature distribution which didn't agree the measured data on the thermocouple plug. In this study, we re-evaluated the inlet boundary conditions and performed another analysis. The calculated temperature distribution on the thermocouple plug had good agreement with the measured data. Thermocouples and flow guide tubes are attached over the subassembly outlets. The calculated temperature at the thermocouples agreed with the temperature of the boundary conditions. And the calculated temperature at the thermocouples had good agreement with the measured data. Therefore, the temperature at the thermocouples can be regarded as the temperature of the subassembly outlets. From these results, the inlet conditions are an appropriate ones.

Journal Articles

Multidimensional thermal-hydraulic analysis on natural circulation behavior in ex-vessel fuel storage tank of MONJU

Ono, Jun; Mori, Takero; Sotsu, Masutake; Ohira, Hiroaki

Proceedings of ASME 2013 International Mechanical Engineering Congress and Exposition (IMECE 2013) (DVD-ROM), 9 Pages, 2013/11

The severe accident evaluation on the EVST of MONJU has ever been performed by one-dimensional flow network code "Super-COPD". However, it is difficult to predict thermal-hydraulics in the EVST accurately because the fluid in the EVST is driven by natural circulation. Thus we have performed multidimensional thermal-hydraulic analysis in order to clarify the thermal-hydraulic behavior and evaluate the appropriateness of the flow network model. As a result, it was noted that the multidimensionality on temperature and velocity in the EVST was small enough and the flow network model would be almost appropriate. It should be noted that flow resistance of the supporting plates or the heat transfer center of the cooling coils should be set conservatively for the safety analysis.

Journal Articles

Plant dynamics evaluation of a MONJU ex-vessel fuel storage system during a station blackout

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 450$$^{circ}$$C. 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.

JAEA Reports

MK-III Performance Tests in JOYO; Control Characteristic Tests of Reactor Coolant Temperature Control System

Ito, Keisuke; Kawahara, Hirotaka; Mori, Takero; Jo, Takahisa; Ariyoshi, Masahiko; Isozaki, Kazunori

JNC TN9410 2005-008, 267 Pages, 2005/03

JNC-TN9410-2005-008.pdf:108.4MB

Control characteristic tests of reactor coolant temperature control system were carried out to confirm its controlling constant to make MK-III hear transport system control stably, and stability against actual disturbance to the plant. The control characteristic test consists of three kinds of tests. As a result, the optimum PI parameters of the reactor coolant temperature control system was confirmed that the proportional gain is between from 0.36 to 1.12(approximately half of MK-II), the integral time is 80 respectively. The gain margin of the control system was between from 7 to 19dB of through the vane opening range.

Oral presentation

Evaluation of the MONJU fuel temperature during transients with a fluctuation plutonium isotopic composition, 2; Effect about effective delayed neutron fraction

Yamada, Fumiaki; Mori, Takero; Miyakawa, Akira; Konomura, Mamoru

no journal, , 

no abstracts in English

Oral presentation

Evaluation of the MONJU fuel temperature during transients with a fluctuation of the plutonium isotopic composition, 1; Effect about Reactivity coefficient

Mori, Takero; Yamada, Fumiaki; Miyakawa, Akira; Konomura, Mamoru

no journal, , 

no abstracts in English

Oral presentation

Development of Monju plant dynamics analysis code, 1; Development plan

Yamada, Fumiaki; Kimura, Koichi; Jo, Takahisa; Mori, Takero; Morizono, Koji; Tamayama, Kiyoshi; Miyakawa, Akira

no journal, , 

In this report that development plan of Monju plant dynamics analysis code.

Oral presentation

Development of Monju plant dynamics analysis code, 3; Verification of Super-COPD code based on the test operation data at 40% rated power

Mori, Takero; Araki, Kosuke*; Kato, Mitsuya*; Takano, Masahito*

no journal, , 

An improved analysis model and the actual component characteristic data for the main cooling system are incorporated in Monju plant dynamics analysis code: Super-COPD. The verification of this new analysis model is based on the results of a plant trip test at 40% rated power and on plant control system characteristics.

Oral presentation

Thermal-hydraulic analysis on reactor upper plenum of MONJU

Honda, Kei; Mori, Takero; Sotsu, Masutake; Ohira, Hiroaki

no journal, , 

Thermal-hydraulics analyses of the reactor upper plenum of Monju, Japanese prototype of FBR, were performed in IAEA/Monju-CRP from 2008 to 2012. However, detail temperature and flow rate conditions of the inlets were required for an accurate analysis. In this paper we re-evaluated the inlet boundary condition (subassembly outlets) and performed another thermal-hydraulics analysis with Star-CCM+. The surface of the structures in the upper plenum was precisely modeled. The structures included a fuel transfer machine, in-vessel racks, flow-guide tubes, etc. The result was following: the structure didn't have large influence to the temperature distribution, and the analysis result of the temperature distribution on the thermocouple plug had some difference from the test result.

Oral presentation

Improvement of the analytical model of Monju air cooler for natural circulation

Mori, Takero; Sotsu, Masutake; Ohira, Hiroaki

no journal, , 

no abstracts in English

Oral presentation

Validation of dynamic simulation code super-COPD by using natural circulation experiments in fast reactor JOYO

Nabeshima, Kunihiko; Doda, Norihiro; Hiyama, Tomoyuki; Ohshima, Hiroyuki; Mori, Takero; Ohira, Hiroaki

no journal, , 

Fast reactor plant dynamics simulation code Super-COPD has been validated through the application to the analysis of natural circulation tests in the experimental fast reactor JOYO. The analysis domain is set from the reactor core to the air cooler so as to focus on the simulation accuracy of natural circulation behavior with the full scale of core assembly modeling.

Oral presentation

Study on prevention of loss of heat removal function for fast reactor, 2; Effectiveness of natural circulation cooling

Yamada, Fumiaki; Mori, Takero

no journal, , 

The core cooling by the nature circulation was effective, that evaluated the form of the circulation root of various coolant in the failure of forced circulation of main heat transport system which was an accident sequence of loss of heat removal system more than design basis accident of the FBR from plant dynamics characteristic analysis.

Oral presentation

Study on prevention of loss of heat removal function for fast reactor, 4; Effectiveness of alternative decay heat removal system

Yoshimura, Kazuo; Aizawa, Kosuke; Ichikawa, Kenta; Mori, Takero; Yamada, Fumiaki

no journal, , 

no abstracts in English

Oral presentation

Study on effectiveness evaluations of countermeasures against severe accidents in fast reactor, 5; Validation of reactor analytical model for PLOHS events

Mori, Takero; Ohira, Hiroaki; Sotsu, Masutake; Fukano, Yoshitaka

no journal, , 

In order to evaluate the availability of the decay heat removal by natural circulation which is a measure for the Protected Loss of Heat Sink (PLOHS) in Fast Reactor, a whole core model which could simulate thermal hydraulics of each subassembly in a reactor was applied to Monju core. The Monju core model was validated by a test result of actual plant, MONJU. In addition, an analysis was conducted for PLOHS to investigate the applicability of this model.

Oral presentation

Safety evaluations for MONJU during decommissioning phase, 2; Conservative thermal evaluations on fuel integrity

Mori, Takero; Sotsu, Masutake; Imaizumi, Yuya; Yoshimura, Kazuo; Fukano, Yoshitaka

no journal, , 

no abstracts in English

Oral presentation

Study on effectiveness evaluations of countermeasures against severe accidents in fast reactor, 6; Effectiveness evaluations of countermeasures against anticipated transient without scram

Imaizumi, Yuya; Yamada, Fumiaki; Nishimura, Masahiro; Mori, Takero; Fukano, Yoshitaka

no journal, , 

no abstracts in English

Oral presentation

Study on effectiveness evaluations of countermeasures against severe accidents in fast reactor, 8; Effectiveness evaluations of countermeasures against severe accidents during reactor shutdown

Nishimura, Masahiro; Yoshimura, Kazuo; Taninaka, Hiroshi; Yamada, Fumiaki; Mori, Takero; Nishino, Hiroyuki; Fukano, Yoshitaka

no journal, , 

no abstracts in English

Oral presentation

Study on effectiveness evaluations of countermeasures against severe accidents in fast reactor, 9; Effectiveness evaluations of countermeasures against severe accidents for spent fuel

Mori, Takero; Taninaka, Hiroshi; Kurisaka, Kenichi; Yamada, Fumiaki; Nishimura, Masahiro; Fukano, Yoshitaka

no journal, , 

no abstracts in English

22 (Records 1-20 displayed on this page)