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, 61(7), p.935 - 957, 2024/07

https://doi.org/10.1080/00223131.2023.2290067

Development of new simulation method; Design support of next generation advanced reactors

Uchibori, Akihiro; Okano, Yasushi

Isotope News, (793), p.32 - 35, 2024/06

The design of a containment vessel in a sodium-cooled fast reactor was optimized from simulation on the hypothetical severe accident including sodium leakage and combustion. The simulation method is one of the base technologies of the design optimization system, ARKADIA. The simulation was performed on the different design conditions including volume of the containment vessel and the safety equipment as optimization parameters. The iterative simulation successfully found that the safety under this accident was kept even in the downsized containment vessel by selecting an effective safety equipment. This study demonstrated that the developed method has basic capability for design optimization in ARKADIA.

Development of severe accident simulation code for sodium-cooled fast reactors: SIMMER-V, 2; Development and verification of detailed fuel pin model

Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu

Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 10 Pages, 2024/05

Development of element functions and design optimization procedures for knowledge- and AI-aided advanced reactor lifecycle optimization method, ARKADIA

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

Mechanical Engineering Journal (Internet), 11(2), p.23-00424_1 - 23-00424_13, 2024/04

https://doi.org/10.1299/mej.23-00424

The outline and development status of element functions and design optimization process in ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source are introduced. It is also briefly explained that 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, and merge state-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&Ds with AI technologies.

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.

The Development of a Multiphysics Coupled Solver for Studying the Effect of Dynamic Heterogeneous Configuration on Particulate Debris Bed Criticality and Cooling Characteristics

Li, C.-Y.; Wang, K.*; Uchibori, Akihiro; Okano, Yasushi; Pellegrini, M.*; Erkan, N.*; Takata, Takashi*; Okamoto, Koji*

Applied Sciences (Internet), 13(13), p.7705_1 - 7705_29, 2023/07

https://doi.org/10.3390/app13137705

Experimental study on the breakup of liquid jet discharged from a nozzle with sudden contraction

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

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05

When liquid sodium leaks as a liquid jet from piping in a sodium-cooled fast reactor, the jet impinges with structures to produce splashing droplets which can cause significant combustion. According to previous studies on circular nozzles, the amount of splash is affected by the state of the jet at the moment of impingement. In the present work, a nozzle with a sudden contraction in the cross-sectional area was designed to reproduce a supposed pipe leakage, and the breakup behavior of jet discharged from this nozzle was observed. The result shows that the breakup of jet was accelerated until the jet transformed into a particularly stable state when the jet velocity exceeded a certain value. Once the jet has transformed, it will not turn back unless turning down the flow rate to a very low value. The stable jet after the transformation has a longer breakup length than that before the transformation in the same flow rate.

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

In-vessel source term analysis code TRACER Version 2.4.1 (User's manual)

Ono, Masahiro*; Uchibori, Akihiro; Okano, Yasushi; Takata, Takashi*

JAEA-Testing 2022-004, 193 Pages, 2023/03

JAEA-Testing-2022-004.pdf:3.31MB

A computer code TRACER (Transport phenomena of Radionuclides for Accident Consequence Evaluation of Reactor) version 2.4.1 has been developed to evaluate species and quantities of fission products (FPs) released into cover gas due to a fuel pin failure in an LMFBR. The TRACER version 2.4.1 includes the models related to NUREG-0772 and also new or modified computational program codes in order to possess a new function shown below, and partial modify of coefficient of FP transition model between coolant and cover gas. This manual includes manual conventions for TRACER Version 2.3, addition of reference such as formula, improvement of explanation of input file creation method, addition of improvement of NUREG-0772 model added to TRACER code, modification of figure of sample analysis performed in appendix. It includes modifications and additions of sample analysis.