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Ono, Ayako; Yamashita, Susumu; Sakashita, Hiroto*; Suzuki, Takayuki*; Yoshida, Hiroyuki
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
JAEA is implementing a simulation of a two-phase flow in the reactor core by TPFIT and JUPITER which are developed by JAEA based on the surface tracking method. However, it is impossible to simulate a boiling on the heating surface in the large-scale domain by this type of simulation method since the simulation of boiling based on the surface tracking method needs the fine mesh which sufficiently resolves the initiation of boiling. Therefore, JAEA started to develop the simplified boiling model applied for the two-phase flow in the fuel assemblies. In this study, the simulation results of the convection boiling on a vertical heating surface and the comparison between the simulation results and experimental results are shown.
Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Akayev, A. S.*; Mikisha, A. V.*; Baklanov, V. V.*; Vurim, A. D.*
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
The cooling of the residual core materials after the fuel discharge from the SFR core in the core disruptive accident can significantly affect the distribution fraction of the core materials which is an important factor for the in-vessel retention (IVR). The cooling of the residual core materials is called "in-place cooling". For the evaluation of the in-place cooling, behavior in a SFR core was simulated by SIMMER-III, and method of phenomena identification and ranking table (PIRT) was applied based on the analysis result. Experiment which focuses on the thermal-hydraulic phenomena which were extracted by the PIRT was conducted in the framework of EAGLE-3 project. Continuous oscillation of sodium level which can occur in the phase of in-place cooling of SFRs was observed in the experiment, and analysis by the SIMMER-III was conducted. By investigation of the analysis result, difference between the experiment and analysis results was revealed to be due to remaining and occupation of non-condensable gas above the sodium level which would be unrealistic in the experiment. Gas mixture model between non-condensable gas and sodium vapor was developed to solve this problem, and coincidence between experiment and analysis results was largely improved by this new model.
Emura, Yuki; Kamiyama, Kenji; Yamano, Hidemasa
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
no abstracts in English
Kikuchi, Norihiro; Mori, Takero; Okajima, Satoshi; Tanaka, Masaaki; Miyazaki, Masashi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
In JAEA, the design optimization method for plant structure has been developed on the process to output optimal solution of the thickness of reactor vessel wall against thermal transient and seismic loads in a SFR as a representative problem. Resistance characteristic of the wall on the load derived from thermal transient is one of the most important factors for safety estimation on the structural integrity. Failure probability of components against thermal transient was set to one of variables in the objective function for a common scale to compare with other variables in different failure mechanisms. In the iterative process to achieve the optimal solution, a number of evaluations to measure the influence on the load derived from thermal transient was necessarily conducted. More reduction of required time for evaluations is desired. To perform the iterative evaluation process efficiently, the automatization of parametric analyses was implemented in the optimization process.
Li, C.-Y.; Watanabe, Akira*; Uchibori, Akihiro; Okano, Yasushi
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2022/07
Identifying accident scenarios that could lead to severe accidents and evaluating their frequency of occurrence are essential issues. This study aims to establish the methodology of the dynamic Probabilistic Risk Assessment (PRA) for sodium-cooled fast reactors that can consider the time dependency and the interdependence of each event. Specifically, the Continuous Markov chain Monte Carlo (CMMC) method is newly applied to the SPECTRA code, which analyzes the severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. Currently, a fault-tree model of air coolers of decay heat removal system is implemented as the CMMC method, and a series of preliminary analysis of the plant's transient characteristics under the scenario of volcanic ashfall has been conducted.
Tanaka, Masaaki; Doda, Norihiro; Yokoyama, Kenji; Mori, Takero; Okajima, Satoshi; Hashidate, Ryuta; Yada, Hiroki; Oki, Shigeo; Miyazaki, Masashi; Takaya, Shigeru
Dai-26-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 5 Pages, 2022/07
To assist conceptual studies of various reactor systems conducted by private sectors in nuclear power innovation, development of an innovative design system named ARKADIA (Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle) is undergoing to achieve the design of an advanced nuclear reactor as a safe, economic, and sustainable carbon-free energy source. In this paper, focusing on the ARKADIA-Design as a part of it, the progress in the development of optimization processes on the representative problems in the fields of the core design, the plant structure design, and the maintenance schedule planning are introduced.
Okamura, Shigeki*; Kinoshita, Takahiro*; Yamano, Hidemasa; Kurisaka, Kenichi; Futagami, Satoshi; Fukasawa, Tsuyoshi*
no journal, ,
The seismic evaluation of key components such as the reactor vessel is important for the seismic probabilistic risk assessment in sodium-cooled fast reactors. In the past, the correlation of the fatigue failure and the integrated vibration energy to the component was applied to the failure evaluation in the seismic assessment. The objective of this study is to develop a failure probability evaluation method, which evaluates integrated vibration energy to the component during an earthquake. The fatigue failure evaluation method is investigated with simple vibration examination to failure. The vibration examination with the simple examination specimens, of which main vibration mode is the bending vibration, has been carried out by the random wave. This paper describes the estimation of the integrated vibration energy with the random waves. The integrated vibration energy of the random wave was compared with the integrated vibration energy of the sine wave by the integrated energy per unit time. The integrated vibration energy of the random wave is in the range of the approximate line, which is estimated with the integrated vibration energy of the sine wave. The evaluation method investigated in this study is expected to estimate the integrated energy with the random wave by the integrated energy with the sine wave.
Yamamoto, Tomohiko; Yan, X.; Shimada, Takahiro*; Motegi, Haruki*; Kai, Satoru*; Otani, Akihito*
no journal, ,
no abstracts in English
Sawa, Kazuhiro*; Haseda, Masaya*; Aihara, Jun
no journal, ,
In high temperature gas-cooled reactors (HTGRs), Tri-isotropic (TRISO)-coated fuel particles are employed as fuel. In the high burnup coated fuel particle, stress due to fission gas pressure and irradiation-induced pyrolytic carbon (PyC) shrinkage is introduced into the coating layers and consequently the stress could cause failure of coating layers under high burnup irradiation condition. A failure model has developed to predict failure fraction of TRISO-coated particle under high burnup irradiation. In the model, failure probability is strongly dependent on the irradiation characteristics of PyC. This paper describes the outline of the failure model and evaluation result of high burnup fuel irradiation experiment by the model.
Morita, Keisuke; Shimizu, Atsushi; Noguchi, Hiroki; Aoki, Takeshi; Okita, Shoichiro; Mizuta, Naoki; Ishii, Katsunori; Iigaki, Kazuhiko; Sato, Hiroyuki; Sakaba, Nariaki
no journal, ,
no abstracts in English
Yoshimura, Kazuo; Doda, Norihiro; Hamase, Erina; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki
no journal, ,
Sodium-cooled fast reactors have intrinsic safety features decreasing reactor power during the increase of the core inlet temperature by the feedback reactivity of the radial expansion of the core support plate. It is necessary for the composition of the core highly of secure to understand the influence of the safety features with high accuracy. In this paper, first, the 1D-CFD coupling method with cold pool as CFD region which enables the plant dynamics analyses taking account of the thermal stratification in cold pool was applied to the ULOHS (Unprotected Loss Of Heat Sink) test performed in the experimental fast reactor U.S. EBR-II and the evaluation of the core inlet temperature could be improved. Secondly, the sensitivity analyses concerning the core bowing reactivity were carried out with the aim of improving the evaluations of the core deformation reactivity and the applicability of the core bowing reactivity model to the test could be indicated.
Tomita, Naoki*; Furuya, Masahiro*; Asahi, Manabu*; Hisamochi, Rikuya*; Toyota, Kodai; Yada, Hiroki
no journal, ,
Ultrasonic phased array is a phase composite imaging technology developed in the radar field and has recently been used for nondestructive inspection of power generation equipment. However, scattered waves in the inspection target make it difficult to distinguish a flaw from scattering from the edge surface. In this study, we developed a method to discriminate flaws with high accuracy by adjusting parameters such as output voltage and receiver sensitivity to make it easier to see the flaws and by deep learning of the optimized flaw images. First, actual measurements were made using an ultrasonic phased-array flaw detector on a stainless-steel specimen. Next, a model was created to discriminate the presence or absence of flaws using transition learning, one of the machine learning methods. As a result, we found that the highest accuracy was achieved when transition learning was performed using inceptionv3 and resnet101, a convolutional neural network architecture. These results show that the method developed in this study is effective for nondestructive inspection.
Onoda, Yuichi; Yamano, Hidemasa
no journal, ,
In Japan, sodium-cooled fast reactor (SFR) design takes In-Vessel Retention (IVR) strategy to stable cooling the damaged core material in the reactor vessel against severe accidents by various design measures. Although a possibility to fail IVR is extremely unlikely, a probabilistic risk assessment study needs a wide variety of scenarios including the IVR failure. Therefore, in order to study of a wide range of event spectra related to stable cooling of debris in the reactor vessel, the deformation and failure behavior of the reactor vessel due to the debris deposited on the bottom of the reactor vessel are investigated in this study using the FINAS-STAR. Structures including reactor vessel and guard vessel which locate below the core catcher at the bottom of the reactor vessel are modeled. The amount of debris falling to the bottom of the reactor vessel is tentatively determined to 30% of the core inventory. The analyses are conducted in two cases of power density with the aim of establishing conditions for failure the bottom of the reactor vessel. Reactor vessel deforms significantly when the temperature reaches about 1100
C and the reactor vessel failed in both calculation cases in the analysis of current conditions.
