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.

Development of Lagrangian particle method for temperature distribution formed by sodium-water reaction in a tube bundle system

Kosaka, Wataru; Uchibori, Akihiro; Okano, Yasushi; Yanagisawa, Hideki*

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.1150 - 1163, 2023/08

https://doi.org/10.13182/NURETH20-40116

The leakage of pressurized water from a steam generator (SG) and the progress after that are a key issue in the safety assessment or design of a SG in sodium-cooled fast reactor. The analysis code LEAP-III can evaluate a rate of water leakage during the long-term event progress, i.e., from the self-wastage initiated by an occurrence of a microscopic crack in a tube wall to the water leak detection and water/water-vapor blowdown. Since LEAP-III consists of semi-empirical formulae and one-dimensional equations of conservation, it has an advantage in short computation time. Thus, LEAP-III can facilitate the exploration of various new SG designs in the development of innovative reactors. However, there are several problems, such as an excessive conservative result in some case and the need for numerous experiments or preliminary analyses to determine tuning parameters of models in LEAP-III. Hence, we have developed a Lagrangian particle method code, which is characterized by a simpler computational principle and faster calculation. In this study, we have improved the existing particle pair search method for interparticle interaction in this code and developed an alternative model without the pair search. Through the trial analysis simulating in a tube bundle system, it was confirmed that new models reduced the computation time. In addition, it was shown that representative temperatures of the heat-transfer tubes evaluated by this particle method code, which is used to predict the tube failure in LEAP-III, were good agreement with that by SERAPHIM, which is a detailed mechanistic analysis method code.

Development of element functions and design optimization procedures with knowledge- and AI-aided design integration approach for advanced reactor lifecycle in ARKADIA

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

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 11 Pages, 2023/05

Development of dynamic PRA methodology for external hazards in sodium-cooled fast reactor via applying Markov chain Monte Carlo method to severe accident analysis code; Assessment of accident management of assigning independent emergency diesel generators to each air cooler

Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05

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.

The Development of Petri Net-based continuous Markov Chain Monte Carlo methodology applying to dynamic probability risk assessment for multi-state resilience systems with repairable multi-component interdependency under longtermly thereat

Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi

Journal of Nuclear Science and Technology, 23 Pages, 2023/00

https://doi.org/10.1080/00223131.2023.2290067

Validation study of sodium pool fire modeling efforts in MELCOR and SPHINCS codes

Louie, D. L. Y.*; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Luxat, D. L.*

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 6 Pages, 2022/09

Experiment study on the effect of nozzle shape on liquid jet breakup

Sun, G.*; Zhan, Y.*; Okawa, Tomio*; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Okano, Yasushi

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08

When a liquid sodium leakage accident occurs in a sodium-cooled fast reactor, the injected sodium collides with structures to produce splashing droplets, which can result in a violent combustion. According to previous studies on circular nozzles, the amount of splash is affected by the state of the jet at the moment of impact. However, the outlet shape of damaged area is hardly to be circular; and meanwhile it influences the flow pattern of jet a lot. Considering about this, in the present work, high-speed cameras were used to observe the jet discharged from oval nozzles vertically downward to investigate the falling process of the jet. The result shows that surface wave appears on the jet and within a certain range of flow velocity it can be observed obviously, meanwhile accelerate the breakup of jet.

Sodium-cooled Fast Reactors

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

https://shop.elsevier.com/books/sodium-cooled-fast-reactors/morishita/978-0-12-824076-2

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.

Development of integrated severe accident analysis code, SPECTRA for sodium-cooled fast reactor

Uchibori, Akihiro; Sonehara, Masateru; Aoyagi, Mitsuhiro; Takata, Takashi*; Ohshima, Hiroyuki

Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04

A new computational code, SPECTRA, has been developed for integrated analysis of in- and ex-vessel phenomena during severe accidents in sodium-cooled fast reactors. The in-vessel thermal hydraulics module includes coupled analytical models for multidimensional multifluid model considering compressibility and relocation of a molten core. A lumped mass model is employed for computing behavior of ex-vessel compressible multicomponent gas including aerosols. This model is coupled with the models for ex-vessel phenomena such as sodium fire. Loss of reactor level event starting from leakage of sodium coolant was computed. Basic capability to evaluate severe accident progress was demonstrated through this analysis.