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Okajima, Satoshi; Mori, Takero; Kikuchi, Norihiro; Tanaka, Masaaki; Miyazaki, Masashi
Mechanical Engineering Journal (Internet), 10(4), p.23-00042_1 - 23-00042_12, 2023/08
In this paper, we propose the simplified procedure to estimate failure probability of components subjected to thermal transient for the design optimization. Failure probability can be commonly used as an indicator of component integrity for various failure mechanisms. In order to reduce number of analyses required for one estimation, we have adopted the First Order Second Moment (FOSM) method as the estimation method of failure probability on the process of the optimization, and an orthogonal table in experiment design method is utilized to define conditions of the analyses for the evaluation of the input parameters for the FOSM method. The superposition of ramp responses is also utilized to evaluate the time history of thermal transient stress instead of finite element analysis. Through the demonstration study to optimize thickness of cylindrical vessel subjected to thermal transient derived from shutdown, we confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation cost.
Okajima, Satoshi; Mori, Takero; Kikuchi, Norihiro; Tanaka, Masaaki; Miyazaki, Masashi
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
In this paper, we propose the simplified procedure to estimate failure probability of components subjected to thermal transient for the design optimization. Failure probability can be commonly used as an indicator of component integrity for various failure mechanisms. In order to reduce number of analyses required for one estimation, we have adopted the First Order Second Moment (FOSM) method as the estimation method of failure probability on the process of the optimization, and an orthogonal table in experiment design method is utilized to define conditions of the analyses for the evaluation of the input parameters for the FOSM method. Through the demonstration study to optimize thickness of cylindrical vessel subjected to thermal transient derived from shutdown, we confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation cost.
Kikuchi, Norihiro; Mori, Takero; Okajima, Satoshi; Tanaka, Masaaki; Miyazaki, Masashi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
In JAEA, the design optimization method for plant structure has been developed on the process to output optimal solution of the thickness of reactor vessel wall against thermal transient and seismic loads in a SFR as a representative problem. Resistance characteristic of the wall on the load derived from thermal transient is one of the most important factors for safety estimation on the structural integrity. Failure probability of components against thermal transient was set to one of variables in the objective function for a common scale to compare with other variables in different failure mechanisms. In the iterative process to achieve the optimal solution, a number of evaluations to measure the influence on the load derived from thermal transient was necessarily conducted. More reduction of required time for evaluations is desired. To perform the iterative evaluation process efficiently, the automatization of parametric analyses was implemented in the optimization process.
Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Tanaka, Masaaki; Doda, Norihiro; Mori, Takero; Yokoyama, Kenji; Uwaba, Tomoyuki; Okajima, Satoshi; Matsushita, Kentaro; Hashidate, Ryuta; Yada, Hiroki
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Japan Atomic Energy Agency is developing an innovative design system named ARKADIA to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In the first phase of its development, ARKADIA-Design for design study and ARKADIA-Safety for safety assessment will be developed individually. In this paper, focusing on the ARKADIA-Design, the concept of the system is described and numerical analysis codes to be used for the multi-level and multi-physics analyses are introduced. Descriptions of the practical functions composed by the analysis codes and the representative problems in application studies for validation are introduced.
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.
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.
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.
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.
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
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.
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.
Yamada, Fumiaki; Mori, Takero; Miyakawa, Akira; Konomura, Mamoru
no journal, ,
no abstracts in English
Mori, Takero; Yamada, Fumiaki; Miyakawa, Akira; Konomura, Mamoru
no journal, ,
no abstracts in English
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.
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.
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.
Mori, Takero; Sotsu, Masutake; Ohira, Hiroaki
no journal, ,
no abstracts in English
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.
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.
