Progress of sodium-cooled fast reactor developments in Japan taking into account total lifecycle, risk-informed approach, and sustainability

Kamide, Hideki; Asayama, Tai; Wakai, Takashi; Ezure, Toshiki; Uchibori, Akihiro; Kubo, Shigenobu; Takeuchi, Masayuki

Nuclear Engineering and Design, 421, p.113062_1 - 113062_10, 2024/05

https://doi.org/10.1016/j.nucengdes.2024.113062

A sodium cooled fast reactor (SFR) is one of the most relevant and decarbonized energy supply system with higher sustainability on natural resources, footprint, and waste management. It was planned in a strategic roadmap of fast reactor decided by Inter-Ministerial Council for Nuclear Power Japan in 2022 to start a conceptual design of a demonstration reactor from 2024 with a background of accumulated knowledge and experiences of SFR development. For example, a design and lifecycle simulation/evaluation system named ARKADIA has been developed to accelerate such design works. It will enable to take into account plant lifecycle, e.g., operation and maintenance, to the plant design and optimize it based on simulations and knowledgebase. This paper shows research progresses of ARKADIA, safety design and evaluations, codes and standards, fuel cycle, and SFR development projects in Japan.

Numerical simulation technologies for safety evaluation in plant lifecycle optimization method, ARKADIA for advanced reactors

Uchibori, Akihiro; Doda, Norihiro; Aoyagi, Mitsuhiro; Sonehara, Masateru; Sogabe, Joji; Okano, Yasushi; Takata, Takashi*; Tanaka, Masaaki; Enuma, Yasuhiro; Wakai, Takashi; et al.

Nuclear Engineering and Design, 413, p.112492_1 - 112492_10, 2023/11

https://doi.org/10.1016/j.nucengdes.2023.112492

The ARKAIDA has been developed to realize automatic optimization of plant design from safety evaluation for the advanced reactors represented by a sodium-cooled fast reactor. ARKADIA-Design offers functions to support design optimization both in normal operating conditions and design basis events. The multi-level simulation approach by the coupled analysis such as neutronics, core deformation, core thermal hydraulics was developed as one of the main technologies. On the other hand, ARKAIDA-Safety aims for safety evaluation considering severe accidents. As a key technology, the numerical methods for in- and ex-vessel coupled phenomena during severe accidents in sodium-cooled fast reactors were tested through a hypothetical severe accident event. Improvement of the ex-vessel model and development of the AI technology to find best design solution have been started.

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.

ARKADIA; For the innovation of advanced nuclear reactor design

Ohshima, Hiroyuki; Asayama, Tai; Furukawa, Tomohiro; Tanaka, Masaaki; Uchibori, Akihiro; Takata, Takashi; Seki, Akiyuki; Enuma, Yasuhiro

Journal of Nuclear Engineering and Radiation Science, 9(2), p.025001_1 - 025001_12, 2023/04

https://doi.org/10.1115/1.4054726

This paper describes the outline and development plan for ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source. ARKADIA will realize Artificial Intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for the design and operation of a nuclear plant, including optimization of safety equipment. State-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&D are integrated with AI. In the first phase of development, ARKADIA-Design and ARKADIA-Safety will be constructed individually, with the first target of sodium-cooled reactor. In a subsequent phase, everything will be integrated into a single entity applicable not only to advanced rectors with a variety of concepts, coolants, configurations, and output levels but also to existing light-water reactors.

Development of in-service inspection rules for liquid-metal cooled reactors using the system based code concept

Takaya, Shigeru; Asayama, Tai; Yada, Hiroki; Roberts, A. T.*; Schaaf, F.*

Journal of Pressure Vessel Technology, 142(2), p.021601_1 - 021601_5, 2020/04

https://doi.org/10.1115/1.4044344

Inservice inspection rules for liquid-metal cooled plants were historically provided by Section XI, Division 3 of the ASME Boiler and Pressure Vessel Code. However, some parts of the Code remained as being in the course of preparation. Although no major revisions were made to Division 3 since the first issue in 1980, a newly developed and published Code Case N-875, now provides alternative examinations to the methods previously contained in Division 3. The Code Case was developed using the System Based Code concept pursuing rationalization of codes and standards based on reliability targets throughout a plant's service life. In this paper, an overview of the Code Case is presented. The technical foundation to establish the applicability of these alternative examinations as delineated in the Code Case, consists of Stage I and II evaluations with compensating individual considerations. Stage I is a structural integrity evaluation without the contribution of inservice inspections, while Stage II is evaluation of the detectability of a postulated flaw. Not only conventional direct detection methods, but also indirect detection methods are permitted to be employed through the Stage II evaluation. Furthermore, the detailed evaluation procedures are illustrated through the application of the Code Case's evaluation criteria to the primary heat transport piping system of a prototype sodium-cooled fast breeder reactor in Japan, specifically Monju.

Development of core and structural materials for fast reactors

Asayama, Tai; Otsuka, Satoshi

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 15 Pages, 2017/06

This paper summarizes ongoing efforts in Japan Atomic Energy Agency on the development of core and structural materials for sodium-cooled fast reactors. For core materials, oxide dispersion strengthened (ODS) steels and 11Cr ferritic steel (PNC-FMS) will be used for the fuel pin cladding and wrapper tube, respectively. As for ODS steel, 9Cr- and 11Cr-ODS steels have been extensively developed. Their laboratory-scale manufacturing technology has been developed including reliability improvement in tube microstructure and strength homogeneity. As for the PNC-FMS wrapper tube, the development of a dissimilar joining technique with type 316 steel and properties evaluation of dissimilar welds have been carried out. For structural materials, codification of 316FR stainless steel and Modified 9Cr-1Mo steel is ongoing. Acquisition and collection of long-term data of base metal and welded joints are continued and evaluation methodologies are being developed to establish a technical basis for 60-year design.

Determination of in-service inspection requirements for fast reactor components using System Based Code concept

Takaya, Shigeru; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Asayama, Tai

Nuclear Engineering and Design, 305, p.270 - 276, 2016/08

AA2016-0006.pdf:0.51MB

https://doi.org/10.1016/j.nucengdes.2016.05.028

In our previous study, we proposed a new process for determining the in-service inspection (ISI) requirements using the System Based Code concept. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other on plant safety. In this study, the ISI requirements for a reactor guard vessel (RGV) and core support structure (CSS) of a prototype sodium-cooled fast breeder reactor in Japan (Monju) were investigated using the proposed process. It was shown that both components had sufficient reliability even assuming unrealistic severe conditions. The failure occurrences of these components were practically eliminated. Hence, it was concluded that no ISI requirements were needed for these components. The proposed process is expected to contribute to the realization of effective and rational ISI by properly taking into account plant-specific features.

Development of creep-fatigue evaluation method for 316FR stainless steel

Nagae, Yuji; Takaya, Shigeru; Asayama, Tai

Journal of Pressure Vessel Technology, 137(4), p.041407_1 - 041407_5, 2015/08

https://doi.org/10.1115/1.4028497

Application of the system based code concept to the determination of in-service inspection requirements

Takaya, Shigeru; Asayama, Tai; Kamishima, Yoshio*; Machida, Hideo*; Watanabe, Daigo*; Nakai, Satoru; Morishita, Masaki

Journal of Nuclear Engineering and Radiation Science, 1(1), p.011004_1 - 011004_9, 2015/01

https://doi.org/10.1115/1.4026392

A new process for determination of inservice inspection (ISI) requirements was proposed based on the System Based Code concept to realize effective and rational ISI by properly taking into account plant specific features. The proposed process consists of two complementary evaluations, one focusing on structural integrity and the other one on detectability of defects before they would grow to an unacceptable size in light of plant safety. If defect detection was not feasible, structural integrity evaluation would be required under sufficiently conservative hypothesis. The applicability of the proposed process was illustrated through an application to the existing prototype fast breeder reactor, Monju.

Elaboration of the system based code concept; Activities in JSME and ASME, 1; Overview

Asayama, Tai; Miyagawa, Takayuki*; Dozaki, Koji*; Kamishima, Yoshio*; Hayashi, Masaaki*; Machida, Hideo*

Proceedings of 22nd International Conference on Nuclear Engineering (ICONE-22) (DVD-ROM), 7 Pages, 2014/07

https://doi.org/10.1115/ICONE22-30567

This paper is the first one of the series of four papers that describe ongoing activities in the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) on the elaboration of the System Based Code (SBC) concept. A brief introduction to the SBC concept is followed by the technical features of structural evaluation methodologies that are based on the SBC concept. Also described is the ongoing collaboration of JSME and ASME at the Joint Task Group for System Based Code established in the ASME Boiler and Pressure Vessel Code Committee which is developing alternative rules for ASME B&PV Code Section XI Division 3, inservice inspection requirements for liquid metal reactor components.