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Yamano, Hidemasa; Vasile, A.*; Kang, S.-H.*; Summer, T.*; Tsige-Tamirat, H.*; Wang, J.*; Ashurko, I.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 "Methods, Models and Codes", WP-SO-2 "Experimental Programs and Operational Experience", and WP-SO-3 "Studies of Innovative Design and Safety Systems". This paper reports recent activities within the SO project.
Matsunaga, Shoko*; Matsubara, Shinichiro*; Kato, Atsushi; Yamano, Hidemasa; D
derlein, C.*; Guillemin, E.*; Hirn, J.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
This paper presents a design of Curie Point Electro-Magnet (CPEM) which will be installed as a passive shutdown system for a French Sodium-cooled Fast Reactor (ASTRID) development program which is conducted in collaboration between France and Japan. To confirm CPEM design validity, a qualification program for CPEM is developed on the basis of past comprehensive test series of Self-Actuated Shutdown System (SASS) in Japan. The main outcome of this paper is results of holding force tests in hot gas, which satisfy design requirements. Moreover, the result of a numerical magnetic field analysis showed the same tendency as that of the holding force test.
Matsuo, Eiji*; Sasa, Kyohei*; Koyama, Kazuya*; Yamano, Hidemasa; Kubo, Shigenobu; Hourcade, E.*; Bertrand, F.*; Marie, N.*; Bachrata, A.*; Dirat, J. F.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05
Discharged molten-fuel from the core during Core Disruptive Accident (CDA) could become solidified particle debris by fuel-coolant interaction in the lower sodium plenum, and then the debris could form a bed on a core catcher located at the bottom of the reactor vessel. Coolability evaluations for the debris bed are necessary for the design of the core catcher. The purpose of this study is to evaluate the coolability of the debris bed on the core catcher for the ASTRID design. For this purpose, as a first step, the coolability calculations of the debris beds formed both in short term and later phase have been performed by modeling only the debris bed itself. Thus, details of core catcher design and decay heat removal system are not described in this paper. In all the calculations, coolant temperature around the debris bed is a parameter. The calculation tool is the debris bed module implemented into a one-dimensional plant dynamics code, Super-COPD. The evaluations have shown that the debris beds formed both in short term and later phase are coolable by the design which secures sufficient coolant flow around the core catcher located in the cold pool.
Yamano, Hidemasa; Tobita, Yoshiharu
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11
Based on the event tree analysis, the present numerical analyses investigated the capability of fuel discharge through the one-dimensional single fuel assembly geometry and the two-dimensional geometry of a CRGT channel with neighboring fuel assemblies. The single fuel assembly analyses showed that the fuel blockage formed in the lower shielding region because fuel solidified by contacting with cold sodium in case of no fission gas release. On the assumption that fission gas was released, the molten fuel successfully relocated below the core. The next analyses using the CRGT channel indicated a significant fuel discharge through the CRGT channel. This is because the fuel temperature was still high just after the CRGT wall failure and sodium in the CRGT channel was quickly voided just after the ingress of a small amount of molten fuel.
Yamano, Hidemasa; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi; Naruto, Kenichi*
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 15 Pages, 2018/10
Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure, which was achieved through probabilistic risk assessment for the EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. The safety strategy for the EVST involves whole core refueling (early transfer of all core fuel assemblies into the EVST) assuming a severe situation that results in sodium level reduction leading finally to the top of the reactor core fuel assemblies in a long time. This study introduces the success criteria mitigation along the decay heat decrease over time. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, a probability analysis for human error, and quantification of accident sequences. The fuel damage frequency of the EVST was evaluated to be approx. 10
/year. The dominant accident sequence resulted from the static failure and human error for the switching from the stand-by to operation mode in the three stand-by cooling circuits after loss of one circuit for refueling heat removal operation as an initiating phase.
C for severe accident simulationsLiu, X.*; Morita, Koji*; Yamano, Hidemasa
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10
On the basis of experimental results, growth of the eutectic material is modeled by the parabolic rate law. Heat and mass transfer processes are also modeled considering both the equilibrium and non-equilibrium phase changes of eutectic material. Thermophysical properties of eutectic material obtained from the experimental measurements are also included in the analytic thermophysical property model and analytic equation-of-state model.
Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamamoto, Takahiro*
ASCE-ASME Journal of Risk and Uncertainty in Engineering Systems, Part B; Mechanical Engineering, 4(3), p.030902_1 - 030902_9, 2018/09
This paper describes volcanic probabilistic risk assessment (PRA) methodology development for sodium-cooled fast reactors. The volcanic ash could potentially clog air filters of air-intakes that are essential for the decay heat removal. The degree of filter clogging can be calculated by atmospheric concentration of ash and tephra fallout duration and also suction flow rate of each component. The atmospheric concentration can be calculated by deposited tephra layer thickness, tephra fallout duration and fallout speed. This study evaluated a volcanic hazard using a combination of tephra fragment size, layer thickness and duration. In this paper, each component functional failure probability was defined as a failure probability of filter replacement obtained by using a grace period to a filter failure limit. Finally, based on an event tree, a core damage frequency was estimated about 3
10
/year in total by multiplying discrete hazard probabilities by conditional decay heat removal failure probabilities. A dominant sequence was led by the loss of decay heat removal system due to the filter clogging after the loss of emergency power supply. In addition, sensitivity analyses have investigated the effects of a tephra arrival reduction factor and pre-filter covering.
Kokubo, Hiroki*; Nishi, Tsuyoshi*; Ota, Hiromichi*; Yamano, Hidemasa
Nippon Kinzoku Gakkai-Shi, 82(10), p.400 - 402, 2018/09
Percentile：100(Metallurgy & Metallurgical Engineering)It is important to obtain the viscosity of a mixed alloy consisting of molten stainless steel and boron-carbide (SUS316L + BC alloy) for the improvement of severe accident assessment methodology for sodium-cooled fast reactors. In this study, the viscosities of the molten nickel (Ni) and stainless steel (SUS316L) were measured by the oscillating crucible method to confirm the performance of the viscosity measurement apparatus as a first step. The viscosities of molten Ni and SUS316L melts were measured up to 1823 K. It was found that the measured viscosity values of molten Ni and SUS316L were estimated from the deviation of the experimental data, were 
4% and 3%, respectively. It was also found that those of molten Ni and SUS316L were close to those of the literature values of molten Ni and similar composite stainless steels. Moreover, we tentatively measured the viscosity of molten SUS316L-5 mass%BC alloy. The fitted results of the viscosity for molten Ni and SUS316L were obtained.
Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi
Mechanical Engineering Journal (Internet), 5(4), p.18-00093_1 - 18-00093_19, 2018/08
This paper describes the development of a probabilistic risk assessment (PRA) methodology against a combination hazard of strong wind and rainfall. In this combination hazard PRA, a hazard curve is evaluated in terms of maximum instantaneous wind speed, hourly rainfall, and rainfall duration. A scenario analysis has provided event sequences resulting from the combination hazard of strong wind and rainfall. The typical event sequence was characterized by the function loss of auxiliary cooling system, of which heat transfer tubes could crack due to cycle fatigue caused by cyclic contacts with rain droplets. This cycle fatigue crack could occur if rain droplets enter into the air cooler of the system following the coolers roof failure due to strong-wind-generated missile impact. This event sequence has been incorporated into an event tree which addresses component failure caused by the combination hazard. As a result, a core damage frequency has been estimated to be about 10/year in total by multiplying discrete hazard frequencies by conditional decay heat removal failure probabilities. The dominant sequence is the manual operation failure of an air cooler damper following the failure of external fuel tank due to the missile impact. The dominant hazard is the maximum instantaneous wind speed of 20-40 m/s, the hourly rainfall of 20-40 mm/h, and the rainfall duration of 0-10 h.
Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi
Mechanical Engineering Journal (Internet), 5(4), p.18-00079_1 - 18-00079_17, 2018/08
Nishida, Akemi; Choi, B.; Yamano, Hidemasa; Takada, Tsuyoshi*
Proceedings of 2018 ASME Pressure Vessels and Piping Conference (PVP 2018), 11 Pages, 2018/07
This study identified and quantified possible cliff edge effects through a seismic safety evaluation of a nuclear power plant, based on the concepts of risk and defense in depth. Cliff edges of the both physical and knowledge-based type were considered in this study. We investigated a seismic isolation effect, etc., for physical cliff edges, and the modeling of the target structure, boundary conditions, etc., for knowledge-based cliff edges. Response analysis was performed using a sway-rocking (SR) model and a three-dimensional model of the target building. The seismic isolation effect of the base-isolated building was confirmed by comparison to the results of earthquake-resistant building. In the case of a collision with the retaining wall of the base-isolated building, the level of damage was found to depend on the modeling of the collision condition assumed. On the other hand, the study confirmed the differences between the results from the SR model and the three-dimensional model.
Yamano, Hidemasa; Naruto, Kenichi*; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi
Proceedings of 26th International Conference on Nuclear Engineering (ICONE-26) (Internet), 9 Pages, 2018/07
Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure by conducting probabilistic risk assessment for EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, human error probability analysis, and quantification of accident sequences. Fuel damage frequency of the EVST was evaluated approx. 10 /year in this paper. By considering the secondary sodium freezing, the fuel damage frequency was twice increased. The dominant accident sequence resulted from the common cause failure of the damper opening and/or the human error for the switching from the stand-by to the operation mode in the three stand-by cooling circuits. The importance analyses have indicated high risk contributions.
Kikuchi, Shin; Yamano, Hidemasa
Dai-23-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2018/06
In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic reaction between boron carbide (BC) and stainless steel (SS) may probably occur. Elucidation on the behavior of cited eutectic reaction is very important in terms of evaluation of core disruptive accidents in SFRs. For the first step to clarify the kinetic feature of BC-SS eutectic reaction, the preliminary thermogavimetry-differential thermal analysis (TG-DTA) measurements using individual reagent were performed to obtain the fundamental information and to confirm the applicability of sample crucibles. It was found that alumina crucible was applicable in terms of eutectic behavior. Based on the DTA curves at different heating rates, the kinetic parameters were roughly estimated by using Kissinger method.
C in the solid phaseTakai, Toshihide; Furukawa, Tomohiro; Yamano, Hidemasa
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1007 - 1013, 2018/04
CFukuyama, Hiroyuki*; Higashi, Hideo*; Yamano, Hidemasa
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1014 - 1019, 2018/04
In this study, densities, surface tensions, normal spectral emissivities, heat capacities and thermal conductivities of molten SUS316L and SUS316L containing 5mass%-BC were measured by the electromagnetic levitation technique in a static magnetic field.
Okano, Yasushi; Yamano, Hidemasa
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 3 Pages, 2017/11
Forest fire hazard assessment methodologies using a logic tree have been applied for the evaluation of combination hazard curves of a forest fire with lightning as an initiator of a forest fire and with a strong wind being independent from a forest fire. The complex shape of the combinational hazard curve of forest fire and lighting is due to that both lightning and high velocity wind tend to appear under unstable weather conditions, and there is correlation between two hazards. The evaluated combinational hazard curve of forest fire and strong wind for the instantaneous wind velocity over 80 m/s has extremely small frequency in the range below 10
/year.
Yamano, Hidemasa; Naruto, Kenichi*; Kurisaka, Kenichi; Nishino, Hiroyuki; Okano, Yasushi
Proceedings of Asian Symposium on Risk Assessment and Management 2017 (ASRAM 2017) (USB Flash Drive), 3 Pages, 2017/11
Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan (JSFR). The objective of this study is to identify dominant accident sequences leading to fuel failure by conducting PRA for EVST. The EVST heat removal system in JSFR consists of four independent loops with for primary and secondary ones. Based on the JSFR design information, this study has identified initiating events, event and /fault tree analyses, human reliability analysis, and quantification of accident sequences. Fuel damage frequency of the EVST was evaluated approx. 10 /year in this paper. The main contributor of the fuel damage frequency is the loss of heat removal function of the cooling system. The dominant initiating event was the loss of one circuit of normal heat removal operation.
Okano, Yasushi; Yamano, Hidemasa
Proceedings of International Topical Meeting on Probabilistic Safety Assessment and Analysis (PSA 2017) (USB Flash Drive), p.1334 - 1342, 2017/09
This study evaluates a hazard curve of smoke effects generated by a forest fire by applying a new method using a logic tree which consists of variable parameters on a forest fire, weather conditions, types of vegetation and topography, and simulation conditions. A response surface of the smoke spatial density was evaluated using two simulation codes: FARSITE for forest fire propagation and ALOFT-FT for smoke transportation. It is followed by a Monte Carlo simulation to evaluate the hazard curve representing the annual exceedance frequency of the total amount of the smoke captured on air filters. The evaluated hazard curve is about 110
per year for 3.5 kg/m
/(m/s).
Nishida, Akemi; Choi, B.; Yamano, Hidemasa; Takada, Tsuyoshi*
Transactions of 24th International Conference on Structural Mechanics in Reactor Technology (SMiRT-24) (USB Flash Drive), 9 Pages, 2017/08
In this research, the seismic safety of nuclear power plants (NPP) is treated as a system in which the various cliff edge effects are identified and quantified based on the concepts of risk and defense in depth. A methodology is then developed for avoiding these cliff edge effects. The first step was to carry out a preliminary elastic-plastic analysis of the NPP building system. From the analysis, some knowledge was obtained for the modeling factor dependence of cliff edge effects. Next, a preliminary fragility evaluation of the reactor vessel and piping was carried out; it was found that introducing a horizontal seismic isolation system was very effective for avoiding the cliff edge.
Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
