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Fukano, Yoshitaka
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
An evaluation on the consequences of a hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) using SAS4A code was also performed in the past study. SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in Monju by this developed SAS4A code was performed. It was clarified by the analyses considering power control system that the reactor would be safely shut down by the plant protection system triggered by either of 116 percent over power or delayed neutron detector trip signals. Therefore the conclusion in the past study that the consequences of HTIB event would be much less severe than that of unprotected loss-of-flow event was strongly supported by this study. Furthermore SAS4A code was newly validated using an in-pile experiment which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study.
Li, F.; Mihara, Takeshi; Udagawa, Yutaka; Amaya, Masaki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07
Negyesi, M.; Amaya, Masaki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
Ota, Yoshimi*; Nishida, Akemi; Tsubota, Haruji; Li, Y.
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07
Many empirical formulae have been proposed to evaluate the local damage to reinforced concrete structures caused by the impact of rigid projectiles. Most of these formulae have been derived based on impact tests perpendicular to the target structures. To date, few impact tests oblique to the target structures have been conducted. The purpose of this study is to propose a new formula for evaluating the local damage caused by oblique impacts based on experiments and simulations. The new formula is derived by modifying an empirical formulation for normal impact and the agreement with results of past oblique impact tests is discussed.
Nishida, Akemi; Ota, Yoshimi*; Tsubota, Haruji; Li, Y.
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
Many empirical formulae have been proposed for evaluating the local damage to reinforced concrete structures caused by rigid projectile impact. Most of these formulae are based on impact tests perpendicular to the target structures. To date, few impact tests oblique to the target structures have been conducted. In this study, we aim to obtain a new formula for evaluating the local damage caused by oblique impacts based on previous experimental and simulation results. We analyze and simulate the local damage owing to impact by deformable projectiles. The experimental and simulation results were in good agreement and confirmed the validity of the proposed analytical method. Furthermore, the internal energy of the deformable projectile absorbed upon impact was approximately 60% of the total energy. In comparison to a rigid projectile, it is possible to reduce the impact load and consequently the damage to the target.
Tsubota, Haruji; Ota, Yoshimi*; Nishida, Akemi; Li, Y.
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
Many empirical formulae have been proposed for evaluating the local damage to reinforced concrete structures caused by rigid projectile impact. Most of these formulae have been derived based on impact tests perpendicular to target structures. To date, few impact tests oblique to target structures have been carried out. Therefore, in this study, we propose a new formula for evaluating the local damage caused by oblique impact based on past experimental results and the results of simulation analyses. In this paper, we present the results of simulation analyses for evaluating the local damage to reinforced concrete panels caused by normal and oblique impact of deformable projectiles, focusing especially on perforation phenomena. Based on the analytical results, we investigate the differences in impact response characteristics between normal and oblique impact.
Matsumoto, Toshinori; Sato, Masatoshi; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07
Goto, Minoru; Ueta, Shohei; Aihara, Jun; Inaba, Yoshitomo; Fukaya, Yuji; Tachibana, Yukio; Okamoto, Koji*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07
A PuO
-YSZ fuel kernel with a ZrC coating, which enhances safety, security and safeguard, namely: 3S-TRISO fuel, was proposed to introduce to the plutonium-burner HTGR. In this study, the efficiency of the ZrC coating as the free-oxygen getter was examined based on a thermochemical calculation. A preliminary study on the feasibility of the 3S-TRISO fuel was conducted focusing on the internal pressure. Additionally, a nuclear feasibility of the reactor core was studied. As a result, all the amount of the free-oxygen is captured by a thin ZrC coating under 1600
C and coating ZrC on the fuel kernel should be very effective method to suppress the internal pressure. The internal pressure of the 3S-TRISO fuel at 500 GWd/t is lower than that of UO kernel TRISO fuel whose feasibility had been already confirmed and the 3S-TRISO fuel should be feasible. The fuel shuffling allows to achieve 500 GWd/t. The temperature coefficient of reactivity is negative during the operation period and thus the nuclear feasibility of the reactor core should be achievable.
Yamashita, Susumu; Uesawa, Shinichiro; Yoshida, Hiroyuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
Uesawa, Shinichiro; Yamashita, Susumu; Shibata, Mitsuhiko; Yoshida, Hiroyuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
Fujita, Shunya*; Abe, Yutaka*; Kaneko, Akiko*; Chonan, Fuminori*; Yuasa, Tomohisa*; Yamaki, Tatsunori*; Segawa, Tomoomi; Yamada, Yoshikazu
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 8 Pages, 2017/07
From the observation results, in the process of flushing, the behaviors leading to flushing were classified divided into three types. First type is that first generation bubble from heating leads to flushing. Second type is that nucleate boiling continues during heating and stop, finally single bubble generates and leads to flushing. Third type is defined that gradual evaporation occurs without bubbles. It was revealed that the total quantities of heat released by flushing are approximately equal when assuming the flushing mechanism, it can be triggered that a large amount of micro bubbles are instantaneously generated and grew.
C/steel interaction and oxidation during severe accidentsIshikawa, Jun; Shiotsu, Hiroyuki; Sugiyama, Tomoyuki; Maruyama, Yu
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07
In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.
Nava, M.*; Ito, Daisuke*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Suzuki, Toru*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 5 Pages, 2017/07
Zhan, Y.*; Oya, Naoki*; Enoki, Koji*; Okawa, Tomio*; Ohno, Shuji; Aoyagi, Mitsuhiro; Takata, Takashi
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 6 Pages, 2017/07
It is important to set the amount of sodium droplet mechanistically for appropriate numerical evaluations of sodium leak and fire behavior in a sodium-cooled fast reactor plant. In the present work, fundamental experiments are performed to measure the splash ratio during the vertical water jet impact onto a horizontal wall. It is shown that the splash ratio can be correlated well as a function of the impact Weber number and dimensionless impact frequency.
Tanaka, Masaaki; Murakami, Satoshi*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
Thermal striping on the core instrumentation plate (CIP) around the primary control rod (PCR) and backup control rod (BCR) channels and the radial blanket fuel assemblies (RBAs) may be caused. Since the interaction between neighbor areas exists in the UIS and the cold sodium flowing from the RBA is affected by the external flow around the UIS, a spatial connection method consisting of the numerical model for the whole upper plenum and the local target area has been developed. The numerical results were compared with the experimental results to confirm applicability of the method to the practical problem. And, sensitivity of mesh arrangement to the numerical results was discussed by using wide and narrow area models with two different spatial resolutions in each model. Through the examinations, appropriate local model for the spatial connection mothed could be proposed.
Yamada, Fumiaki; Imaizumi, Yuya; Nishimura, Masahiro; Fukano, Yoshitaka; Arikawa, Mitsuhiro*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
The loss-of-reactor-level (LORL) is one of the loss-of-heat-removal-system (LOHRS) of beyond-DBA (BDBA) severe accident. An evaluation method for the LORL which is caused by the coolant leakage in two positions of the primary heat transport system (PHTS) was developed for prototype JSFR which is loop-type sodium-cooled fast reactor. The secondary leakage in cold standby which occurred in different loop from that of the first leakage in rated power operation can lead LORL by excessive declining of the sodium level. Therefore, the sodium level behavior in RV was studied in a representative accident sequence by considering the sodium pumping up into RV, siphon-breaking to stop pumping out from RV and maintain the sodium level, and calculation programs for the transient sodium level in RV. The representative sequence with lowest sodium level was selected by considering combinations of possible leakage positions. As a result of the evaluation considering the countermeasures above, it was revealed that the LOHRS can be prevented by maintaining the sodium level for the operation of decay heat removal system, even in the leakages in two positions of PHTS which corresponds to BDBA.
Mano, Akihiro; Yamaguchi, Yoshihito; Katsuyama, Jinya; Li, Y.
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 7 Pages, 2017/07
Recently, cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel based alloy welds in the primary piping of pressurized water reactors. Structural integrity assessments taking PWSCC into account have become important. Probabilistic fracture mechanics (PFM) is expected as one of rational methods for the assessments because it can account for uncertainty of the influencing factors and evaluate the failure probabilities of components. In JAEA, a PFM analysis code PASCAL-SP was developed to evaluate the failure probability of nuclear pipe. This paper details improvement of the PASCAL-SP to evaluate the failure probability taking PWSCC into account. As numerical examples, the failure probabilities for circumferential and axial cracks due to PWSCC are evaluated. Influence of inspection on failure probabilities are evaluated. As the results, we conclude that the improved PASCAL-SP is useful for evaluating the failure probability taking PWSCC into account.
Li, Y.; Katsumata, Genshichiro*; Masaki, Koichi*; Hayashi, Shotaro*; Itabashi, Yu*; Nagai, Masaki*; Suzuki, Masahide*; Kanto, Yasuhiro*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock transients. In this study, as a part of the verification activities, a working group was established in Japan, with seven organizations from industry, universities and institutes voluntarily participating as members. The source program of PASCAL was released to the members of the working group. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group including the verification plan, approaches and results.
Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
