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Wakui, Takashi; Saito, Shigeru; Futakawa, Masatoshi
Materials, 17(23), p.5925_1 - 5925_14, 2024/12
The ductile properties of irradiated materials are one of the important indicators related to their structural integrity. Indentation tests are used for evaluating the ductile properties easily and rapidly. Constants in the material constitutive equation were identified via inverse analysis using the Kalman filter, such that the numerical experimental results reproduced the indentation test results. Numerical tensile experiments were conducted using stress-strain curves with the identified constants to obtain nominal stress-strain curves. Furthermore, two methods were proposed for evaluating the total elongation. Evaluated minimum total elongation was 10 %. The evaluation results of ion-irradiated materials were similar to the tensile test results of irradiated materials.
Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
We developed the measures for improving resilience of the sodium-cooled fast reactor structure using the failure mitigation technology and evaluated the effectiveness of the measures. To prevent core damage in the event of an accident progressing to an ultra-high temperature state, both measures to prevent overpressure in the reactor vessel and measures to cool the reactor core are required. As a core cooling measure, we developed a core cooling concept that promotes radiant heat transfer from the reactor vessel and cools the containment vessel outer surface by natural convection named Containment Vessel Auxiliary Cooling System (CVACS). We developed a method to use the reduction rate of core damage frequency as an indicator for effectiveness of the measures for improving resilience. The core damage frequency was evaluated by calculating the core cooling performance using CVACS, reflecting the results of structural analysis and human reliability analysis. By implementing measures for improving resilience in addition to existing measures, the core damage frequency of Japan loop-type sodium-cooled fast reactor caused by LOHRS has been reduced to about one-hundredth of the previous level.
Taniguchi, Yoshinori; Mihara, Takeshi; Kakiuchi, Kazuo; Udagawa, Yutaka
Annals of Nuclear Energy, 195, p.110144_1 - 110144_11, 2024/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)Sawa, Kazuhiro*; Haseda, Masaya*; Aihara, Jun
Nihon Kikai Gakkai Rombunshu (Internet), 89(921), p.22-00314_1 - 22-00314_6, 2023/05
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.
Oka, Hiroshi*; Kaito, Takeji; Ikusawa, Yoshihisa; Otsuka, Satoshi
Nuclear Engineering and Design, 370, p.110894_1 - 110894_8, 2020/12
Times Cited Count：1 Percentile：9.76(Nuclear Science & Technology)The objective of this study is to evaluate the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins were obtained from steady state irradiation in the EBR-II. Post irradiation examinations revealed that FP gas pressure was the main cause of creep damage in cladding, and that the stress contribution from FCMI was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to FP gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. The results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects are adopted.
Udagawa, Yutaka; Fuketa, Toyoshi*
Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08
Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki
Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06
Times Cited Count：3 Percentile：29.45(Nuclear Science & Technology)Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Li, J.*; Jang, S.*
Mechanical Engineering Journal (Internet), 7(3), p.19-00548_1 - 19-00548_11, 2020/06
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium. To improve the evaluation accuracy for the temperature distribution, a Lagrangian particle model for simulating reacting jet was also developed as an alternative method and its basic function was confirmed.
and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki
Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05
Times Cited Count：3 Percentile：29.45(Nuclear Science & Technology)Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki
Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05
Times Cited Count：1 Percentile：9.76(Nuclear Science & Technology)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.
Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki
Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12
Times Cited Count：8 Percentile：61.40(Nuclear Science & Technology)no abstracts in English
Taniguchi, Yoshinori; Udagawa, Yutaka; Mihara, Takeshi; Amaya, Masaki; Kakiuchi, Kazuo
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.551 - 558, 2019/09
Sato, Ikken
Journal of Nuclear Science and Technology, 56(5), p.394 - 411, 2019/05
Times Cited Count：12 Percentile：72.57(Nuclear Science & Technology)Water columns were adopted in the pressure measurement system of Fukushima-Daiichi Unit-3. Part of these water columns evaporated during the accident condition jeopardizing correct understanding on actual pressure. Through comparison of RPV (Reactor Pressure Vessel) and S/C pressures with D/W pressure, such water-column effect was evaluated. Correction for this effect was developed enabling clarification of slight pressure difference among RPV, S/C and D/W. This information was then integrated with other available data such as, water level, CAMS and environmental dose rate, into an interpretation of accident focusing on RPV and PCV pressurization/depressurization and radioactive material release to environment. It is suggested that dryout of in-vessel and ex-vessel debris was likely causing pressure decrease. S/C water poured into pedestal heated by relocated debris was the likely cause of pressurization. Cyclic reflooding of pedestal debris and dryout was likely.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 6 Pages, 2018/11
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium.
Uchibori, Akihiro; Takata, Takashi; Yanagisawa, Hideki*; Li, J.*; Jang, S.*
Proceedings of 2018 ANS Winter Meeting and Nuclear Technology Expo; Embedded Topical International Topical Meeting on Advances in Thermal Hydraulics (ATH 2018) (USB Flash Drive), p.1289 - 1294, 2018/11
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium. To improve the evaluation accuracy for the temperature distribution, a Lagrangian particle model for simulating reacting jet was also developed as an alternative method and its basic function was confirmed.
Haga, Katsuhiro; Kogawa, Hiroyuki; Wakui, Takashi; Naoe, Takashi; Takada, Hiroshi
Journal of Nuclear Science and Technology, 55(2), p.160 - 168, 2018/02
Times Cited Count：5 Percentile：42.11(Nuclear Science & Technology)The mercury target vessel used for the spallation neutron source in J-PARC has multi-walled structure made of stainless steel type 316L, which comprises a mercury vessel and a water shroud. In 2015, water leak incidents from the water shroud occurred while the mercury target was operated with a proton beam power of 500 kW. Several investigations were conducted to identify the cause of failure. The results of the visual inspections, mockup tests, and analytical evaluations suggested that the water leak was caused by the combination of two factors. One was the diffusion bonding failure due to the large thermal stress induced by welding of the bolt head, which fixes the mercury vessel and the water shroud, during the fabrication process. The other was the thermal fatigue failure of the seal weld due to the repetitive beam trip during the operating period. These target failures point to the importance of eliminating initial defects from welding lines and to secure the rigidity and reliability of welded structures. The next mercury target was fabricated with an improved design which adopted parts of monolithic structure machined by wire EDM to reduce welding lines, and intensified inspections to eliminate the initial defects. The operation with the improved target is planned to be started in October 2017.
Amaya, Masaki; Udagawa, Yutaka; Narukawa, Takafumi; Mihara, Takeshi; Taniguchi, Yoshinori
Proceedings of 2017 Water Reactor Fuel Performance Meeting (WRFPM 2017) (USB Flash Drive), 10 Pages, 2017/09
Aoyagi, Kazuhei; Ishii, Eiichi; Ishida, Tsuyoshi*
Journal of MMIJ, 133(2), p.25 - 33, 2017/02
In the construction of a deep underground facility, the hydromechanical properties of the rock mass around an underground opening are changed significantly due to stress redistribution. This zone is called an excavation damaged zone (EDZ). In high-level radioactive waste disposal, EDZs can provide a shortcut for the escape of radionuclides to the surface environment. Therefore, it is important to develop a method for predicting the detailed characteristics of EDZs. For prediction of the EDZ in the Horonobe Underground Research Laboratory of Japan, we conducted borehole televiewer surveys, rock core analyses, and repeated hydraulic conductivity measurements. We observed that niche excavation resulted in the formation of extension fractures within 0.2 to 1.0 m into the niche wall, i.e., the extent of the EDZ is within 0.2 to 1.0 m into the niche wall. These results are largely consistent with the results of a finite element analysis implemented with the failure criteria considering failure mode. The hydraulic conductivity in the EDZ was increased by 3 to 5 orders of magnitude compared with the outer zone. The hydraulic conductivity in and around the EDZ has not changed significantly in the two years following excavation of the niche. These results show that short-term unloading due to excavation of the niche created a highly permeable EDZ.
Uchibori, Akihiro; Ohshima, Hiroyuki
Mechanical Engineering Journal (Internet), 3(3), p.15-00620_1 - 15-00620_9, 2016/06
For assessment of the wastage environment under tube failure accident, a mechanistic computer code called SERAPHIM calculating compressible multicomponent multiphase flow with sodium-water chemical reaction has been developed. In this study, applicability of the SERAPHIM code was investigated through the analysis of the experiment on water vapor discharging in liquid sodium under actual condition of the steam generator. The numerical result showed that the underexpanded jet appeared and impinged on the target tube located above the discharging tube. The calculated temperature distribution agreed with the measurement result well. The liquid droplet entrainment and its transport were considered in this analysis. The region with higher impingement velocity of the liquid droplet was close to the wastage region confirmed in the experiment. It was demonstrated that the SERAPHIM code could predict the temperature distribution and the environment of LDIE under the actual condition.
