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Yamashita, Susumu; Uesawa, Shinichiro; Ono, Ayako; Yoshida, Hiroyuki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10
no abstracts in English
Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10
Horiguchi, Naoki; Yoshida, Hiroyuki; Kaneko, Akiko*; Abe, Yutaka*
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 6 Pages, 2022/10
For safety evaluation of nuclear reactors in severe accidents, it is important to estimate physical quantities of fragments generated from the molten fuel jet, which falls in a pool and breaks up. The evaluation method has been developed for the behavior as liquid jet with hydrodynamic interaction including fuel coolant interaction (FCI). In case of a shallow pool assumed in ex-vessel, the molten fuel jet is assumed to behave as wall-impinging liquid jet and to form liquid film flow spreading on the floor with/without fragmentation. In our research, focusing on hydrodynamic interaction and the transient 3-dimensional spreading on the floor, we have developed the evaluation method by numerical simulation using the two-phase flow simulation code with interface tracking method (TPFIT) developed by JAEA and, the experimental method using the 3D-LIF method in liquid-liquid system for the validation data. In our previous studies, we investigated the wall-impinging liquid jet behavior with fragmentation and observed that the liquid film flow had some characteristic parts transiently. Since it indicates that the quantities change depending on the parts and affect the safety evaluation, it is important to measure the quantities of the fragments generated from each part. This paper explains the measurement of the physical quantities of the fragments generated from each part of the wall-impinging liquid jet in a shallow pool for the validation of the numerical simulation. We conducted an experiment with the 3D-LIF method and segmented the experimental data based on the fragmentation point over the liquid film flow using the dispersed phase tracking method, developed by JAEA. Then, we measured the diameter and amount of the fragments from the segmented experimental data and investigated their changing trend.
Nakamura, Hironori*; Hayakawa, Satoshi*; Shibata, Akihiro*; Sasa, Kyohei*; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10
In order to evaluate long-term coolablity of the debris-bed with decay heat, a three-dimensional calculation method coupled with the debris bed module was developed in this study. The coupled code calculation results show that natural circulation of the coolant between the hot pool and the cold pool is established through the four intermediate heat exchangers after the activation of the dipped direct heat exchangers. The cold pool with the debris-bed is continually cooled not only by the natural circulation flow, but also by heat transfer to the hot pool through the plenum separation plate between the hot pool and the cold pool. The effect of the three-dimensional flow field around the core catcher on the temperature in the debris-bed is about 20K under the current calculation condition.
Tobita, Daiki*; Monji, Hideaki*; Yamashita, Susumu; Horiguchi, Naoki; Yoshida, Hiroyuki; Sugawara, Takanori
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 5 Pages, 2022/10
Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 9 Pages, 2022/10
To improve the prediction accuracy of the plant dynamics analysis code named Super-COPD, JAEA has joined the IAEA benchmark for the FFTF Loss of Flow Without Scram Test No.13. In the first blind phase, there was the challenge to perform outlet temperatures of fuel assemblies more accurately. Hence, the renewed analysis was performed with the whole core multi-channel model in which each assembly was modelled to simulate the radial heat transfer among assemblies and the flow redistribution induced by the buoyancy in the NC conditions. Then, to validate the coupled transient analysis between the whole core multi-channel model and the one-point kinetics model, the analysis considering major reactivity feedbacks such as GEM, assembly bowing was performed. As a result, the second peak of outlet temperatures was reproduced successfully, and it was observed that the plant dynamics analysis could follow the measured data.
Doda, Norihiro; Kato, Shinya; Iida, Masaki*; Yokoyama, Kenji; Tanaka, Masaaki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10
In the conventional core design in sodium-cooled fast reactors (SFRs), negative reactivity feedback due to core deformation was neglected because of large uncertainty in analytical evaluation. To optimize core design, it is necessary to develop an analytical evaluation method and eliminate excessive conservativeness. An evaluation method for core deformation reactivity has been developed by coupling analysis of neutronics, thermal-hydraulics, and structural mechanics. For the verification study of the neutronics calculation method, the reactivity was calculated for the deformed core geometry in which the fuel assembly (FA) moved horizontally in the radial direction for each row from the core center. Compared to reference values derived from Monte Carlo calculations, the calculated reactivity due to FA displacement agreed well in the core region and was overestimated in the reflector region. The modification challenges in development of the core deformation model were identified.
Aizawa, Kosuke; Tsuji, Mitsuyo; Kobayashi, Jun; Kurihara, Akikazu
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 7 Pages, 2022/10
In sodium-cooled fast reactors (SFRs), optimizing the design and operate decay heat removal systems (DHRSs) is important for safety enhancement against severe accidents. Thus, it is required to evaluate the cooling capability of DHRSs including the natural circulation behavior inside the reactor vessel during heat-removal phase that the fuel debris relocated in the reactor vessel is cooled by DHRSs. In this study, the experiments which simultaneously operations of the dipped-type DHX and the penetrated-type DHX were conducted to investigate the effect of operating multiple decay heat removal system on the natural circulation behavior in the reactor vessel. After achieving the stable conditions by operating the dipped-type DHX or the penetrated-type DHX, the other DHX was operated and the transient behavior was clarified by the temperature measurements. The clear temperature rise in the reactor vessel was confirmed by operating the penetrated-type DHX as second DHX operation under the condition of the dipped-type DHX operation at the beginning and the high heater power of fuel debris on the core catcher. Therefore, it was confirmed that the inhibition of the cooling for the decay heat occurred by operating multiple DHXs.
Koike, Akari*; Nakashima, Risako*; Nemoto, Masaya*; Sakai, Takaaki*; Doda, Norihiro; Tanaka, Masaaki
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 4 Pages, 2022/10
Due to global warming, the amount of snowfall in abnormal snowfall events may increase in the future. In order to evaluate the effect of global warming on the probability of exceeding the limit temperature at the core outlet as a core damage factor in a sodium-cooled fast reactor, a hazard curve of snowfall was developed considering global warming, and a dynamic PRA was performed. As a result, it was found that the amount of snowfall in abnormal snowfall events increases due to global warming, and the probability of exceeding the limit temperature increases.
Tsuji, Mitsuyo; Aizawa, Kosuke; Kobayashi, Jun; Kurihara, Akikazu
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 6 Pages, 2022/10
In sodium-cooled fast reactors (SFRs), decay heat removal after a core disruptive accident (CDA) is an important issue for the safety enhancement. Therefore, water experiments using a 1/10 scale experimental apparatus (PHEASANT) that simulates the reactor vessel of an SFR are conducted to investigate the natural circulation phenomena in the reactor vessel. In this study, experiments under the operation of the dipped-type DHX were conducted to investigate the effect of the heat generation ratio between the fuel debris on the core catcher in lower plenum and the reactor core remnant on the natural circulation behavior in the reactor vessel. The temperature distribution and the velocity distribution were measured under two heat generation conditions. Thus, the effect of the heat generation ratio between the fuel debris in the lower plenum and the reactor core remnant on the natural circulation behavior was quantitatively grasped under the dipped-type DHX operating conditions.
Onoda, Yuichi; Yamano, Hidemasa
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 9 Pages, 2022/10
In Japan, sodium-cooled fast reactor design takes In-Vessel Retention (IVR) strategy to stably cool damaged core materials in the reactor vessel during a severe accident with various design measures. Although a possibility to fail IVR is extremely low, a probabilistic risk assessment study needs a wide variety of scenarios including the IVR failure. Therefore, in order to study a wide range of event spectra related to stable cooling of debris in the reactor vessel, this study numerically investigated the deformation and failure behavior of the reactor vessel due to the debris deposited onto the skirt of the core catcher using the FINAS-STAR structural analysis code. The analyses are conducted in two cases of power density with the aim of investigating failure conditions of the bottom of the reactor vessel. Reactor vessel deforms significantly when the temperature reaches about 1100 
C and the reactor vessel reaches the failure criteria in high-power-density case.
Yamamoto, Seishiro*; Odaira, Naoya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 4 Pages, 2022/10
Takatsuka, Daichi*; Morita, Koji*; Liu, W.*; Zhang, T.*; Nakamura, Takeshi*; Kamiyama, Kenji
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 10 Pages, 2022/10