Watanabe, Tadashi*; Katsuyama, Jinya; Mano, Akihiro
International Journal of Nuclear and Quantum Engineering (Internet), 13(11), p.516 - 519, 2019/10
The estimation of leak flow rates through narrow cracks in structures is of importance for nuclear reactor safety, since the leak flow could be detected before occurrence of loss-of-coolant accidents. The two-phase critical leak flow rates are calculated using the system analysis code, and two representative non-homogeneous critical flow models, Henry-Fauske model and Ransom-Trapp model, are compared. The pressure decrease and vapor generation in the crack, and the leak flow rates are found to be larger for the Henry-Fauske model. It is shown that the leak flow rates are not affected by the structural temperature, but affected largely by the roughness of crack surface.
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.
Katsuyama, Jinya; Uno, Shumpei*; Watanabe, Tadashi*; Li, Y.
Frontiers of Mechanical Engineering, 13(4), p.563 - 570, 2018/12
For the structural integrity assessments on reactor pressure vessels (RPVs) under pressurized thermal shock (PTS) events, thermal hydraulic (TH) behavior of coolant water is one of the most important influence factors. Configuration of plant equipment and their dimensions, and operator action have large influences on TH behavior. In this study, to investigate the influence of operator action on TH behavior during a PTS event, we developed an analysis model for a typical Japanese plant, and performed TH and structural analyses. Two different operator action times were assumed based on the Japanese and US' rules. From the analysis results, it was clarified that differences in operator action times have a significant effect on TH behavior and loading conditions, that is, following the Japanese rule may lead to lower stresses compared to that when following the US rule because earlier operator action caused lower pressure in the RPV.
Ishigaki, Masahiro; Watanabe, Tadashi*
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 10 Pages, 2018/10
When coolant in one of the secondary side of steam generator (SG) is lost under some accident condition, the NC in the loop with the affected SG may terminate. Hence, the experiment was done in order to discuss the behavior of the natural circulation flow when the secondary side of the intact SG was depressurized stepwisely and that of the affected SG was empty of coolant. In this paper, we analyzed this NC experiment using the LSTF by the TRACE code. The objective of this analysis is to clarify the sensitivity of the code to the NC behavior. The calculated mass flow rate in the intact loop was slightly underestimated compared with the experimental result. On the other hand, the calculated mass flow rate in the affected loop was overestimated compared with the experimental result. In addition, we did the sensitivity analysis of the NC behavior in the case that the cooldown rate was changed.
Watanabe, Tadashi*; Ishigaki, Masahiro*; Katsuyama, Jinya
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 9 Pages, 2018/10
The analyses of LSTF experiment and PWR plant for 5% cold-leg break LOCA are performed using the RELAP5/MOD3.3 code. The discharge coefficient of critical flow model is determined so as to obtain the agreement of pressure transient between the LSTF experiment and the experimental analysis, and used for the PWR analysis. The characteristics of thermal-hydraulic phenomena in the experiment are shown to be simulated well by the two analyses. The decrease in core differential pressure during the loop-seal clearing is, however, underestimated by the two analyses, and the core heat up is not predicted. The loop flow rates are also underestimated by the two analyses. Although the duration of core heat up during the boil-off period is longer in the experimental analysis, the results of two analyses agree well, and the effect of scaling is found to be small between the experimental analysis and the PWR analysis.
Tani, Norio; Watanabe, Yasuhiro; Hotchi, Hideaki; Harada, Hiroyuki; Yamamoto, Masanobu; Kinsho, Michikazu; Igarashi, Susumu*; Sato, Yoichi*; Shirakata, Masashi*; Koseki, Tadashi*
Proceedings of 13th Annual Meeting of Particle Accelerator Society of Japan (Internet), p.708 - 711, 2016/11
At the J-PARC Main Ring (MR), there have been various investigation carried out at the moment aiming at the beam operation of MW order. As one of the investigations, a study of the Rapid-Cycling Synchrotron (RCS) magnets was implemented. Increase of the extraction energy of RCS was needed to reduce beam loss, as beam loss in the MR injection region was large under influence of Space Charge effect at the injection beam of 3GeV. Therefore conceptual design of the extraction energy upgrade using dipole and quadrupole magnets of RCS was performed. In this paper, we will report the contents of the study in extraction energy upgrade of RCS magnets and problems which became clear as a result.
Uno, Shumpei; Katsuyama, Jinya; Watanabe, Tadashi*; Li, Y.
Proceedings of 2016 ASME Pressure Vessels and Piping Conference (PVP 2016) (Internet), 8 Pages, 2016/07
For structural integrity assessment on reactor pressure vessels (RPVs) of pressurized water reactor during the pressurized thermal shock (PTS) events, thermal history of coolant water and heat transfer coefficient between coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analyses simulating PTS events and Jackson-Fewster correlation. Using these values, subsequently, loading conditions for structural integrity assessments of RPVs are evaluated by structural analyses. Nowadays, three-dimensional TH and structural analyses are recognized as precise assessment method for structural integrity of RPVs. In this study, we performed the TH and structural analyses using three-dimensional models of cold-leg, downcomer and RPV in order to evaluate loading conditions during a PTS event more accurately. Distributions of temperature of coolant water and heat transfer coefficient on the surface of RPV were calculated by TH analysis. Loading condition evaluation was then performed by using these values and taking the weld residual stress due to weld-overlay cladding and post-weld heat treatment into consideration. From these analyses, we obtained histories and distributions of loading conditions at the reactor core region of RPV. We discussed the conservativeness of current structural integrity assessment method of RPV prescribed in the current codes through the comparison in the loading conditions due to PTS event.
Tomizawa, Hiromitsu*; Sato, Takahiro*; Ogawa, Kanade*; Togawa, Kazuaki*; Tanaka, Takatsugu*; Hara, Toru*; Yabashi, Makina*; Tanaka, Hitoshi*; Ishikawa, Tetsuya*; Togashi, Tadashi*; et al.
High Power Laser Science and Engineering, 3, p.e14_1 - e14_10, 2015/04
no abstracts in English
Katsuyama, Jinya; Katsumata, Genshichiro; Onizawa, Kunio; Watanabe, Tadashi*; Nishiyama, Yutaka
Proceedings of 2014 ASME Pressure Vessels and Piping Conference (PVP 2014) (DVD-ROM), 9 Pages, 2014/07
For structural integrity assessment on reactor pressure vessel (RPV) of pressurized water reactor during the pressurized thermal shock (PTS) events, temperature of coolant water and heat transfer coefficient between coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events. Using these values, structural integrity assessment of RPV is performed by thermal-structural analysis, e.g. loading that affects the crack initiation and propagation is evaluated. In this study, we performed the TH and thermal-structural analyses using three-dimensional model of cold-leg, downcomer and RPV to assess loading conditions during the PTS more accurate. We obtained the loading histories at the reactor core region of RPV where a crack is postulated in the structural integrity assessment. Through the comparison between analysis results and current evaluation method, conservatism of current method will be discussed.
Takeda, Takeshi; Watanabe, Tadashi; Maruyama, Yu; Nakamura, Hideo
NEA/CSNI/R(2013)8/PART2 (Internet), p.173 - 183, 2013/12
An experiment simulating a PWR cold leg IBLOCA with 17% break at cold leg was conducted. The post-test analysis by RELAP5/MOD188.8.131.52 code revealed that cladding surface temperature was underpredicted due to later major core uncovery. The post-test analysis conditions were considered as Base Case assuming the discrepancies were caused by uncertainties in the code predictability and input data. Key phenomena and related important parameters, which may affect the cladding surface temperature, were selected based on the LSTF test data analysis and post-test analysis results. Sensitivity analyses were performed by changing the parameters relevant to the key phenomena within the ranges to investigate influences of the parameters onto the cladding surface temperature. It was confirmed that both constant C of Wallis CCFL correlation at the core exit and inter-phase drag in the core are more sensitive to the cladding surface temperature.
Sato, Takahiro*; Iwasaki, Atsushi*; Owada, Shigeki*; Yamanouchi, Kaoru*; Takahashi, Eiji*; Midorikawa, Katsumi*; Aoyama, Makoto; Yamakawa, Koichi; Togashi, Tadashi*; Fukami, Kenji*; et al.
Journal of Physics B; Atomic, Molecular and Optical Physics, 46(16), p.164006_1 - 164006_6, 2013/08
By introducing 13th- (61.7 nm) and 15th-order harmonics (53.4 nm) of femtosecond laser pulses at 800 nm into an undulator of SCSS (SPring-8 Compact SASE Source) test accelerator at RIKEN, these harmonic pulses were amplified by a factor of more than 10 with a high contrast ratio through the interaction between accelerated electron bunches and the harmonic pulses. From numerical simulations of the amplification processes of high-order harmonic pulses in the undulator, optimum conditions of the electron bunch duration interacting with the high-order harmonic pulses were investigated for generating full-coherent and intense pulses in the extreme ultraviolet wavelength region.
Ogawa, Kanade*; Sato, Takahiro*; Matsubara, Shinichi*; Okayasu, Yuichi*; Togashi, Tadashi*; Watanabe, Takahiro*; Takahashi, Eiji*; Midorikawa, Katsumi*; Aoyama, Makoto; Yamakawa, Koichi; et al.
Proceedings of 10th Conference on Lasers and Electro-Optics Pacific Rim and 18th OptoElectronics and Communications Conference and Photonics in Switching 2013 (CLEO-PR & OECC/PS 2013) (USB Flash Drive), 2 Pages, 2013/06
no abstracts in English
Togashi, Tadashi*; Takahashi, Eiji*; Midorikawa, Katsumi*; Aoyama, Makoto; Yamakawa, Koichi; Sato, Takahiro*; Iwasaki, Atsushi*; Owada, Shigeki*; Yamanouchi, Kaoru*; Hara, Toru*; et al.
Proceedings of Ultrafast Optics IX (CD-ROM), 2 Pages, 2013/03
We have demonstrated free-electron laser radiation seeded by high-order harmonics in the extreme-ultraviolet region. Strong enhancement of the radiation intensity by a factor of 104 was observed with timing control of an electro-optical sampling technique.
Kino, Chiaki; Watanabe, Tadashi*; Nishida, Akemi; Takemiya, Hiroshi
Nippon Kikai Gakkai Rombunshu, B, 78(796), p.2113 - 2126, 2012/12
Flow around an in-line forced oscillating circular cylinder was simulated numerically by using OpenFOAM in order to clarify the mechanism of flow induced vibration. Immersed boundary Method is used to solve the moving boundary. Reynolds number is set to 1000 and the reduced velocity is set to the range from 0 to 10. In the first excitation region, it is shown that negative drag force which is a factor for an in-line oscillation of a circular cylinder comes from contacting between high pressure region and a circular cylinder. The present simulation shows that twin vortex has an important role on the contact phenomena. In the second excitation region, it is shown that time averaged lift drag doesn't become zero on some oscillating conditions. It is considered that a cross-flow oscillation comes from the phenomena.
Takeda, Takeshi; Watanabe, Tadashi; Nakamura, Hideo
Proceedings of 9th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-9) (CD-ROM), 12 Pages, 2012/09
Ishigaki, Masahiro; Watanabe, Tadashi; Nakamura, Hideo
Proceedings of OECD/NEA and IAEA Workshop on Experimental Validation and Application of CFD and CMFD Codes in Nuclear Reactor Technology (CFD4NRS-4) (CD-ROM), 11 Pages, 2012/09
In this research, two-phase critical flow with the phase change in the Super Moby Dick convergent-divergent nozzle is analyzed numerically by the CFD code FLUENT, and we discuss the performance of the numerical code by comparing the mass flow rates, pressure and void fractions by the numerical simulations with those values by the experiments. The cavitation model is applied together with the evaporation-condensation model. The estimations of mass flow rate agree with the experimental results for the low axial pressure difference. On the other hand, the mass flow rates are under-estimated when the axial pressure difference is high. It is found that the mass flow rate is improved much by taking into account the effect of the wall vapor generation.
Iwasaki, Atsushi*; Sato, Takahiro*; Owada, Shigeki*; Togashi, Tadashi*; Takahashi, Eiji*; Midorikawa, Katsumi*; Aoyama, Makoto; Yamakawa, Koichi; Matsubara, Shinichi*; Okayasu, Yuichi*; et al.
Proceedings of International Conference on Ultrafast Phenomena 2012 (UP 2012) (Internet), 3 Pages, 2012/07
no abstracts in English
Takeda, Takeshi; Maruyama, Yu; Watanabe, Tadashi; Nakamura, Hideo
Journal of Power and Energy Systems (Internet), 6(2), p.87 - 98, 2012/06
Ishigaki, Masahiro; Watanabe, Tadashi; Nakamura, Hideo
Journal of Power and Energy Systems (Internet), 6(2), p.264 - 274, 2012/06
Two-phase critical flow in the nozzle tube is analyzed numerically by the best estimate code TRACE and the CFD code FLUENT, and the performance of the mass flow rate estimation by the numerical codes is discussed. For the best estimate analysis by the TRACE code, the critical flow option is turned on. The mixture model is used with the cavitation model and the evaporation-condensation model for the numerical simulation by the FLUENT code. Two test cases of the two-phase critical flow are analyzed. One case is the critical flashing flow in a convergent-divergent nozzle (Super Moby Dick experiment), and the other case is the break nozzle flow for a steam generator tube rupture experiment of pressurized water reactors at Large Scale Test Facility of Japan Atomic Energy Agency. The calculation results of the mass flow rates by the numerical simulations show good agreements with the experimental results.
Uzawa, Ken; Watanabe, Tadashi*
Journal of Power and Energy Systems (Internet), 6(2), p.229 - 240, 2012/06
To prevent anomalous events of equipment and structures in a spent fuel pool from occurring, it is essential to comprehend overflow of water and fluid pressure caused by sloshing. For this purpose, it is necessary to quantitatively evaluate turbulent dissipation near water surface and inner structures instead of an empirical value which depends on size of each pool. Therefore, we test the efficacy of turbulence models by performing numerical simulation of dam break phenomenon as an elementary process of a sloshing to evaluate turbulent dissipation based on action mechanism. In this research, we derived a conservation equation of turbulent kinetic energy of RANS models and compared contributions of pressure, gravity and turbulent dissipation terms to the kinetic energy. As a result, we demonstrated that turbulent dissipation term is a predominant factor on the dynamics of dam break flow and turbulent dissipation is overestimated in eddy viscosity model.