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Journal Articles

Development of instrumentation systems for severe accidents, 5; Basic properties of hydrogen sensor with solid electrolyte for safety measures of LWRs

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

Journal Articles

Development of numerical simulation method for relocation behavior of molten materials in nuclear reactors; Relocation behavior in a simplified core structures

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.

Journal Articles

Development research of corrosion-resistant structural material using Fe-Si alloy lining centrifugal cast-iron for thermochemical water-splitting iodine-sulfur process

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$$^{circ}$$C-900$$^{circ}$$C) was executed. There was not the interface detachment and it was confirmed to have enough interfacial strength.

Journal Articles

Development of failure evaluation method for BWR lower head in severe accident, 2; Applicability evaluation of the FEM using uni-axial material data for multi-axial deformation analysis

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.

Journal Articles

Seismic response simulation of High-Temperature Engineering Test Reactor building against 2011 Tohoku earthquake

Nishida, Akemi; Nakajima, Norihiro; Kawakami, Yoshiaki; Iigaki, Kazuhiko; Sawa, Kazuhiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

The R&D on the three dimensional vibration simulation technologies for a nuclear facility is one of missions of Center for Computational Science and e-Systems, Japan Atomic Energy Agency. Until now, three dimensional building and equipment models of HTTR (High Temperature Engineering Test Reactor) have been constructed and been performed validation of the models by comparison with seismic observed records. In this report, the results obtained by seismic observation simulation on the Tohoku earthquake occurred in the 3/11/2011 using three dimensional models of the HTTR building are shown. The simulation results show good agreement with the real observation data.

Journal Articles

Modeling approach to various time and spatial scale environmental issues in Fukushima; Related to radioactive cesium migration in aquatic systems

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.

Journal Articles

JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors, 4; Balance of plant

Chikazawa, Yoshitaka; Kato, Atsushi; Nabeshima, Kunihiko; Otaka, Masahiko; Uzawa, Masayuki*; Ikari, Risako*; Iwasaki, Mikinori*

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

Design study and evaluation for SDC and safety SDG on the BOP of the demonstration JSFR including fuel handling system, power supply system, component cooling water system, building arrangement are reported. For the fuel handling system, enhancement of storage cooling system has been investigated adding diversified cooling systems. For the power supply, existing emergency power supply system has been reinforced and alternative emergency power supply system is added. For the component cooling system and air conditioning, requirements and relation between safety grade components are investigated. Additionally for the component cooling system, design impact when adding decay heat removal system by sea water has been investigated. For reactor building, over view of evaluation on the external events and design policy for distributed arrangement is reported. Those design study and evaluation provides background information of SDC and SDG.

Journal Articles

Seismic analysis of the ITER blanket remote handling system

Noguchi, Yuto; Maruyama, Takahito; Takeda, Nobukazu; Kakudate, Satoshi

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 9 Pages, 2015/05

This paper reports the seismic analysis of the ITER Blanket RH system (BRHS) during blanket module handling operation. Since the BRHS, which is composed of the articulated rail and the vehicle manipulator, which travels on the rail deployed in the vacuum vessel in the toroidal direction, has various configurations and the rail system has flexibility, evaluation of dynamic response of the system is of essential importance. Via parameter sensitivity study on position and posture of the vehicle manipulator, the most unfavorable configuration for each component of the BRHS has been specified by the modal and spectrum analyses with the global BRHS FE model. Then using the quasi-static equivalent loads on the individual components obtained by the global BRHS seismic analysis, the structural verifications of the structural members of the BRHS have been carried out with detailed partial FE models. The system seismic resistance of the BRHS to a safe shutdown earthquake was confirmed.

Journal Articles

Mechanical properties of fuel debris for defueling toward decommissioning

Hoshino, Takanori; Kitagaki, Toru; Yano, Kimihiko; Okamura, Nobuo; Ohara, Hiroshi*; Fukasawa, Tetsuo*; Koizumi, Kenji

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05

Journal Articles

HTTR demonstration program for nuclear cogeneration of hydrogen and electricity

Sato, Hiroyuki; Sumita, Junya; Terada, Atsuhiko; Ohashi, Hirofumi; Yan, X.; Nishihara, Tetsuo; Tachibana, Yukio; Inagaki, Yoshiyuki

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

This paper explains the outline and schedule of HTTR demonstration program with a plant concept of the heat application system directed at establishing an HTGR cogeneration system with 950$$^{circ}$$C reactor outlet temperature for production of power and hydrogen as recommended by the task force.

Journal Articles

Radioactivity decontamination in and around school facilities in Fukushima

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.

Journal Articles

Dipole tracer migration and diffusion tests in fractured sedimentary rock at Horonobe URL

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

Journal Articles

Validation of core cooling capability analysis in Monju during guillotine pipe break at primary heat transport system

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.

Journal Articles

Sludge behavior in centrifugal contactor operation for nuclear fuel reprocessing

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

Journal Articles

Burn-up dependency of control rod position at zero power criticality in the high temperature test engineering reactor

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.

Journal Articles

Improvement of cell model for control rod in reflector region of high temperature test engineering reactor

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.

Journal Articles

Development of risk assessment methodology of decay heat removal function against external hazards for sodium-cooled fast reactors, 1; Project overview and margin assessment methodology against snow

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.

Journal Articles

Welding joint design of ITER toroidal field coil structure under cryogenic environment

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.

Journal Articles

Development of failure evaluation method for BWR lower head in severe accident, 3; Creep damage evaluation based on thermal-hydraulics and structural analyses

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.

Journal Articles

Development of failure evaluation method for BWR lower head in severe accident, 1; High temperature creep test and creep damage model

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

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