Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08
An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.
Tanaka, Masaaki; Miyake, Yasuhiro*; Ezure, Toshiki; Hamase, Erina
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
The numerical analysis model for the computational fluid dynamics (CFD) code for the design study is developed to evaluate the thermal-hydraulics in the core under the core-plenum interaction (CPI) during the decay heat removal using the dipped type direct heat exchanger (D-DHX). To judge the adequacy of the numerical results for a validation study with the sodium experiment results conducted at PLANDTL-2 facility, the degree of difference (DoD) between the numerical and experimental results must be measured by using the area validation metrics (AVM). Through the examinations, the applicability of the AVM and MAVM based on the p-box method was confirmed.
Tanaka, Masaaki; Kudo, Yoshiro*; Nakada, Kotaro*; Koshizuka, Seiichi*
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.1473 - 1484, 2019/08
Verification and validation (V&V) including uncertainty quantification on modeling and simulation activities has been very much focused on. Due to increase of requirement for standardization of the procedures on the V&V and prediction process to enhance the simulation credibility, "Guideline for Credibility Assessment of Nuclear Simulations (AESJ-SC-A008: 2015)" was published on July 2016 from the AESJ through ten-year discussion. The paper describes brief history of discussion in the AESJ to the publication and introductory explanation of the procedures in the five major elements and one scheme described in the Guideline. And also, a practical experience of the V&V activity according to the fundamental concept indicated in the Guideline is introduced.
Nakamura, Hideo
Nihon Genshiryoku Gakkai-Shi ATOMO
, 61(4), p.270 - 272, 2019/04
no abstracts in English
Tanaka, Masaaki
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 14 Pages, 2018/10
A numerical simulation code MUGTHES has been developed to estimate high cycle thermal fatigue in SFRs. In development of numerical simulation code, verification, validation, and uncertainty quantification (VVUQ) are indispensable. In this study, numerical simulation at impinging jet condition in the WATLON experiment which was the water experiment of a T-junction piping system was performed for the fundamental validation. Based on the previous studies, the simplified least square version GCI method and the area validation metrics were employed as reference methods to quantify uncertainty and to measure the degree of difference between the numerical and the experimental results, respectively. Through the examinations, the potential applicability of the MUGTHES to the thermal striping phenomena was indicated and requirements of modification in the simulation was suggested in accordance with the uncertainty values.
Kikuchi, Norihiro; Doda, Norihiro; Hashimoto, Akihiko*; Yoshikawa, Ryuji; Tanaka, Masaaki; Ohshima, Hiroyuki
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2018/06
For the thermal-hydraulic design regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been developed by JAEA. ASFRE was applied to numerical simulations of several kinds of water and sodium experiments as its validation studies and it was confirmed that pressure drops and temperature distributions measured in the experiments can be well reproduced. To enhance safety of sodium-cooled fast reactor, it is required to evaluate thermal-hydraulics in a core during decay heat removal by natural circulation. It is necessary to estimate radial heat transfer phenomena between fuel assemblies. In this study, a numerical simulation of a 37-pin bundle sodium experiment with radial heat flux was carried out and it was confirmed that ASFRE can be qualitatively reproduced temperature distributions in a fuel assembly affected by radial heat transfer.
Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*
Dai-22-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2017/06
In JAEA, a numerical simulation code named MUGTHES which can deal with conjugate heat transfer between the fluid and the structure parts has been developed for estimation of the thermal fatigue issue. In fundamental validation, the benchmark analysis was considered using the experiment of planar triple parallel jet sodium test (PLAJEST). Three specific experimental conditions at Vr=1, 1.56, and 5.56 were employed for the benchmark analyses according to the knowledge in the literatures. Through the benchmarks, applicability of the large eddy simulation (LES) approach with the standard Smagorinsky model in MUGTHES to simulate thermal striping phenomena was potentially confirmed and issues to be modified in the future works were indicated.
Tanaka, Masaaki
Keisan Kogaku Koenkai Rombunshu (CD-ROM), 22, 4 Pages, 2017/05
In the development of the simulation code and the numerical estimation for high cycle thermal fatigue on a structure caused by thermal striping phenomena in sodium cooled fast reactors, implementation of verification and validation (V&V) process is indispensable. A procedure named V2UP (Verification and Validation plus Uncertainty quantification and Prediction) has been made by referring to the existing guidelines regarding the V&V and the methodologies of the safety assessment. In this paper, a challenging installation of quality management procedures into the V2UP procedure is attempted based on the JSCES Standard for "A Model Procedure for Engineering Simulation".
Doda, Norihiro; Ohira, Hiroaki; Kamide, Hideki
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.1618 - 1625, 2016/04
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, an analysis of EBR-II (Experimental Breeder Reactor II) shutdown heat removal test using the method was performed. The results demonstrated that the evaluation method sufficiently predicts the whole plant thermal hydraulic behaviors and the maximum coolant temperature in a fuel subassembly during natural circulation decay heat removal operations.
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.
Ohgama, Kazuya; Ikeda, Kazumi*; Ishikawa, Makoto; Kan, Taro*; Oki, Shigeo
no journal, ,
no abstracts in English
Ohgama, Kazuya
no journal, ,
no abstracts in English
Suyama, Kenya
no journal, ,
We, researchers of Japan, have been participating in numerous international neutronics computer code benchmarks organized by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD/NEA) for many years. This presentation shows examples of the international computer code benchmarks which the speaker has been involved and the role of the benchmarks for the verification and validation of computer codes in future would be discussed.
Ohgama, Kazuya
no journal, ,
no abstracts in English
Yokoyama, Kenji
no journal, ,
no abstracts in English
Tanaka, Masaaki
no journal, ,
In the development of the simulation code and the numerical estimation for high cycle thermal fatigue on a structure caused by thermal striping phenomena in sodium cooled fast reactors, implementation of verification and validation (V&V) process is indispensable. A procedure named V2UP (Verification and Validation plus Uncertainty quantification and Prediction) has been made by referring to the existing guidelines regarding the V&V published by AESJ. In this presentation, installation of quality management procedures into the V2UP procedure is attempted based on the existing standard regarding guidelines of quality management system.
Ohgama, Kazuya; Takegoshi, Atsushi; Katagiri, Hiroki*; Hazama, Taira
no journal, ,
no abstracts in English
Tanaka, Masaaki; Kobayashi, Jun; Nagasawa, Kazuyoshi*
no journal, ,
A numerical simulation code named MUGTHES has been developed for estimation of the thermal striping issue. In fundamental validation, the benchmark analysis was considered using the experimental data of planar triple parallel jet sodium test (PLAJEST). Three specific experimental conditions of PLAJEST at Vr=1, 1.56, and 5.56 in ration of discharged velocities of the side jet to the center jet were employed for the benchmark analyses. Through the benchmarks, applicability of the large eddy simulation (LES) approach in MUGTHES to simulate thermal striping phenomena was potentially confirmed and issues to be modified in the future works were indicated.
Watanabe, Tomoaki; Suyama, Kenya; Tada, Kenichi; Ferrer, R.*; Hykes, J.*
no journal, ,
The CASMO5 neutron data library based on JENDL-5 is generated and validated by comparison with the conventional ENDF/B-VII.1 library analyzing TCA criticality experiments and Takahama-3 PIE data. In the present library, such data as cross section, decay data, fission yield, etc based on ENDF/B-VII.1 are replaced by those of JENDL-5 as much as possible. As a result of the TCA experiment analysis, k
with the JENDL-5 library is slightly greater than those with the ENDF/B-VII.1 library for all cases analyzed. This tendency is also observed by same analysis using MVP3.0. For the PIE analysis, C/E values of nuclide inventory are almost comparable to those with the ENDF/B-VII.1 library and improved for a few nuclides such as Cs.
Onoe, Hironori*; Saegusa, Hiromitsu*; Tanaka, Tatsuya*; Ishida, Keisuke*; Fujisaki, Kiyoshi*; Sawada, Atsushi
no journal, ,
no abstracts in English
