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Journal Articles

Transient response of LWR fuels (RIA)

Udagawa, Yutaka; Fuketa, Toyoshi*

Comprehensive Nuclear Materials, 2nd Edition, Vol.2, p.322 - 338, 2020/08

Journal Articles

Fracture-mechanics-based evaluation of failure limit on pre-cracked and hydrided Zircaloy-4 cladding tube under biaxial stress states

Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki

Journal of Nuclear Science and Technology, 57(6), p.633 - 645, 2020/06

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

Advancement of elemental analytical model in LEAP-III code for tube failure propagation

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.

Journal Articles

Analytical study of SPERT-CDC test 859 using fuel performance codes FEMAXI-8 and RANNS

Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

Journal Articles

The Effect of base irradiation on failure behaviors of UO$$_{2}$$ and chromia-alumina additive fuels under simulated reactivity-initiated accidents; A Comparative analysis with FEMAXI-8

Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki

Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

JAEA Reports

Code-B-2.5.2 for stress calculation for SiC-TRISO fuel particle

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$$_{2}$$ (PuO$$_{2}$$-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.

Journal Articles

Thresholds for failure of high-burnup LWR fuels by pellet cladding mechanical interaction under reactivity-initiated accident conditions

Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

Journal of Nuclear Science and Technology, 56(12), p.1063 - 1072, 2019/12

 Times Cited Count:2 Percentile:28.85(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Behavior of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

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

Journal Articles

An Interpretation of Fukushima-Daiichi Unit 3 plant data covering the two-week accident-progression phase based on correction for pressure data

Sato, Ikken

Journal of Nuclear Science and Technology, 56(5), p.394 - 411, 2019/05

 Times Cited Count:1 Percentile:53.3(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.

Journal Articles

Development of numerical analysis method for tube failure propagation under sodium-water reaction accident

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.

Journal Articles

Advancement of numerical analysis method for tube failure propagation

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.

Journal Articles

Technical investigation on small water leakage incident occurrence in mercury target of J-PARC

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:4 Percentile:26.29(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.

Journal Articles

Behavior of high-burnup advanced LWR fuels under design-basis accident conditions

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

Journal Articles

Field observations and failure analysis of an excavation damaged zone in the Horonobe Underground Research Laboratory

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.

Journal Articles

Applicability of a mechanistic numerical method for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

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.

Journal Articles

Applicability of a mechanistic numerical method for sodium-water reaction phenomena in steam generators of sodium-cooled fast reactors

Uchibori, Akihiro; Ohshima, Hiroyuki

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

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.

JAEA Reports

The Evaluation of the influence of hydride rim and biaxial stress condition on the cladding failure under a reactivity-initiated-accident by using EDC test method

Shinozaki, Takashi; Mihara, Takeshi; Udagawa, Yutaka; Sugiyama, Tomoyuki; Amaya, Masaki

JAEA-Research 2014-025, 34 Pages, 2014/12

JAEA-Research-2014-025.pdf:6.05MB

EDC test is a test method on the mechanical property of fuel cladding tube, and it focuses on the stress condition generated by PCMI under a RIA. We conducted EDC tests which simulate the mechanical conditions during a RIA by using the unirradiated cladding tubes which simulate hydride rim. Circumferential residual strains observed in post-test specimens tended to decrease with increasing the hydrogen concentration in the test cladding tubes and the thickness of the hydride rim. We also prepared RAG tube and performed EDC tests on it. It was observed that circumferential total strains at failure tended to decrease with increasing pre-crack depth on the outer surface of RAG tube specimen. We conducted biaxial stress tests by applying longitudinal tensile load onto RAG tube specimens. It was observed that circumferential total strains at failure under biaxial stress conditions tended to decrease compared to the results under uniaxial tensile condition.

Journal Articles

Effects of secondary depressurization on core cooling in PWR vessel bottom small break LOCA experiments with HPI failure and gas inflow

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

Journal of Nuclear Science and Technology, 43(1), p.55 - 64, 2006/01

 Times Cited Count:10 Percentile:38.25(Nuclear Science & Technology)

Effects of non-condensable gas from the accumulator tanks on secondary depressurization, as one of accident management (AM) measures in case of high pressure injection system failure, are studied at the ROSA-V/LSTF experiments simulating a ten instrument-tube break LOCA at the PWR vessel bottom. In an experiment with no gas inflow, the secondary depressurization at -55 K/h initiated by SI signal with 10 minutes delay succeeded in the LPI actuation. On the other hand, the gas inflow in another experiment degraded the primary depressurization and resulted in core uncovery before the LPI start. The third experiment with rapid secondary depressurization and continuous auxiliary feedwater supply, however, showed a possibility of long-term core cooling by the LPI actuation. RELAP5/MOD3 code analyses well predicted these experiment results and clarified that condensation heat transfer was largely degraded by the gas in the U-tubes. In addition, a primary pressure - coolant mass map was found to be useful for indication of key plant parameters of AM measures.

Journal Articles

Analysis on split failure of cladding of high burnup BWR rods in reactivity-initiated accident conditions by RANNS code

Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi

Nuclear Engineering and Design, 236(2), p.128 - 139, 2006/01

 Times Cited Count:6 Percentile:55.14(Nuclear Science & Technology)

A computer code RANNS was developed to analyze fuel rod behaviors in the RIA conditions. The code performs thermal and FEM-mechanical calculation for a single rod in axis-symmetric geometry to predict temperature profile, PCMI contact pressure, stress-strain distribution and their interactions. An experimental analysis by RANNS begins with pre-test conditions of irradiated rod which are given by FEMAXI-6. Analysis was performed on the simulated RIA experiments in NSRR, FK-10 and FK-12, of high burnup BWR rods in a cold start-up conditions, and PCMI process was discussed extensively. It was revealed that pellet thermal expansion dominates cladding deformation and subjects the cladding to bi-axial stress state, and thermal expansion in the cladding makes the stress in the inner region significantly lower than that in the outer region. Simulation calculations with wider pulses were carried out and the resulted cladding hoop stress was compared with the failure stress estimated in the NSRR experiments.

Journal Articles

Development of structural reliability evaluation method for aged piping considering uncertainty of seismic motions

Sugino, Hideharu*; Ito, Hiroto*; Onizawa, Kunio; Suzuki, Masahide

Nippon Genshiryoku Gakkai Wabun Rombunshi, 4(4), p.233 - 241, 2005/12

The purpose of this research is to establish the reliability evaluation method of aged nuclear power components for seismic events from a viewpoint of long-term use of the existing light water reactor nuclear power plants. For this purpose, we developed a piping failure probability evaluation code "PASCAL-SC" based on probabilistic fracture mechanics, and a probabilistic seismic hazard evaluation code "SHEAT-FM" for calculating the seismic occurrence probability of a plant site, paying attention to aging such as fatigue crack progress by the stress corrosion cracking and seismic load in primary coolant piping system. We proposed the reliability evaluation method of aged piping for seismic events by combination of these codes. Using this method, we evaluated the reliability of a weld line in the PLR(Primary Loop Recirculation system) piping of the BWR model plant for seismic events.

230 (Records 1-20 displayed on this page)