Analysis of fuel assemblies inclination due to upper core support plate deflection for reactivity evaluation

Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Uwaba, Tomoyuki; Tanaka, Masaaki; Nemoto, Toshiyuki*

Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 8 Pages, 2024/03

To investigate possibility of the insertion of the reactivity by the deflection of the upper core support plate, structural mechanics analyses of the domain consisting of the fuel assemblies and core support plates and evaluation of the reactivity due to the inclination of the fuel assemblies in EBR-II were carried out. As a result, it was indicated that the upper core support plate deflected downward larger at the low flowrate condition than that at the high flowrate condition and positive reactivity was inserted due to the inclination of the fuel assemblies at the low flowrate condition.

Development of virtual plant model for design rationalization of fast reactors by multi-level simulation system; Confirmation of functionality in application to U.S. experimental fast reactor EBR-II

Yoshimura, Kazuo; Doda, Norihiro; Nakamine, Yoshiaki*; Fujisaki, Tatsuya*; Igawa, Kenichi*; Iida, Masaki*; Tanaka, Masaaki

Dai-27-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2023/09

In Japan Atomic Energy Agency, a virtual plant model of the sodium-cooled fast reactor plant composed in a computer is being developed to reduce the development cost, by replacing the experiments to the numerical simulations with coupled analyses of the physical phenomena accounting for the interaction between components under various plant conditions. Through the numerical analysis of the ULOHS test in the U.S. experimental fast reactor named EBR-II, applicability of the virtual plant model was confirmed in comparison with the measured data including the core inlet temperature and the reactor power.

Orientation mapping of YbSn single crystals based on Bragg-dip analysis using a delay-line superconducting sensor

Shishido, Hiroaki*; Vu, TheDang*; Aizawa, Kazuya; Kojima, Kenji M*; Koyama, Tomio*; Oikawa, Kenichi; Harada, Masahide; Oku, Takayuki; Soyama, Kazuhiko; Miyajima, Shigeyuki*; et al.

Journal of Applied Crystallography, 56(4), p.1108 - 1113, 2023/08

https://doi.org/10.1107/S1600576723005204

Application of quasi-Monte Carlo and importance sampling to Monte Carlo-based fault tree quantification for seismic probabilistic risk assessment of nuclear power plants

Kubo, Kotaro; Tanaka, Yoichi; Hakuta, Yuto*; Arake, Daisuke*; Uchiyama, Tomoaki*; Muramatsu, Ken

Mechanical Engineering Journal (Internet), 10(4), p.23-00051_1 - 23-00051_17, 2023/08

https://doi.org/10.1299/mej.23-00051

The significance of probabilistic risk assessments (PRAs) of nuclear power plants against external events was re-recognized after the Fukushima Daiichi Nuclear Power Plant accident. Regarding the seismic PRA, handling correlated failures of systems, components, and structures (SSCs) is very important because this type of failure negatively affects the redundancy of accident mitigation systems. The Japan Atomic Energy Research Institute initially developed a fault tree quantification methodology named the direct quantification of fault tree using Monte Carlo simulation (DQFM) to handle SSCs' correlated failures in detail and realistically. This methodology allows quantifying the top event occurrence probability by considering correlated uncertainties related to seismic responses and capacities with Monte Carlo sampling. The usefulness of DQFM has already been demonstrated. However, improving its computational efficiency would allow risk analysts to perform several analyses. Therefore, we applied quasi-Monte Carlo and importance sampling to the DQFM calculation of simplified seismic PRA and examined their effects. Specifically, the conditional core damage probability of a hypothetical pressurized water reactor was analyzed with some assumptions. Applying the quasi-Monte Carlo sampling accelerates the convergence of results at intermediate and high ground motion levels by an order of magnitude over Monte Carlo sampling. The application of importance sampling allows us to obtain a statistically significant result at a low ground motion level, which cannot be obtained through Monte Carlo and quasi-Monte Carlo sampling. These results indicate that these applications provide a notable acceleration of computation and raise the potential for the practical use of DQFM in risk-informed decision-making.

Development of a statistical evaluation method for core hot spot temperature in sodium-cooled fast reactor under natural circulation conditions

Doda, Norihiro; Igawa, Kenichi*; Iwasaki, Takashi*; Murakami, Satoshi*; Tanaka, Masaaki

Nuclear Engineering and Design, 410, p.112377_1 - 112377_15, 2023/08

https://doi.org/10.1016/j.nucengdes.2023.112377

To enhance the safety of sodium-cooled fast reactors, the decay heat in the core must be removed by natural circulation even if the AC power supply to the forced circulation equipment is lost. Under natural circulation conditions, sodium flow is driven by buoyancy, and flow velocity and temperature distribution influence each other. Thus, it is difficult to evaluate the core hot spot temperature by deterministically considering the uncertainties affecting flow and heat. In this study, a statistical evaluation method is developed for the core hot spot temperature by using Monte Carlo sampling methods. The applicability of the core hotspot evaluation method was confirmed in three representative events during natural circulation decay heat removal operations in loop-type sodium-cooled fast reactors.

Improvement of reactivity model of core deformation in plant dynamics analysis code during unprotected loss of heat sink event in EBR-II

Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

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

The benchmark analyses for the unprotected loss of heat sink (ULOHS) tests in the pool-type experimental SFR in the United States, EBR-II (BOP-301 and BOP-302R) have been conducted in order to validate the evaluation method of the reactivity feedback equipped in the plant dynamics analysis code named Super-COPD. In this study, 1D-CFD coupled analyses adding the core bowing reactivity model were conducted. Through the analysis, the applicability of the modified reactivity model was confirmed for the BOP-301 test. For the BOP-302R test, consideration of the core restraint system in the core and modeling the control rod driveline expansion reactivity was indicated.

Development plan for coupling technology between high temperature gas-cooled reactor HTTR and Hydrogen Production Facility, 2; Development plan for coupling equipment between HTTR and Hydrogen Production Facility

Mizuta, Naoki; Morita, Keisuke; Aoki, Takeshi; Okita, Shoichiro; Ishii, Katsunori; Kurahayashi, Kaoru; Yasuda, Takanori; Tanaka, Masato; Isaka, Kazuyoshi; Noguchi, Hiroki; et al.

Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 6 Pages, 2023/05

Validation of feedback reactivity evaluation models for plant dynamics analysis code during unprotected loss of heat sink event in sodium-cooled fast reactors

Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Journal of Nuclear Engineering and Radiation Science, 9(2), p.021601_1 - 021601_9, 2023/04

https://doi.org/10.1115/1.4055328

Feedback reactivity automatically caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In order to validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD, the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of BOP-302R and BOP-301 in an experimental SFR, EBR-II were conducted and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS even, by comparing the numerical results and the experimental data.

Application of 1D-CFD coupling method to unprotected loss of heat sink event in EBR-II focusing on thermal stratification in cold pool

Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08

To confirm the applicability of the reactivity model, the authors have been conducting the benchmark exercises of the unprotected loss of heat sink event tests in a pool-type experimental fast reactor EBR-II. In the blind phase in the benchmark analyses using the plant dynamics analysis (1D) code in which the cold pool was modeled by means of the perfect mixing volume, it was found the increase of the core inlet temperature was evaluated lower than that of the measured data and the feedback reactivity was underestimated, because the thermal stratification in the cold pool was ignored. Then, the detailed model of the cold pool for the computational fluid dynamics (CFD) code was introduced and the 1D-CFD codes coupling method was applied to the benchmark analyses. It was confirmed that both the thermal stratification in the cold pool and the increase of the core inlet temperature were successfully reproduced.

Application of first-order method to estimate structural integrity in a probabilistic form of component subjected to thermal transient for optimization of design parameter

Okajima, Satoshi; Mori, Takero; Kikuchi, Norihiro; Tanaka, Masaaki; Miyazaki, Masashi

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08

In this paper, we propose the simplified procedure to estimate failure probability of components subjected to thermal transient for the design optimization. Failure probability can be commonly used as an indicator of component integrity for various failure mechanisms. In order to reduce number of analyses required for one estimation, we have adopted the First Order Second Moment (FOSM) method as the estimation method of failure probability on the process of the optimization, and an orthogonal table in experiment design method is utilized to define conditions of the analyses for the evaluation of the input parameters for the FOSM method. Through the demonstration study to optimize thickness of cylindrical vessel subjected to thermal transient derived from shutdown, we confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation cost.