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CNguyen, B. V. C.*; Murakami, Kenta*; Chena, L.*; Phongsakorn, P. T.*; Chen, X.*; Hashimoto, Takashi; Hwang, T.*; Furusawa, Akinori; Suzuki, Tatsuya*
Nuclear Materials and Energy (Internet), 39, p.101639_1 - 101639_9, 2024/06
Lu, K.; Takamizawa, Hisashi; Li, Y.; Masaki, Koichi*; Takagoshi, Daiki*; Nagai, Masaki*; Nannichi, Takashi*; Murakami, Kenta*; Kanto, Yasuhiro*; Yashirodai, Kenji*; et al.
Mechanical Engineering Journal (Internet), 10(4), p.22-00484_1 - 22-00484_13, 2023/08
Lu, K.; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.
International Journal of Pressure Vessels and Piping, 199, p.104706_1 - 104706_13, 2022/10
Times Cited Count：3 Percentile：60.63(Engineering, Multidisciplinary)Iwata, Keiko; Hata, Kuniki; Tobita, Toru; Hirota, Takatoshi*; Takamizawa, Hisashi; Chimi, Yasuhiro; Nishiyama, Yutaka
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07
Takeda, Takeshi
JAEA-Data/Code 2021-006, 61 Pages, 2021/04
JAEA-Data-Code-2021-006.pdf:2.78MB
An experiment denoted as SB-PV-09 was conducted on November 17, 2005 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-PV-09 simulated a 1.9% pressure vessel top small-break loss-of-coolant accident in a pressurized water reactor (PWR). The test assumptions included total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). In the experiment, liquid level in the upper-head was found to control break flow rate. When maximum core exit temperature reached 623 K, steam generator (SG) secondary-side depressurization was initiated by fully opening the relief valves in both SGs as an accident management (AM) action. The AM action, however, was ineffective on the primary depressurization until the SG secondary-side pressure decreased to the primary pressure. Meanwhile, the core power was automatically reduced when maximum cladding surface temperature of simulated fuel rods exceeded the pre-determined value of 958 K to protect the LSTF core due to late and slow response of core exit temperature. After the automatic core power reduction, loop seal clearing (LSC) was induced in both loops by steam condensation on the ACC coolant injected into cold legs. The whole core was quenched because of core recovery after the LSC. After the ACC tanks started to discharge nitrogen gas, the pressure difference between the primary and SG secondary sides became larger. After the continuous core cooling was confirmed through the actuation of low pressure injection system of ECCS, the experiment was terminated. This report summarizes the test procedures, conditions, and major observations in the ROSA/LSTF experiment SB-PV-09.
Lu, K.; Katsuyama, Jinya; Li, Y.; Yoshimura, Shinobu*
Journal of Pressure Vessel Technology, 143(2), p.021505_1 - 021505_8, 2021/04
Times Cited Count：1 Percentile：10.51(Engineering, Mechanical)Lu, K.; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 142(5), p.051501_1 - 051501_10, 2020/10
Times Cited Count：2 Percentile：15.81(Engineering, Mechanical)Lu, K.; Katsuyama, Jinya; Li, Y.
Proceedings of ASME 2020 Pressure Vessels and Piping Conference (PVP 2020) (Internet), 10 Pages, 2020/08
Lu, K.; Katsuyama, Jinya; Li, Y.; Miyamoto, Yuhei*; Hirota, Takatoshi*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*
Mechanical Engineering Journal (Internet), 7(3), p.19-00573_1 - 19-00573_14, 2020/06
Lu, K.; Katsuyama, Jinya; Li, Y.
Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04
Times Cited Count：6 Percentile：42.63(Engineering, Mechanical)Katsuyama, Jinya; Osakabe, Kazuya*; Uno, Shumpei*; Li, Y.; Yoshimura, Shinobu*
Journal of Pressure Vessel Technology, 142(2), p.021205_1 - 021205_10, 2020/04
Times Cited Count：2 Percentile：15.81(Engineering, Mechanical)no abstracts in English
Aihara, Jun; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio
JAEA-Data/Code 2019-018, 22 Pages, 2020/01
JAEA-Data-Code-2019-018.pdf:1.39MB
Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO
(PuO-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. On the other hand, we have developed Code-B-2.5.2 for prediction of pressure vessel failure probabilities of SiC-tri-isotropic (TRISO) coated fuel particles for HTGRs under operation by modification of an existing code, Code-B-2. The main purpose of modification is preparation of applying code for CFPs of Pu-burner HTGR. In this report, basic formulae are described.
Lu, K.; Katsuyama, Jinya; Li, Y.; Yoshimura, Shinobu*
Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 9 Pages, 2019/07
Katsuyama, Jinya; Masaki, Koichi; Lu, K.; Watanabe, Tadashi*; Li, Y.
Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 7 Pages, 2019/07
For reactor pressure vessel (RPV) of pressurized water reactor, temperature of coolant water in emergency core cooling system (ECCS) may have influence on the structural integrity of RPV during pressurized thermal shock (PTS) events. Focusing on a mitigation measure to raise the coolant water temperature of ECCS for aged RPVs in order to reduce the effect of thermal shock due to PTS events, we performed thermal hydraulic analyses and probabilistic fracture mechanics analyses by using RELAP5 and PASCAL4, respectively. From the analysis results, it was shown that the failure probability of RPV was dramatically reduced when the coolant temperature in accumulator as well as high and low pressure injection systems (HPI/LPI) was raised, although raising the coolant temperature of HPI/LPI only did not cause reduction in the failure probability.
Lu, K.; Katsuyama, Jinya; Li, Y.; Miyamoto, Yuhei*; Hirota, Takatoshi*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05
Ito, Chikara; Naito, Hiroyuki; Ishikawa, Takashi; Ito, Keisuke; Wakaida, Ikuo
JPS Conference Proceedings (Internet), 24, p.011038_1 - 011038_6, 2019/01
A high-radiation resistant optical fiber has been developed in order to investigate the interiors of the reactor pressure vessels and the primary containment vessels at the Fukushima Daiichi Nuclear Power Station. The tentative dose rate in the reactor pressure vessels is assumed to be up to 1 kGy/h. We developed a radiation resistant optical fiber consisting of a 1000 ppm hydroxyl doped pure silica core and 4 % fluorine doped pure silica cladding. We attempted to apply the optical fiber to remote imaging technique by means of fiberscope. The number of core image fibers was increased from 2000 to 22000 for practical use. The transmissive rate of infrared images was not affected after irradiation of 1 MGy. No change in the spatial resolution of the view scope by means of image fiber was noted between pre- and post-irradiation. We confirmed the applicability of the probing system, which consists of a view scope using radiation-resistant optical fibers.
Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio
JAEA-Data/Code 2018-013, 60 Pages, 2018/11
JAEA-Data-Code-2018-013.pdf:1.67MB
Mechanical properties of materials including fracture toughness are extremely important for evaluating the structural integrity of reactor pressure vessels (RPVs). In this report, the published data of mechanical properties of nuclear RPVs steels, including neutron irradiated materials, acquired by the Japan Atomic Energy Agency (JAEA), specifically tensile test data, Charpy impact test data, drop-weight test data, and fracture toughness test data, are summarized. There are five types of RPVs steels with different toughness levels equivalent to JIS SQV2A (ASTM A533B Class 1) containing impurities in the range corresponding to the early plant to the latest plant. In addition to the base material of RPVs, the mechanical property data of the two types of stainless overlay cladding materials used as the lining of the RPV are summarized as well. These mechanical property data are organized graphically for each material and listed in tabular form to facilitate easy utilization of data.
Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 8 Pages, 2018/07
Lu, K.; Masaki, Koichi; Katsuyama, Jinya; Li, Y.; Uno, Shumpei*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 10 Pages, 2018/07
Chimi, Yasuhiro; Iwata, Keiko; Tobita, Toru; Otsu, Takuyo; Takamizawa, Hisashi; Yoshimoto, Kentaro*; Murakami, Takeshi*; Hanawa, Satoshi; Nishiyama, Yutaka
JAEA-Research 2017-018, 122 Pages, 2018/03
JAEA-Research-2017-018.pdf:44.03MB
Warm pre-stress (WPS) effect is a phenomenon that after applying a load at a high temperature fracture does not occur in unloading during cooling, and then the fracture toughness in reloading at a lower temperature increases effectively. Engineering evaluation models to predict an apparent fracture toughness in reloading are established using experimental data with linear elasticity. However, there is a lack of data on the WPS effect for the effects of specimen size and surface crack in elastic-plastic regime. In this study, fracture toughness tests were performed after applying load-temperature histories which simulate pressurized thermal shock transients to confirm the WPS effect. The experimental results of an apparent fracture toughness tend to be lower than the predictive results using the engineering evaluation models in the case of a high degree of plastic deformation in preloading. Considering the plastic component of preloading can refine the engineering evaluation models.
