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Doda, Norihiro; Yoshimura, Kazuo; Hamase, Erina; Yokoyama, Kenji; Uwaba, Tomoyuki; Tanaka, Masaaki
Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 3 Pages, 2022/09
ARKADIA-Design is being developed to support the optimization of sodium-cooled fast reactors in the conceptual design stage. Design optimization requires various types of numerical analysis: 1-D plant dynamics analysis for efficient evaluation of various design options and multi-dimensional analysis for a detailed evaluation of local phenomena, including multi-physics. For those analyses, ARKADIA-Design performs whole plant analyses based on the multi-level simulation (MLS) technique in which the analysis codes are coupled to simulate the phenomena in an intended degree of resolution. This paper describes an outline of the coupling analysis methods in the MLS of the ARKADIA-Design and the numerical simulations of the experimental fast breeder reactor EBR-II tests by the coupled analysis.
Doda, Norihiro; Hamase, Erina; Kikuchi, Norihiro; Tanaka, Masaaki
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 10 Pages, 2022/04
In conventional design studies of sodium-cooled fast reactors, plant dynamics and local phenomena were evaluated separately by using simple models and detailed models, respectively, and their interaction was considered through the boundary conditions settings with conservativeness for each individual analysis. Thus, the final result through the analyses may contain excessive conservativeness. Therefore, JAEA began to develop a multi-level simulation system in which detailed analysis codes are coupled with a plant dynamics analysis code. Focusing on core thermal-hydraulics, a coupled analysis method using a plant dynamics analysis code Super-COPD and a subchannel analysis code ASFRE has been developed. The analysis on a test in the experimental fast reactor EBR-II was performed to validate the coupled analysis. Through the comparison of the analysis results and the measurement, it was confirmed that the coupled analysis could predict the transient temperature distribution in the subassembly, and the multi-level simulation by changing the level of detail in analysis model could be performed for core thermal-hydraulics.
Doda, Norihiro; Uwaba, Tomoyuki; Nemoto, Toshiyuki*; Yokoyama, Kenji; Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 26, 4 Pages, 2021/05
For design optimization of fast reactors, in order to consider the feedback reactivity due to thermal deformation of the core when the core temperature rises, which could not be considered in the conventional design analysis, a neutronics, thermal-hydraulics, and structure mechanics coupled analysis method has been developed. Neutronics code, plant dynamics code, and structural mechanics code are coupled by a control module in python script. This paper outlines the coupling method of analysis codes and the results of its application to an experiment in an actual plant.
Doda, Norihiro; Hamase, Erina; Yokoyama, Kenji; Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 25, 4 Pages, 2020/06
With the aim of advancing the design optimization in fast reactors, neutronics and thermal-hydraulics coupled analysis method which can consider the temporal change of neutron flux distribution in the core has been developed. A three-dimensional neutronics analysis code and a plant dynamics analysis code are coupled on a platform using Python programing. In this report, outlines of the coupling method of analysis codes, the results of its application to the actual plant under a virtual accidental condition, and the future development is described.
secondary system and turbomachineryHumrickhouse, P. W.*; Sato, Hiroyuki; Imai, Yoshiyuki; Sumita, Junya; Yan, X.
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 9 Pages, 2018/10
This work describes the development of a RELAP5-3D model of the HTTR-GT/H plant secondary system. The RELAP5-3D model presently includes detailed models of several of the heat exchangers in the secondary system as well as the turbomachinery, which includes two compressors and two gas turbines connected to a common shaft and motor. The predictions of the model agreed well to design parameters in both sole power generation and hydrogen co-generation modes in most instances. Both the turbomachinery and heat exchanger models rely on extensive customization via RELAP5-3D control variables, and these implementations are outlined in detail. Potential improvements to the RELAP5-3D turbine model are discussed.
Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki
Nihon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00431_1 - 16-00431_11, 2017/04
A plant dynamics analysis code Super-COPD is being developed in JAEA for the design and safety assessments of sodium-cooled fast reactors (SFRs). In this study, the friction loss coefficients in the whole core thermal-hydraulic model was modified to improve the prediction accuracy of the sodium temperature distribution in a fuel subassembly under the natural circulation conditions. The modified whole core model was applied to analyses of experiments that were performed by using JAEA's test facility PLANDTL as a part of the code validation study. The obtained numerical results of sodium temperature distributions in the core showed good agreement with the measured data. It implies that the modified whole core model can properly reproduce dominant thermal-hydraulic phenomena in the core region under natural circulation conditions, i.e., flow redistribution among fuel subassemblies as well as in a fuel subassembly and inter-subassembly heat transfer.
JNC TN4400 2000-002, 33 Pages, 2000/06
JNC-TN4400-2000-002.pdf:5.22MB
An on-site plant analyzer can provide analysis support in evaluating plant dynamic characteristics when unplanned events occur in a nuclear power station. The plant analyzer contains a plant-dynamics analysis code, which efficiently and quickly analyzes the plant dynamic characteristics. Elements being developed for the on-site plant analyzer include utilities to build plant models for performing analyses and to retrieve plant operation data. The addition of these elements to the analysis code supports the plant-dynamics analysis works in MONJU, in particular, to assist the analyses of start up tests. The system contains the FBR plant-dynamics analysis code "Super-COPD", which is based on the "COPD" code that was used in the safety licensing of MONJU. One feature of the system is that all operations, e.g., assembling plant models for analysis, are prepared using a GUI (Graphical user Interface). In addition to this feature, the system is able to retrieve directly on- and off-line plant data from MIDAS, the Monju Integrated Data Acquisition System. These plant data are used to supply time-dependent boundary conditions for the plant analysis models. For this report, two case studies were performed. First, the analysis result of a turbine trip test at 40% power operation using the full plant model is described. For the second, performance of the IHX model was evaluated using retrieved plant data for boundary conditions. With the development of this system, improvement in the efficiency of analyses of MONJU start-up tests is expected.
Sakai, Takaaki; *; Ohshima, Hiroyuki; Yamaguchi, Akira
JNC TN9400 2000-033, 94 Pages, 2000/04
JNC-TN9400-2000-033.pdf:4.36MB
The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. ln this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube fairer accidents in a steam generator. ln this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in "Equivalent plant" with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. ln conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to conform the heat transfer reduction by the oxidize film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance.
Ohshima, Hiroyuki; Sakai, Takaaki; ; Yamaguchi, Akira; Nishi, Yoshihisa*; Ueda, Nobuyuki*; *
JNC TN9400 2000-077, 223 Pages, 1999/05
JNC-TN9400-2000-077.pdf:6.24MB
The feasibility study (Phase l) is being carried out at JNC to build up new design concepts of practical fast reactors (FRs) from the viewpoint of economy, safety, effective use of resources, reduction of environmental burden and non-proliferation. This report describes the results of the investigation, related to decay heat removal, core/fuel-assembly thermal-hydraulics and thermal-hydraulic correlations, that was performed in fiscal l999 as a part of the feasibility study. ln the study of the decay heat removal, the effects of several design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) in a middle-scale sodium-cooled FR were clarified by using a plant dynamic analysis code. The upper limit of RVACS performance was preliminarily estimated at approximately 0.5
0.6 MWe. Numerical methods for the plant dynamic analysis of gas-and heavy-metal-cooled FRs were also developed. They were applied to the preliminary calculations of the transition from scram to natural circulation and the transient characteristics in tentative plant design concepts were clarified. ln addition, a dimensionless number indicating natural circulation performance was deduced for the comparison of several plant design concepts. With respect to the core/fuel-assembly thermal-hydraulics, numerical analysis methods were improved for the pin-type fuel assembly of gas-and heavy-metal-cooled FRs, the coated-particle- type fuel assembly of helium-gas-cooled FR, and the ductless core of sodium-and heavy-metal-cooled FRs. As preliminary evaluations, thermal-hydraulics in the heavy-metal-cooled FR fuel assembly was compared with sodium-cooled one and thermal-hydraulic analyses of carbon-dioxide- and helium-gas-cooled FR fuel assemblies were performed. The analysis for the fuel assembly with inside duct of sodium-cooled FR was also carried out. The correlations of pressure loss and heat transfer coefficient were investigated for the thermal-hydraulic ...
Hirano, Masashi; Hada, Kazuhiko
JAERI-M 90-071, 68 Pages, 1990/04
no abstracts in English
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.
Doda, Norihiro; Araseki, Hideo; Nabeshima, Kunihiko; Takata, Takashi; Tanaka, Masaaki; Ohshima, Hiroyuki
no journal, ,
JAEA has started the development of a multi-level plant simulation system as a powerful tool for the design studies of sodium-cooled fast reactors. This system enables from highly-efficient to highly-accurate analyses of various operating statuses including normal operation, abnormal transient and accidents by employing multi-level models, and is able to respond flexibly to most of the design options. The conceptual design of the new system, the coupling method of codes as a key of the system development, and the development plan are reported in this presentation.
Doda, Norihiro; Araseki, Hideo; Nabeshima, Kunihiko; Takata, Takashi; Tanaka, Masaaki; Ohshima, Hiroyuki
no journal, ,
JAEA has developed a multi-level plant simulation system to enhance a safety assessment technology for sodium cooled fast reactor. The system, in which high-efficient plant dynamics analyses with a one-dimensional system code and high-accurate local phenomena analyses with multi-dimensional codes are coupled, may apply to the various design options of sodium cooled reactor. In this report, the present status in the development of the basic functions in the system are reported.
Doda, Norihiro; Hamase, Erina; Kikuchi, Norihiro; Tanaka, Masaaki
no journal, ,
In the conventional analysis in the sodium-cooled fast reactor (SFR) plant design, the physical phenomena occurring in each part of the plant are evaluated individually and the interaction between the phenomena is considered under conservative boundary conditions. Therefore, it can be a conservative evaluation in viewpoint of the total balance in the plant design. In this investigation, the integrated platform wchich can couple the detailed analysis codes with the plant dynamics analysis code in order to consider the interaction between the phenomena has been developed as a support tool for achieving optimized design of the SFR plant. In this report, outlines of the development purpose of the integrated platform, the coupling method of analysis codes, and the future development is described.
Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
JAEA has developed a multi-level plant simulation system to enhance a safety assessment technology for sodium cooled fast reactor (SFR). The system, in which high-efficient plant dynamics analyses with a one-dimensional system code and high-accurate local phenomena analyses with multi-dimensional codes are coupled, may apply to the various design options of SFR. We have developed two coupling methods with plant dynamics analysis code for fuel assembly thermal-hydraulics analysis code and for neutronics analysis code, respectively. Using the coupling methods, we performed analyses on the thermal-hydraulics coupling problem between whole core and a fuel assembly and on the nuclear-thermal coupling problem under unprotected conditions. The coupled analysis results were compared with the 1D code results of the same problems. The results showed that each coupling method was validated.
Doda, Norihiro; Yokoyama, Kenji; Tanaka, Masaaki; Takata, Takashi; Ohshima, Hiroyuki
no journal, ,
A multi-level plant simulation system has developed to enhance a safety assessment technology for sodium cooled fast reactor (SFR). The system, in which high-efficient analyses with a one-dimensional plant dynamics analysis code and high-accurate local phenomena analyses with multi-dimensional codes are coupled, may apply to the various design options of SFR. For a validation of the system, the analysis on the experimental fast reactor EBR-II with the SHRT-45R test condition and a virtual condition of control rod withdrawn condition were performed. From the comparison of the analysis results and the test results, it was confirmed that the system could perform the evaluation by changing the level of detail of the analysis model according to the intended use.
Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki
no journal, ,
Validation of an analysis model for a plant dynamic analysis code named Super-COPD including neutronics calculation of a one-point reactor kinetics model necessitates the further work on the beyond design basis accident. Therefore, JAEA participated in IAEA benchmark for Loss of Flow without Scram (LOFWOS) test No.13 performed at the Fast Flux Test Facility (FFTF), and the transient analysis at the first blind phase considering with major reactivity feedback mechanisms was carried out. It was observed that the whole plant dynamics analysis could follow the measured data. As a future work, the gap conductance model for transient, the upper plenum of reactor vessel with dividing several regions or multi-dimensional modeling, and the core model that can evaluate the radial heat transfer rate more accurately will be refined.
