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Ohgama, Kazuya; Kawashima, Katsuyuki*; Oki, Shigeo
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
In order to evaluate transient behavior of Japan sodium-cooled fast reactor (JSFR) with fuel sub-assemblies with the innerduct structure (FAIDUS) precisely, a new model for a plant dynamics code HIPRAC was developed. In this new model, inner core and outer core channels can be divided into three channels, respectively, such as interior, edge and near innerduct channel, and calculate coolant redistribution and coolant temperature in each channel. Coolant temperature distribution of interior and edge channels calculated by this model was compared with previous study by the general-purpose thermal-hydraulics code 
-FLOW. Coolant temperature behavior inside the innerduct was analyzed by a commercial thermal hydraulics code STAR-CD ver. 3.26. Based on this result, horizontally-uniformed coolant temperature in the innerduct was assumed as a heat transfer model of the innderduct. Reactivity coefficients for 750 MWe JSFR with low -decontaminated transuranic (TRU) fuel were evaluated. Transient behaviors of an unprotected loss-of-flow (ULOF) accident for JSFR with 750 MWe output calculated by previous and new models were compared. The results showed that the detailed evaluation of coolant temperature improved overestimation of the coolant temperature and coolant temperature feedback reactivity of the peripheral channels including coolant inside the innerduct and in the inter-wrapper gap.
Ioka, Ikuo; Kuriki, Yoshiro*; Iwatsuki, Jin; Kubo, Shinji; Inagaki, Yoshiyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
The thermochemical water-splitting (IS process) for hydrogen production has been developed by JAEA as application of a high-temperature gas cooled reactor. The IS process includes a severe corrosion environment which is made to boil and decompose concentrated sulfuric acid. Two kinds of brittleness materials, SiC and Fe-high Si alloy, are reported as materials having enough corrosion resistance in this corrosion environment. The two-layer pipe consisted of the Fe-high Si alloy with boiling sulfuric acid-resistant and the carbon steel with the ductility was produced by centrifugal casting. The evaluation of characteristics was carried out. The Fe-high Si alloy lining showed enough corrosion resistance in boiling sulfuric acid. As evaluation of the Fe / Fe-high Si alloy interface, thermal cycle test (100
C-900C) was executed. There was not the interface detachment and it was confirmed to have enough interfacial strength.
Saegusa, Hiromitsu; Onoe, Hironori; Kohashi, Akio; Watanabe, Masahisa
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company is facing contaminated water issues. The amount of contaminated water is continuously increasing due to groundwater leakage into the underground part of reactor and turbine buildings. Therefore, it is important to understand the groundwater flow conditions at the site and to predict the impact of countermeasures taken for isolating groundwater from the source of the contamination, i.e. the reactor buildings. Installations, such as of land-side and sea-side impermeable walls have been planned as countermeasures. In this study, groundwater flow modeling has been performed to estimate the response of groundwater flow conditions to the countermeasures. From the modeling, groundwater conditions and changes in response to implementation of the countermeasures could be reasonably estimated. The results indicate that the countermeasures will decrease the volume of inflow into underground part of the buildings. This means that the countermeasures will be effective in reducing the discharge volume of contaminated groundwater to ocean.
Otsuka, Noriaki; Matsui, Yoshinori; Tsuchiya, Kunihiko; Matsui, Tetsuya*; Arita, Setsuo*; Wada, Shohei*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
no abstracts in English
Saegusa, Jun; Tagawa, Akihiro; Kurikami, Hiroshi; Iijima, Kazuki; Yoshikawa, Hideki; Tokizawa, Takayuki; Nakayama, Shinichi; Ishida, Junichiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
After the Fukushima nuclear accident, JAEA lead off demonstration tests to find out effective decontamination methods for various school facilities in Fukushima. It included (1) dose reduction measures at schoolyards, (2) purification of swimming pool water and (3) removal of surface contamination of playground equipments. Through these tests, they established practical methods suitable for each situation; (1) At school yards, dose rates were drastically reduced by removing topsoil which was then placed in trenches of 1 m deep; (2) For the purification of pool water, the flocculation-coagulation treatment was found to be effective for collecting radiocesium dissolved in the water; (3) Demonstration tests for playground equipments, such as horizontal bars and a sandbox wood frame, suggested that effectiveness of decontamination considerably varied depending on the material, paint or coating condition. This paper reviews these demonstrations.
Tanaka, Shingo*; Yokota, Hideharu; Ono, Hirokazu; Nakayama, Masashi; Fujita, Tomoo; Takiya, Hiroaki*; Watanabe, Naoko*; Kozaki, Tamotsu*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
Yamada, Fumiaki; Arikawa, Mitsuhiro*; Fukano, Yoshitaka
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
In sodium-cooled fast reactor, since the coolant does not need to be pressurized, a pipe break due to the internal pressure does not occur physically. For safety margin in Japanese prototype fast breeder reactor (Monju), the guillotine pipe break accident, i.e., loss of integrity (LOPI) has been analyzed as an extreme assumption for beyond design basis accidents (B-DBAs) in the licensing application for the permit. The cooling capability of the core was re-evaluated in this paper during a large-scale, more specifically guillotine pipe break at the primary heat transport system (PHTS) in Monju, newly considering the following latest findings: (a) Experimental data on sodium boiling in fuel assemblies, (b) Actual PHTS pump coast-down characteristics, and (c) Transient burst test data on irradiated fuel claddings. The analysis models were the validated and simulations were re-performed also using the actual Monju data such as the response time to the trip signals, etc. As a result, it was clarified that the ratio of failed fuel claddings does not exceed around 3% of all of fuel assemblies, as in the past licensing analysis. The safety has been reconfirmed to be secured without significant core damage even under an extreme assumption of a double-ended guillotine pipe break at the PHTS in Monju.
Kurikami, Hiroshi; Kitamura, Akihiro; Yamada, Susumu; Machida, Masahiko
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
Several numerical models have been prepared to deal with various time- and spatial-scale issues related to radioactive cesium migration in environment in Fukushima area. This paper describes fragments of the JAEA's approaches of modeling to deal with the issues corresponding to radioactive cesium migration in environment with some case studies.
Honda, Yuki; Fujimoto, Nozomu; Sawahata, Hiroaki; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
The operating data of the HTTR with burn-up is very important for developments of HTGRs. There is a temperature distribution in a core in full power operation. The temperature distribution in a core makes it difficult to validate of calculation code. On the other hands, the data of the control rod position at criticality at zero power have been measured at the beginning of each operation cycle. The temperature distribution in a core at zero power is uniform. Therefore, the data at zero power are suitable for validation of calculation code. In this study, the calculated control rod positions at zero power criticality with burn-up are compared with the experimental data with correlation of core temperature. The calculated results of criticality control rod position at zero power show good agreement to the experimental data. It means that calculated result shows appropriate decrease in uranium and accumulation in plutonium decrease in burnable absorber with burn-up.
Honda, Yuki; Fujimoto, Nozomu; Sawahata, Hiroaki; Sawa, Kazuhiro
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
In the HTTR, a two-step control rods insertion method for reactor scram is adopted. In the method, control rods at reflector region are inserted at the scram signal is initiated. The core should keep its subcriticality by reflector region control rods. Therefore, precise evaluation of control rods reactivity worth for reflector region is necessary. However, all cross section of control rods has been prepared for control rod in fuel region because the reactivity value of control rods in the fuel region is larger than that of control rods in the reflector region. This paper proposed the revised method of preparing the control rod cross section for first step control rod in reflector region.
Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi; Okano, Yasushi; Sakai, Takaaki; Yamamoto, Takahiro*; Ishizuka, Yoshihiro*; Geshi, Nobuo*; Furukawa, Ryuta*; Nanayama, Futoshi*; et al.
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 10 Pages, 2015/05
This paper describes mainly snow margin assessment methodology development in addition to the project overview. For the snow margin assessment, the index is a combination of a snowfall rate and duration. Since snow removal can be expected during the snowfall, the developed snow margin assessment methodology is such that the margin was regarded as the snowfall duration up to the decay heat removal failure which was defined as when the snow removal rate was smaller than the snowfall rate.
Iguchi, Masahide; Sakurai, Takeru; Nakahira, Masataka; Koizumi, Norikiyo; Nakajima, Hideo
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
Application of partial penetration welding (PPW) to ITER Toroidal Field Coil structure has been proposed because of limited accessability for weld due to complex geometry and low stress and low importance components. In order to obtain fatigue crack growth (FCG) behavior of PPW joint in cryogenic environment, Japan Atomic Energy Agency performed FCG test at 4K by using Compact Tension (CT) specimens having as-weld notch of PPW. These CT specimens were made from mockups having one of actual joint shape of PPW, double J-groove. As the result of this test, it was observed that crack propagated in weld metal having inclination from as-weld notch. Moreover it was shown that FCG rate of as-weld CT specimens had high FCG rate region in early stage of crack propagation due to residual stress distribution. In addition, application method of this FCG rate to designing of PPW joint was proposed and verified in this study.
Itabashi, Kiyoshi; Tagawa, Akihiro; Sugiyama, Kenji; Yamamoto, Tomoyo
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
Kotaeru kai for questions about radiation (below is called Kotaeru kai) has started since July 2011 in Fukushima prefecture which influenced especially after the accident of Fukushima Daiichi Nuclear Power Station on 11st March 2011. The purpose is to have mainly parents and educators (such as of kindergartens, nursery schools etc.) understand correctly about radiation and the influence to health. Receiving requests from educators in Fukushima prefecture, about 4 staff members teamed up to explain. They were selected from 500 JAEA staffs entered beforehand. The members explain about radiation and the influence to health by using illustrations and metaphors in materials and answer questions asked in advance at schools or so which require explain. When they explain, they place much vale for communication with participants. Until the end of September 2014, Kotaeru kai have enforced 237 times for about 20 thousand of participants. According to 7,613 participants' questionnaires (which were collected from July 2011 to the end of 2012), it seems that participants understood well about radiation and the influence to health. Certain junior high school requested to explain for students, and JAEA will continue this activity in order to meet these expectations and requirements.
Nemoto, Yoshiyuki; Kato, Hitoshi; Kaji, Yoshiyuki; Yoshida, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
For the evaluation of reactor pressure vessel (RPV) lower head rupture probably occurred during the severe accident in Fukushima Daiichi Nuclear Power Plants, JAEA is conducting the thermal-hydraulics / mechanical coupling analysis. In the mechanical analysis based on the finite element method (FEM), material property data previously obtained from uni-axial material tests are applied. The lower head of BWR such as Fukushima NPP, has complicated structure compared to PWR, with control rod guide tubes, stub tubes, etc., therefore the mechanical analyses need to treat multi-axial deformation of the materials. To perform such mechanical analysis, the applicability of the analytical model using uni-axial data for multi-axial deformation analysis must be validated. In this study, the internal pressure creep tests were performed because which can realize the multi-axial deformation condition. In addition, mechanical analyses were conducted and the analytical results were compared with the experimental data.
Katsuyama, Jinya; Yamaguchi, Yoshihito; Kaji, Yoshiyuki; Yoshida, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
In existing severe accident codes, rupture of reactor pressure vessel (RPV) by relocated molten core is judged using simple models. However, it is difficult to assess rupture behavior of the lower head of RPV in boiling-water-type nuclear power plants due to severe accident like Fukushima Daiichi since boiling water reactors (BWRs) have geometrically complicated structure with a lot of penetrations. Therefore, we have been developing an analysis method to predict time and location of RPV lower head rupture of BWRs. The detailed three-dimensional model of RPV lower head is constructed. Using temperature distribution in relocated debris obtained from thermal-hydraulics analysis, structural analysis is carried out to evaluate creep damage distributions using damage criterions of "considere", strain, Kachanov, and Larson-Miller-parameter criteria. It is shown that failure regions of BWR lower head are only penetrations, although there is a large difference in damaged time.
Yamaguchi, Yoshihito; Katsuyama, Jinya; Kaji, Yoshiyuki; Yoshida, Hiroyuki; Li, Y.
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
Kasahara, Seiji; Murata, Tetsuya*; Kamiji, Yu; Terada, Atsuhiko; Yan, X.; Inagaki, Yoshiyuki; Mori, Michitsugu*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 11 Pages, 2015/05
A heat transport analysis of a district heating and snow melting system in Sapporo and Ishikari, Hokkaido was carried out assuming application of waste heat from GTHTR300, a design of high temperature gas-cooled reactor. The following components in the system were modeled; pipelines of the water loops between GTHTR300 and heat demand district and heat exchangers to transport the heat from the water loops to water loops in the district. Double pipes for the pipeline has disadvantage that pumping electricity consumption was 2.74 times to that of single pipes due to pressure loss in annulus channel. On the other hand, the double pipe was advantageous in less heat loss and excavation load. Heat loss was 33% smaller because heat loss from inner tube was recovered in annulus channel. Excavation area was 23% smaller because water loop was made by one double pipe. Total heat loss from the GTHTR300s to the water loop in the district was 4.2% and ratio of pump electricity to power generation from the GTHTR300s was 0.8%. In January, the maximum heat demand in a year, 97.0% of the heat demand was supplied by 2 GTHTRs. Less distance between GTHTR300 and heat demand district from 40 km to 20 km would make cost of the heat transfer system 22% smaller.
Sakamoto, Atsushi; Sano, Yuichi; Takeuchi, Masayuki; Okamura, Nobuo; Koizumi, Kenji
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
Kamide, Hideki; Ando, Masato*; Ito, Takaya*
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05
JAEA, JAPC and MFBR have been conducted design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lesson learned from the TEPCO's Fukushima Dai-ichi Nuclear Power Plants accident, in the frame work of generation IV international forum (GIF), the design study is focusing on the design measures against sever external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated.
Yamashita, Susumu; Takase, Kazuyuki; Yoshida, Hiroyuki
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 5 Pages, 2015/05
In accidents of the Fukushima Daiichi Nuclear Power Plants, by stop of the emergency core cooling system, fuel rods were overheated due to the radioactive decay heat and the oxidization of fuel cladding. Although it is inferred that the core degradation occurred, condition inside the core still has not been revealed. Especially, in order to precisely understand the accumulation condition of debris in lower plenum, detailed and phenomenological relocation process of molten fuel is quite important. In this problem, since an experiment is extremely difficult, numerical simulation will be useful tool for investigating conditions in reactor core. However, existing codes can not be phenomenologically treated relocations process. Therefore a phenomenologically-based numerical simulation method for predicting the melting core behavior including solidification and relocation based on the computational fluid dynamics has been developed in JAEA. Last paper, ICONE 23, in order to distinguish a fuel component which possesses a heat source such as a decay heat from structures which does not possess a heat source such as a core plate, control rod guide tubes and so on, we developed three phases and three component multiphase flow simulation code and performed preliminary analysis of molten core relocation behavior using simplified core structures. As a result, we obtained reasonable results. However, we have not carried out validations for the numerical code yet. In this paper, we show that the results of the numerical test for evaluating the validity of the numerical code and also show that the development of the radiation heat transfer model and its preliminary analysis. In addition, we will report the preliminary analysis of the relocation behavior of molten materials in the simplified core support structure.
