JSME series in thermal and nuclear power generation Vol.3 (Sodium-cooled fast reactor development; R&Ds on thermal-hydraulics and safety assessment towards social implementation)

Tanaka, Masaaki; Uchibori, Akihiro; Okano, Yasushi; Yokoyama, Kenji; Uwaba, Tomoyuki; Enuma, Yasuhiro; Wakai, Takashi; Asayama, Tai

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

The book, JSME Series in Thermal and Nuclear Power Generation Vol.3 Sodium-cooled Fast Reactor, was published as a 30th anniversary memorial project of Power & Energy Systems Division. This paper describes an introduction of the book on a part of key technologies regarding safety assessment, thermal-hydraulics, neutronics, and fuel and material development. This introductory paper also provides an overview of an integrated evaluation system named ARKADIA to offer the best possible solutions for challenges arising during the design process, safety assessment, and operation of a nuclear plant over its life cycle, in active use of the R&D efforts and knowledges on thermal-hydraulics and safety assessment with state-of-the-art numerical analysis technologies.

Automatization of parametric analyses of influence factor on load derived from thermal transient in design optimization method for plant structure in sodium-cooled fast reactor

Kikuchi, Norihiro; Mori, Takero; Okajima, Satoshi; Tanaka, Masaaki; Miyazaki, Masashi

Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07

https://doi.org/10.1299/jsmepes.2022.26.A223

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.

Development of a surface heat flux model for urban wind simulation using locally mesh-refined lattice Boltzmann method

Onodera, Naoyuki; Idomura, Yasuhiro; Hasegawa, Yuta; Nakayama, Hiromasa

Dai-35-Kai Suchi Ryutai Rikigaku Shimpojiumu Koen Rombunshu (Internet), 3 Pages, 2021/12

A detailed wind simulation is very important for designing smart cities. Since a lot of tall buildings and complex structures make the air flow turbulent in urban cities, large-scale CFD simulations are needed. We develop a GPU-based CFD code based on a Lattice Boltzmann Method (LBM) with a block-based Adaptive Mesh Refinement (AMR) method. In order to reproduce real wind conditions, the wind condition and ground temperature of the mesoscale weather forecasting model are given as boundary conditions. In this research, a surface heat flux model based on the Monin-Obukhov similarity theory was introduced to improve the calculation accuracy. We conducted a detailed wind simulation in Oklahoma City. By executing this computation, wind conditions in the urban area were reproduced with good accuracy.

Methodology development for transient flow distribution analysis in high temperature gas-cooled reactor

Aoki, Takeshi; Sato, Hiroyuki; Ohashi, Hirofumi

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

https://doi.org/10.1115/ICONE2020-16199

The flow distribution analysis, which is a part of thermal hydraulic design of the prismatic-type of the high temperature gas cooled reactor (HTGR) considering unintended flows between graphite blocks, has been performed for steady and conservative conditions. On the other hand, the transient analysis for satisfactorily realistic conditions will be helpful for the design improvement of prismatic-type HTGR. The present study aims to develop the transient flow distribution analysis code and confirm its applicability for the transient flow distribution analysis for prismatic-type HTGRs during anticipated operational occurrences and accidents utilizing experiences on high temperature engineering test reactor (HTTR) design. The calculation model and code were developed and validated for analysis of the unintended flows in the core and the molecular diffusion dominant in beginning air ingress behavior in an air ingress accident.

Numerical simulation of laser welding different kinds of materials using a thermohydraulics computational science numerical simulation code SPLICE

Muramatsu, Toshiharu; Sato, Yuji; Kamei, Naomitsu; Aoyagi, Yuji*; Shobu, Takahisa

Nihon Kikai Gakkai Dai-13-Kai Seisan Kako, Kosaku Kikai Bumon Koenkai Koen Rombunshu (No.19-307) (Internet), p.157 - 160, 2019/10

no abstracts in English

Development of numerical estimation method for thermal hydraulics in reactor vessel of sodium-cooled fast reactor under decay heat removal system operation conditions; Preliminary thermal hydraulics simulation for simulated reactor vessel in sodium experimental apparatus PLANDTL-2

Tanaka, Masaaki; Ono, Ayako; Hamase, Erina; Ezure, Toshiki; Miyake, Yasuhiro*

Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2018 Koen Rombunshu (CD-ROM), 4 Pages, 2018/08

Decay heat removal system (DHRS) by using the natural circulation without depending on the pump as the mechanical equipment is recognized as one of the most effective methodologies for the sodium-cooled fast reactor from the viewpoint of the safety enhancement. The numerical estimation method which can predict thermal hydraulic phenomena in the natural circulation under the plant cooling process by operating the various DHRSs including the severe accident is necessarily required. In this paper, the numerical results of the preliminary analysis for the sodium experiment condition with the apparatus of PLANDTL-2, in which the core and the upper plenum with a dipped-type direct heat exchanger (DHX) were modeled, were discussed, in order to establish an appropriate numerical models for the direct heat exchanger (DHX).

Study on applicability of fast reactor plant dynamics analysis code to core thermal hydraulics under natural circulation decay heat removal conditions

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

https://doi.org/10.1299/transjsme.16-00431

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.

Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

Kamide, Hideki; Ohshima, Hiroyuki; Sakai, Takaaki; Tanaka, Masaaki

Nuclear Engineering and Design, 312, p.30 - 41, 2017/02

https://doi.org/10.1016/j.nucengdes.2016.09.026

In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related Research and Development results on innovative technologies and lessons learned from Fukushima Dai-ichi Nuclear Power Plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V and V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

Critical heat flux prediction for subcooled flow boiling in annulus

Liu, W.

Dai-20-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu, p.391 - 392, 2015/06

Subcooled flow boiling is a boiling that begins and develops even though the mean enthalpy of liquid phase is lower than saturation. This forced convective boiling is one of the most efficient ways for the removal of high heat flux. It is widely used in the high heat flux components such as nuclear reactor cores, accelerator targets and fusion reactor components. The thermal outputs of these systems are restricted by Critical Heat Flux(CHF). Because of the importance of the CHF, lots of researches, including both experimental and mechanistic modelling, have been performed. However, the CHF prediction for rod bundles still remains unsolved. As the first step for the CHF prediction in rod bundles, in this paper, we tried to predict the CHF in annulus, which is the most basic flow geometry simplified from a rod bundle. We performed the CHF prediction by using liquid sublayer dryout model, combined with Nouri single phase velocity distribution correlation for annulus. The results show that the CHF in annulus can be predicted in an accuracy of about 20%.

Energy confinement and transport of H-mode plasmas in tokamak

Urano, Hajime

JAERI-Research 2004-027, 131 Pages, 2005/02

JAERI-Research-2004-027.pdf:5.69MB

no abstracts in English