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Yamashita, Shinichiro
Nihon Genshiryoku Gakkai-Shi ATOMO
, 65(4), p.233 - 237, 2023/04
In the wake of the accident at the Fukushima Daiichi Nuclear Power Plant (NPP) of TEPCO due to the Great East Japan Earthquake in 2011, interest in the early implementation of accident tolerant fuel (ATF) not only for many existing NPPs but also for future NPPs, which is expected to dramatically improve the safety of light water reactors, has increased globally, and research and development is currently underway in many countries around the world. In this article, an overview of domestic ATF technology development that has been carried out with the support of the Ministry of Economy, Trade and Industry since 2015, will be introduced.
Madokoro, Hiroshi; Yamashita, Takuya; Sato, Ikken; Gaus-Liu, X.*; Cron, T.*; Fluhrer, B.*; St
ngle, R.*; Wenz, T.*; Vervoortz, M.*; Mizokami, Shinya
Proceedings of 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (Internet), 16 Pages, 2022/03
Nagai, Haruyasu
Fission Product Behavior under Severe Accident, p.112 - 116, 2021/05
no abstracts in English
Madokoro, Hiroshi; Sato, Ikken
Nuclear Engineering and Design, 376, p.111123_1 - 111123_15, 2021/05
Times Cited Count：6 Percentile：70.8(Nuclear Science & Technology)Kawaguchi, Munemichi; Miyahara, Shinya*; Uno, Masayoshi*
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021305_1 - 021305_9, 2020/04
Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater were performed to investigate the chemical reaction beneath the internal heater (800
C), which was used to simulate the obstacle and heating effect on SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. The internal heater on the concrete hindered the transport of Na into the concrete. Therefore, Na could start to react with the concrete at the periphery of the internal heater, and the concrete ablation depth at the periphery was larger than under the internal heater. The high Na pool temperature of 800C increased largely the Na aerosol release rate, which was explained by Na evaporation and hydrogen bubbling, and formed the porous reaction product layer, whose porosity was 0.54-0.59 from the mass balance of Si and image analyzing EPMA mapping. They had good agreement with each other. The porous reaction products decreased the amount of Na transport into the reaction front. The Na concentration around the reaction front became about 30wt.% despite the position of the internal heater. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.
Yoshikawa, Shinji
JAEA-Technology 2019-024, 22 Pages, 2020/03
JAEA-Technology-2019-024.pdf:1.76MB
JAEA-Technology-2019-024-appendix(CD-ROM).zip:73.55MB
In various technical fields of nuclear energy, computer codes are often used for transient simulations of target phenomena. Some of the codes were developed many years ago and have been revised with newly acquired findings, rather than newly developed, because of many encompassed numerical models and complexity of algorithms. In many cases, available outputs for users are output text files and graphs showing temporal variations of parameters, despite diversified and huge number of output information items are posing difficulty to the users in grasping the whole picture of the reproduced phenomena. This report compiles a series of know-hows in building a post-processor software for large simulation codes which serves as an interactive tool for code users in understanding the reproduced consequence with visually understandable information items. These know-hows are acquired through post-processor developments for LWR severe accident simulation codes RELAP/SCDAPSIM and MELCOR.
Takeuchi, Tomoaki; Otsuka, Noriaki; Nakano, Hiroko; Iida, Tatsuya*; Ozawa, Osamu*; Shibagaki, Taro*; Komanome, Hirohisa*; Tsuchiya, Kunihiko
QST-M-16; QST Takasaki Annual Report 2017, P. 67, 2019/03
no abstracts in English
Nemoto, Yoshiyuki; Kaji, Yoshiyuki; Onizawa, Takashi*; Kanazawa, Toru*; Nakashima, Kazuo*; Tojo, Masayuki*
Proceedings of Annual Congress of the European Federation of Corrosion (EUROCORR 2018) (USB Flash Drive), 8 Pages, 2018/09
The authors proposed oxidation models based on oxidation data which previously obtained in high temperature oxidation tests on small sample of Zircalloy-2 (Zry2) cladding in dry air and in air/steam mixture environment. The oxidation models were implemented in computational fluid dynamics (CFD) code to analyse oxidation behavior of long cladding sample in hypothetical spent fuel pool (SFP) accident conditions. The oxidation tests were conducted using Zry2 cladding sample 500 mm in length. The oxide layer growth in dry air was well reproduced in the calculation using the oxidation model, meanwhile which in air/steam mixture was overestimated atmosphere composition change anticipated in the spent fuel rack during the accident, and its influence on the oxidation behaviour of the cladding were discussed in consideration of the oxidation model improvement.
Nakano, Hiroko; Hirota, Noriaki; Shibata, Hiroshi; Takeuchi, Tomoaki; Tsuchiya, Kunihiko
Mechanical Engineering Journal (Internet), 5(2), p.17-00594_1 - 17-00594_12, 2018/04
no abstracts in English
Nagaishi, Ryuji
Radioisotopes, 66(11), p.601 - 610, 2017/11
no abstracts in English
Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Komanome, Hirohisa*; Miura, Kuniaki*; Ishihara, Masahiro
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
After the accident at the Fukushima Dai-ichi (1F) Nuclear Power Plant (NPP), the Japanese Government referred to "Enhancement of instrumentation to identify the status of the reactors and PCVs", in the report of Japanese government to the IAEA ministerial conference in June 2011. In response to these provisions, a research and development of a monitoring system for NPPs situations during severe accidents started in November 2012. The objectives of the R&D are composed of radiation-resistant monitoring camera, radiation-resistant in-water transmission system, and heat-resistant signal cable. For all the three objectives, the elemental technologies have been already developed and now trial system are being fabricated and tested under simulated conditions of severe accidents. The results will enable us to determine the basic specifications of the systems and to provide the information about application limits for users.
Hino, Ryutaro; Takegami, Hiroaki; Yamazaki, Yukie; Ogawa, Toru
JAEA-Review 2016-038, 294 Pages, 2017/03
JAEA-Review-2016-038.pdf:11.08MB
In the aftermath of the Fukushima nuclear accident, safety measures against hydrogen in severe accident have been recognized as a serious technical problem in Japan. Therefore, efforts have begun to form a common knowledge base between nuclear engineers and experts on combustion and explosion, and to secure and improve future nuclear energy safety. As one of such activities, we have prepared the "Handbook of Advanced Nuclear Hydrogen Safety" under the Advanced Nuclear Hydrogen Safety Research Program funded by the Agency for Natural Resources and Energy of the Ministry of Economy, Trade and Industry. The concepts of the handbook are as follows: to show advanced nuclear hydrogen safety technologies that nuclear engineers should understand, to show hydrogen safety points to make combustion-explosion experts cooperate with nuclear engineers, to expand information on water radiolysis considering the situation from just after the Fukushima accidents and to the waste management necessary for decommissioning after the accident, etc.
irradiationTakeuchi, Tomoaki; Otsuka, Noriaki; Uehara, Toshiaki; Kumahara, Hajime*; Tsuchiya, Kunihiko
QST-M-2; QST Takasaki Annual Report 2015, P. 80, 2017/03
no abstracts in English
Tsuchiya, Kunihiko; Takeuchi, Tomoaki; Komanome, Hirohisa*; Miura, Kuniaki*; Araki, Masanori; Ishihara, Masahiro
Nihon Hozen Gakkai Dai-13-Kai Gakujutsu Koenkai Yoshishu, p.375 - 378, 2016/07
no abstracts in English
*; *
PNC TN9410 88-055, 111 Pages, 1988/06
In this study, decay heat removal capability of the Maintenance Cooling System (MCS) of Monju has been investigated with respect to protected accidents. The protected accidents of the Liquid Metal Fast Breeder Reactors (LMFBRs), such as Protected Loss-of-Heat-Sink (PLOHS) or Loss-of-Reactor-Level (LORL), are of great importance from the viewpoint of the annual frequency of core damage. The progression of the protected accidents is mild in general because reactor decay heat can be dispersed from the core by natural circulation. The decay heat for Monju is to be removed by the Intermediate Reactor Auxiliary Cooling system (IRACS). It is essential to keep the intactness of coolant flow path from the reactor core to the heat sink and the availability of heat sink itself. If the either of them is degraded, it is taken for granted d that protected slow meltdown follows. However, the reactor core can be prevented from any damage or meltdown if the decay heat can be removed through MCS. The plant thermohydraulics of the procected accidents is analyzed using SSC-L to develop success criteria in the decay heat removal by the MCS. Parametric calculations are performed with respect to: available heat capacity in the heat transport system, cooling time before the loss-of-heat-sink and MCS starting time. It has been found, for example, that (1)MCS can remove the decay heat immediately after the reactor shutdown if heat capacity of more than two main coolant loops is available; (2)after two hours cooling time by natual circulation, MCS can remove the decay heat even if no coolant flow is assumed in all the main heat transport system; (3)LORL caused by the failure in sodium make-up can be recovered by the MCS operation. In the PLOHS condition, the coolant temperature may exceed conservative design limit of the MCS piping. However, the conservativeness of the design limit and the method of qualification make compensation for the deterioration in structural strength. Finally, ...
*; *; Aoi, Sadanori*
PNC TN941 82-74VOL1, 151 Pages, 1982/03
PNC-TN941-82-74VOL1.pdf:7.53MB
A study of hypothetical core disruptive accidents (HCDAs) in the prototype fast breeder reactor Monju (714 MWt) has been conducted by using the SAS3D
accident analysis code. A loss-of-flow (LOF) due to the loss of off-site power and a transient overpower (TOP) due to control assembly withdrawal, both at rated power, are considered as the HCDA initiators with a postulated total failure of the reactor shutdown system. The accident scenarios of each postulated anticipated transient without scram are studied for the three burnup stages of Monju: the beginning-of-initial cycle (BOIC) ; a beginning-of-equilibrium cycle (BOEC); and an end-of-equilibrium cycle (EOEC). The neutronics data used in this study has been obtained by a 3-dimensional HEX-Z diffusion code and the first order perturbation calculations. The reactivity coefficients used in this study are the design nominal values without taking into account their uncertainties. The nominal design value of the maximum positive sodium void worth in Monju is a relatively small value of 2.5$ in the EOEC core. In the 2 cents/sec TOP, the reactor power shows a sudden increase following the onset of FCIs (Molten-Fuel/Coolant Interactions) in high-powered fuel assemblies but the maximum power level reached is less than 5 times the rated power and due to the fuel sweepout negative reactivity in the FCI fuel assemblies, the reactor is shutdown within 0.1 sec at the latest after the first FCI onset. The extent of damaged fuel assemblies is largest in the clean (FP-gas free) BOIC core in which the radial power peaking is smaller than in BOEC and EOEC cores, and about 17% of the fuel assemblies are damaged in the central region of the core. In the equilibrium cycle cores the damage extents are limited to about 5% core-center assemblies and this is smaller than in the BOIC core because of the larger radial power peaking and the rapid fuel sweepout reactivity insertion accelerated by the FP-gas pressure in the ...
*; *; *; *
PNC TN941 82-55, 284 Pages, 1982/03
The postdisassembly expansion phase of the Hypothetical Core Disruptive Accident (HCDA) in the MONJU reactor was analyzed by using the SIMMER-II code. Hitherto, the isentropic expansion of fuel vapor has been assumed after the core disassembly phase to estimate the system work energy following a postulated energetic disassembly. Recently, the SIMMER code was applied to analyze che postdisassembly expansion phase for the Clinch River Breeder Reactor (CRBR), and it was shown that the system work energy as a result of an HCDA was remarkably reduced compared with the isentropic expansion. The SIMMER code has attracted attension in the field of postdisassembly expansion analysis because of this possibility of work energy reduction. SIMMER-II was installed at PNC, the O-arai Engineering Center, in May 1980, and has been operational since November 1980. This report is divided into two parts. The first deals with the parameter survey based on the study of the MONJU postdisassembly expansion under simplified initial conditions by Kondo and Aizawa at Los Alamos National Laboratory. The other is based on the results of the initiating phase and of the core disassembly phase analyses by the SAS 3D and VENUS-PM codes, performed at PNC. In the latter, we adopted two cases which yielded largest system kinetic energy in the MONJU system, and we estimated the maximum energy released in the MONJU HCDA by using the SIMMER-II code. The main results obtained are shown below. (1)The maximum system kinetic energy released during the postdisassembly phase of the MONJU HCDA is at most 10 MJ when the active core, upper axial blanket and fission gas plenum are all voided at the initial state. (The maximum system work energy associated with isentropic expansion of the fuel vapor to 1 atm is 992 MJ.) (2)Under the same initial average core fuel temperature, a higher peaking factor of temperature distribution causes a larger system kinetic energy. For example, for temperature ...
Washiya, Tadahiro
no journal, ,
no abstracts in English
