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/

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.

A Comparative study of sampling techniques for dynamic probabilistic risk assessment of nuclear power plants

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

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.308 - 315, 2020/10

Dynamic probabilistic risk assessment (PRA) is a method for improving the realism and completeness of conventional PRA. However, enormous calculation costs are incurred by these improvements. One solution is to select an appropriate sampling method. In this paper, we applied the Monte Carlo, Latin hypercube, grid-point, and quasi-Monte Carlo sampling methods to the dynamic PRA of a simplified accident sequence and compared the results for each method. Quasi-Monte Carlo sampling was found to be the most effective method in this case.

Analysis for the accident at unit 2 of the Fukushima Daiichi NPS with THALES2/KICHE code in BSAF2 project

Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.100 - 111, 2019/08

JAEA participated in the OECD/NEA BSAF2 project with our integrated severe accident analysis code, THALES2/KICHE, in order to analyze and discuss the accident progression and source term of the accident at the Fukushima Daiichi NPS. One of important characteristics of THALES2/KICHE code is that it has the capability of predicting iodine chemistry based on reaction kinetics in the aqueous phase. JAEA performed the three week analysis for the accident at unit 2 on the basis of the boundary conditions and assumptions proposed by the BSAF2 project and our own assumptions. One of focusing points in the BSAF2 project was the trend of measured data of reactor vessel from 20:00 March 14 to 02:00 March 15. An assumption was made that the lower part of the suppression chamber failed to form a water leakage path. The released iodine and cesium within three weeks after the earthquake were predicted to be approximately 3% and 0.1% of the initial inventory, respectively.

Sensitivity analysis of source term in the accident of Fukushima Dai-ichi Nuclear Power Station Unit 1 using THALES2/KICHE

Tamaki, Hitoshi; Ishikawa, Jun; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Asian Symposium on Risk Assessment and Management 2018 (ASRAM 2018) (USB Flash Drive), 6 Pages, 2018/10

In the accidents at Fukushima Dai-ichi Nuclear Power Station, Tsunami caused loss of electric power supply and this event led to core melt and failure of Containment vessel. Finally, fission products were released to the environment. Currently, the activities for understanding of accident progressions are carried out based on the measured data during the accident, accident progression analysis using integrated severe accident analysis codes and investigation of inside of reactor buildings and containment vessels. On the other hand, there are some research activities with combination of accident progression analysis and accident consequence analysis. In Japan Atomic Energy Agency (JAEA), the research project of combination of these analyses using the computational simulation codes has been started. The results obtained from the combination analysis are expected to have broad width of uncertainty because of many uncertainty factors in this combined analysis. In order to perform the analysis efficiently, sensitivity analysis for failure location on containment vessel and its failure size were carried out by THALES2/KICHE developed by JAEA at first. This analysis was performed on unit 1, since it was the first plant to release radioactive materials to the environment during the accident and its consequence had no effect from other plants. The authors focused on the failure of containment vessel head flange, penetration seal and vacuum breaker pipe, and possibility of partial open of vent valve based on the investigations of reactor building inside performed by TEPCO. This paper presents the results obtained from this sensitivity analysis.

Study on characterisation of colloidal silica grout under condition of sea water

Toguri, Satohito*; Okihara, Mitsunobu*; Tsuji, Masakuni*; Nakashima, Hitoshi*; Sugiyama, Hirokazu*; Saito, Akira*; Sato, Toshinori; Aoyagi, Kazuhei; Masunaga, Kosuke

JAEA-Research 2017-013, 131 Pages, 2018/02

JAEA-Research-2017-013.pdf:8.49MB

The discussions on scientifically promising site for the geological disposal has been made at the council of studying group on techniques for geological disposal of radioactive wastes, which is held by Resources and Energy Agency. From the aspect of ensuring safety during the transportation of disposal waste, the coastal area is discussed to be a more suitable area. This report shows the result of the first year of this project as following items; Study on the state-of-art technology and remain tasks; laboratory tests on characterization of colloidal silica grout under sea water; Study on the development of grouting technology (design and the evaluation method of influence on the rock mass).

Current status of research for the accident of evaporation to dryness caused by boiling of reprocessed high level radioactive liquid waste

Tamaki, Hitoshi; Yoshida, Kazuo; Abe, Hitoshi; Sugiyama, Tomoyuki; Maruyama, Yu

Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 9 Pages, 2017/11

An accident of evaporation to dryness caused by boiling of high level radioactive liquid waste (HLLW) is postulated as one of severe accidents caused by the loss of cooling function at the fuel reprocessing plant. This accident can be divided into early boiling stage, late boiling stage and dry-out stage by characteristics of accident evolution. It is important to estimate the amount of fission product (FP) transport between the liquid and gas phases, and the amount of FP deposition on the walls in each stage in order to estimate the release amount of FP to the environment. Various research activities have been carried out for this issue. This paper reviews these activities and presents the recent activities at JAEA for development of simulation code for this type of accident.

Development of hydrogen behavior simulation code system; Outline of code system and validation using existing data

Terada, Atsuhiko; Matsumoto, Masaaki*; Sugiyama, Hitoshi*; Kamiji, Yu; Kadowaki, Satoshi*; Hino, Ryutaro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

In the Fukushima Daiichi Nuclear Power Station (NPS) accident, hydrogen was generated by oxidation reaction of the cladding and water etc, then leaked into the NPS building, and finally led to occurrence of hydrogen explosion in the building. This resulted in serious damage to the environment. To improve the safety performance of the NPS, especially on the hydrogen safety under severe accident conditions, a simulation code system has been developed to analyze hydrogen behaviors including diffusion, combustion, explosion and structural integrity evaluation. This developing system consists of CFD and FEM tools in order to support various hydrogen user groups of students, researchers and engineers. Preliminary calculated results obtained with above mentioned tools, damage of piping induced by hydrogen combustion, agreed well with existing test data.

Numerical analysis of turbulent flow in a rectangular duct containing straight cylinder placed near a bottom wall

Sugiyama, Hitoshi*; Otemori, Toshiki*; Kato, Naoto*; Terada, Atsuhiko; Kamiji, Yu; Hino, Ryutaro

Jidosha Gijutsukai Gakujutsu Koenkai Zensatsushu, (147-14), p.25 - 30, 2014/10

Numerical analysis has been performed for three-dimensional developing turbulent flow in a rectangular duct containing straight cylinder placed near a bottom wall. In this kind of turbulent flow, it has been pointed out as a characteristic features that anisotropic turbulence generates the secondary flow of the second kind and pulsation flow is measured in narrow gap region located between cylinder and bottom wall. Comparisons with the experimental results suggest that the present method is able to predict velocity profiles correctly and reproduce both the secondary flow of the second kind and the generation of pulsation flow near the narrow region between cylinder and bottom wall.

Numerical analysis of turbulent flow in periodically arranged ribs

Okagaki, Yuria*; Sugiyama, Hitoshi*; Kato, Naoto*; Hino, Ryutaro

Jidosha Gijutsukai Rombunshu, 43(4), p.949 - 955, 2012/07

https://doi.org/10.11351/jsaeronbun.43.949

Turbulent heat transfer enhancement using periodically arranged ribs processed on the graphite sleeve of a fuel rod aims to increase heat generation density of the fuel rod, which can potentially improve the economics of a block type high-temperature gas-cooled reactor. Rectangular cross section of rib is selected because of easy processing. Furthermore, it is important to find optimum rib topology such as pitch and height to maximize turbulent heat transfer performance. In order to develop a turbulent model as a design tool for rib-roughened fuel rod, numerical analysis has been conducted to verify applicability of an algebraic Reynolds stress model, which is computationally efficient, to non-isotropic turbulent flow in rib roughened channels. The results of the analysis on a square duct with periodically arranged ribs on the bottom wall are shown to reproduce very well the flow characteristics such as flow separation and reattachment behind the ribs. The model is therefore expected to be applicable to rib-enhanced fuel rod design.