Main outputs from the OECD/NEA ARC-F Project

Maruyama, Yu; Sugiyama, Tomoyuki*; Shimada, Asako; Lind, T.*; Bentaib, A.*; Sogalla, M.*; Pellegrini, M.*; Albright, L.*; Clayton, D.*

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.4782 - 4795, 2023/08

http://dx.doi.org/10.13182/NURETH20-40302

Improvement of JASMINE code for ex-vessel molten core coolability in BWR

Matsumoto, Toshinori; Kawabe, Ryuhei*; Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Maruyama, Yu

Annals of Nuclear Energy, 178, p.109348_1 - 109348_13, 2022/12

https://doi.org/10.1016/j.anucene.2022.109348

The Japan Atomic Energy Agency extended the applicability of their fuel-coolant interaction analysis code JASMINE to simulate the relevant phenomena of molten core in a severe accident. In order to evaluate the total coolability, it is necessary to know the mass fraction of particle, agglomerated and cake debris and the final geometry at the cavity bottom. An agglomeration model that considers the fusion of hot particles on the cavity floor was implemented in the JASMINE code. Another improvement is introduction of the melt spreading model based on the shallow water equation with consideration of crust formation at the melt surface. For optimization of adjusting parameters, we referred data from the agglomeration experiment DEFOR-A and the under-water spreading experiment PULiMS conducted by KTH in Sweden. The JASMINE analyses reproduced the most of the experimental results well with the common parameter set, suggesting that the primary phenomena are appropriately modelled.

Revolatilization of iodine by bubbly flow in the suppression pool during an accident

Nanjo, Kotaro; Ishikawa, Jun; Sugiyama, Tomoyuki; Pellegrini, M.*; Okamoto, Koji*

Journal of Nuclear Science and Technology, 59(11), p.1407 - 1416, 2022/11

https://doi.org/10.1080/00223131.2022.2062065

Uncertainty analysis of dynamic PRA using nested Monte Carlo simulations and multi-fidelity models

Zheng, X.; Tamaki, Hitoshi; Takahara, Shogo; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Probabilistic Safety Assessment and Management (PSAM16) (Internet), 10 Pages, 2022/09

https://www.iapsam.org/PSAM16/

Simulation-based dynamic probabilistic risk assessment of an internal flooding-initiated accident in nuclear power plant using THALES2 and RAPID

Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Proceedings of the Institution of Mechanical Engineers, Part O; Journal of Risk and Reliability, 11 Pages, 2022/04

https://doi.org/10.1177/1748006X221091604

Probabilistic risk assessment (PRA) is a method used to assess the risks associated with large and complex systems. However, the timing at which nuclear power plant structures, systems, and components are damaged is difficult to estimate if the risk of an external event is evaluated using conventional PRA based on event trees and fault trees. A methodology coupling thermal-hydraulic analysis with external event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID) is therefore proposed to overcome this limitation. A flood propagation model based on Bernoulli's theorem was applied to represent internal flooding in the turbine building of the pressurized water reactor. Uncertainties were also taken into account, including the flow rate of the floodwater source and the failure criteria for the mitigation systems. The simulated recovery actions included the operator isolating the floodwater source and using a drainage pump; these actions were modeled using several simplifications. Overall, the results indicate that combining isolation and drainage can reduce the conditional core damage probability upon the occurrence of flooding by approximately 90%.

Radiochemical analysis of the drain water sampled at the exhaust stack shared by Units 1 and 2 of the Fukushima Daiichi Nuclear Power Station

Shimada, Asako; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Ohira, Saki; Iida, Yoshihisa; Sugiyama, Tomoyuki; Amaya, Masaki; Maruyama, Yu

Scientific Reports (Internet), 12(1), p.2086_1 - 2086_11, 2022/02

https://doi.org/10.1038/s41598-022-05924-2

no abstracts in English

Numerical analysis for FP speciation in VERDON-2 experiment; Chemical re-vaporization of iodine in air ingress condition

Shiotsu, Hiroyuki; Ito, Hiroto*; Sugiyama, Tomoyuki; Maruyama, Yu

Annals of Nuclear Energy, 163, p.108587_1 - 108587_9, 2021/12

https://doi.org/10.1016/j.anucene.2021.108587

Development of evaluation framework for ex-vessel core coolability

Matsumoto, Toshinori; Iwasawa, Yuzuru; Sugiyama, Tomoyuki

Proceedings of Reactor core and Containment Cooling Systems, Long-term management and reliability (RCCS 2021) (Internet), 8 Pages, 2021/10

A methodological framework is being developed in JAEA for evaluating debris coolability at ex-vessel during the severe accident (SA) of BWR under the wet cavity strategy. The probability of ex-vessel debris coolability under the wet cavity strategy is analyzed to demonstrate the evaluation approach. Probabilistic distribution of the melt conditions ejected from the RPV was obtained as the result of the iterative analyses with MELCOR code. Five uncertainty parameters relating with the core degradation and transfer process were chosen. Parameter sets were generated by Latin hypercube sampling (LHS). JASMINE code plays the physical model to predict the mass fraction of agglomerated debris and melt pool spreading on the floor. Fifty-nine input parameter set for JASMINE code were generated by LHS again using the probabilistic distribution of melt condition determined from the results of MELCOR analyses. The depth of the water pool was set as 0.5, 1.0 and 2.0 m. The accumulated debris height was compared with the criterion to judge the debris coolability. As the result, the success probability of debris cooling was obtained through the sequence of calculations.

Dynamic PRA of flooding-initiated accident scenarios using THALES2-RAPID

Kubo, Kotaro; Zheng, X.; Tanaka, Yoichi; Tamaki, Hitoshi; Sugiyama, Tomoyuki; Jang, S.*; Takata, Takashi*; Yamaguchi, Akira*

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2279 - 2286, 2020/11

https://www.rpsonline.com.sg/proceedings/esrel2020/pdf/4717.pdf

Probabilistic risk assessment (PRA) is one of the methods used to assess the risks associated with large and complex systems. When the risk of an external event is evaluated using conventional PRA, a particular limitation is the difficulty in considering the timing at which nuclear power plant structures, systems, and components fail. To overcome this limitation, we coupled thermal-hydraulic and external-event simulations using Risk Assessment with Plant Interactive Dynamics (RAPID). Internal flooding was chosen as the representative external event, and a pressurized water reactor plant model was used. Equations based on Bernoulli's theorem were applied to flooding propagation in the turbine building. In the analysis, uncertainties were taken into account, including the flow rate of the flood water source and the failure criteria for the mitigation systems. In terms of recovery action, isolation of the flood water source by the operator and drainage using a pump were modeled based on several assumptions. The results indicate that the isolation action became more effective when combined with drainage.

Enhancement of the treatment of system interactions in a dynamic PRA tool

Tanaka, Yoichi; Tamaki, Hitoshi; Zheng, X.; Sugiyama, Tomoyuki

Proceedings of 30th European Safety and Reliability Conference and 15th Probabilistic Safety Assessment and Management Conference (ESREL 2020 and PSAM-15) (Internet), p.2195 - 2201, 2020/11

https://www.rpsonline.com.sg/proceedings/esrel2020/html/3854.xml