Watanabe, Masashi; Kato, Masato; Sunaoshi, Takeo*
Journal of Nuclear Materials, 542, p.152472_1 - 152472_7, 2020/12
The oxygen self-diffusion coefficients in near stoichiometric (U,Pu)O at high temperatures were successfully measured by thermogravimetry combined with the oxygen isotope exchange method. The activation energy for oxygen diffusion in the stoichiometric composition of (U,Pu)O was evaluated from experimental data, and the value was determined to be 248 kJ/mol. In addition, the defect migration energies of (U,Pu)O were derived, and the oxygen self-diffusion coefficients were evaluated using these. As a result, good agreement was found between the experimental data and the oxygen self-diffusion coefficients calculated using the defect migration energies.
Negishi, Hitoshi; Kamide, Hideki; Maeda, Seiichiro; Nakamura, Hirofumi; Abe, Tomoyuki
Nippon Genshiryoku Gakkai-Shi, 62(8), p.438 - 441, 2020/08
Prototype Fast Breeder Reactor, Monju, was under decommission since April, 2018. It is the first time for Japan to make a sodium cooled reactor into decommission. It is significant work and will take 30 years. The Monju has provided wide spectrum and huge amount of findings and knowledge, e.g., design, R&D, manufacturing, construction, and operation up to 40% of full power over 50 years of development history. It is significant to utilize such findings and knowledge for the development and commercialization of a fast rector in Japan.
Takata, Takashi*; Yamano, Hidemasa; Narumiya, Yoshiyuki*
Nippon Genshiryoku Gakkai-Shi, 62(8), p.448 - 451, 2020/08
This report describes risk analysis applicable to selection and evaluation of external hazards in risk assessment. Volcanic ash hazard evaluation is shown as one of hazard evaluations. The objective of the risk assessment is to secure and improve the safety of nuclear installation. This report also describes discussion on the process of response informed based on risk evaluation against external hazards.
Watanabe, So; Sano, Yuichi; Okada, Makoto*; Matsuura, Haruaki*; Hagura, Naoto*; Kada, Wataru*
Nuclear Instruments and Methods in Physics Research B, 477, p.60 - 65, 2020/08
IBIL and EXAFS analyses were applied on strucutral analysis of Eu complex formed in adsorbent developed for extraction chromatography. Those analyses revealed slight structural difference between adsorbent and solvent systems.
Sakamoto, Naoki; Fujishima, Tadatsune; Mizukoshi, Yasutaka
Hozengaku, 19(2), p.125 - 126, 2020/07
The five post-irradiation examination facilities in JAEA's Oarai research and development institute have been operated for over 40 years in order to investigate the irradiation performance of fast reactor fuel materials. The equipment associated with these facilities has been managed to maintain secure from the problems occurred in the process of aging. Therefore, we established a safety assessment method for aging facilities in 2002, and we have been conducting maintenance management of facilities since then. In this study, improvement plans of the safety assessment method are considered in order to solve the issues detected as a result of analysis of past maintenance information.
Doda, Norihiro; Hamase, Erina; Yokoyama, Kenji; Tanaka, Masaaki
Dai-25-Kai Nippon Keisan Kogaku Koenkai Rombunshu (CD-ROM), 4 Pages, 2020/06
With the aim of advancing the design optimization in fast reactors, neutronics and thermal-hydraulics coupled analysis method which can consider the temporal change of neutron flux distribution in the core has been developed. A three-dimensional neutronics analysis code and a plant dynamics analysis code are coupled on a platform using Python programing. In this report, outlines of the coupling method of analysis codes, the results of its application to the actual plant under a virtual accidental condition, and the future development is described.
Ishida, Shinya; Kawada, Kenichi; Fukano, Yoshitaka
Mechanical Engineering Journal (Internet), 7(3), p.19-00523_1 - 19-00523_17, 2020/06
The Phenomena Identification and Ranking Table (PIRT) approach was applied to the validation of SAS4A code in order to indicate the reliability of SAS4A code sufficiently and objectively. Based on this approach, issue and objective were clarified, plant design and scenario were defined, FOM and key phenomena were selected, and the code validation test matrix was completed with the results of investigation about analysis models and test cases. The results of the test analysis corresponding to this matrix show that the SAS4A models required for the IP evaluation were sufficiently validated. Furthermore, the validation with this matrix is highly reliable, since this matrix represents the comprehensive validation that also considers the relation between physical phenomena. In this study, the reliability and validity of SAS4A code were significantly enhanced by using PIRT approach to the sufficient level for CDA analyses in SFR.
Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki
Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06
Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.
Takaya, Shigeru; Fujisaki, Tatsuya*
Mechanical Engineering Journal (Internet), 7(3), p.19-00526_1 - 19-00526_10, 2020/06
In severe seismic conditions, sloshing waves are considered to even reach a roof slab of a reactor vessel. The structural integrity of roof slabs is required to be evaluated against sloshing impacts. However, there is no widely recognized evaluation method for sloshing impact pressure on flat roofs yet. Therefore, in this paper, a simplified evaluation method is proposed based on Wagner's theory, which is a well-known classic theory for evaluating impact pressures on rigid wedges dropping on water surfaces. In the proposed method, we assume an equivalent wedge on a flat roof. The impact pressure on the equivalent wedge is evaluated by applying Wagner's theory. Computational fluid dynamics analysis is conducted to confirm that a key assumption of Wagner's theory is applicable to the evaluation of sloshing impact on a flat roof. In addition, the predictability of the proposed method is investigated by comparing literature data of sloshing experiments with the estimated values.
Takaya, Shigeru; Sasaki, Naoto*
Mechanical Engineering Journal (Internet), 7(3), p.19-00549_1 - 19-00549_9, 2020/06
Seismic buckling of vessels is one of main concerns for the design of fast reactor plants in Japan. In previous studies, we discussed evaluation methods of seismic buckling probability of vessels by taking account of seismic hazards in order to rationalize seismic buckling evaluation, and proposed a rule for seismic buckling of vessels based on the load and resistant factor design method. The proposed method deals with only seismic load, but in actuality, dead weight and internal pressure also exist. In this study, the rule was expanded so that dead weight and internal pressure can be taken into account. Furthermore, the influences of dead weight and internal pressure to seismic buckling evaluation were discussed. As result, it was shown that approximately 10 to 20% of further rationalization of allowable seismic load could be achieved by considering dead weight and internal pressure in the evaluation.
Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Li, J.*; Jang, S.*
Mechanical Engineering Journal (Internet), 7(3), p.19-00548_1 - 19-00548_11, 2020/06
Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium. To improve the evaluation accuracy for the temperature distribution, a Lagrangian particle model for simulating reacting jet was also developed as an alternative method and its basic function was confirmed.
Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Uchita, Masato*; Miyagawa, Takayuki*; Eto, Masao*; Suzuno, Tetsuji*; Matoba, Ichiyo*; Endo, Junji*; Watanabe, Osamu*; et al.
Mechanical Engineering Journal (Internet), 7(3), p.19-00489_1 - 19-00489_16, 2020/06
The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.
Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki
Transactions of the American Nuclear Society, 122(1), p.862 - 865, 2020/06
Development of a new simulation system SPECTRA has been started to enable a simulation of comprehensive in- and ex-vessel events during a severe accident in a sodium-cooled fast reactor. The simulation system SPECTRA consists of two basic modules of thermal-hydraulics; in-vessel basic module and ex-vessel one, and some sub-modules for specific phenomena or events. A sodium fire models are implemented as one sub-module of the ex-vessel module. The sodium fire models are adapted from existing sodium fire analysis codes AQUA-SF and SPHINCS. As the result of verification test, the results show good agreement with the one of original codes. The validation analysis for single droplet falling and combustion corresponds well with the experimental data reasonably.
Otsuka, Satoshi; Tanno, Takashi; Oka, Hiroshi; Yano, Yasuhide; Tachi, Yoshiaki; Kaito, Takeji; Hashidate, Ryuta; Kato, Shoichi; Furukawa, Tomohiro; Ito, Chikara; et al.
2018 GIF Symposium Proceedings (Internet), p.305 - 314, 2020/05
Oxide Dispersion Strengthened (ODS) steel has been developed worldwide as a high-strength and radiation-tolerant steel used for advanced nuclear system. Japan Atomic Energy Agency (JAEA) has been developing ODS steel as the primary candidate material of Sodium cooled Fast Reactor (SFR) high burn-up fuel cladding tube. Application of high burn-up fuel to SFR core can contribute to improvement of economical performance of SFR in conjunction with volume and hazardousness reduction of radioactive waste. This paper described the current status and future prospects of ODS tempered martensitic steel development in JAEA for SFR fuel application.
Watanabe, Masashi; Matsumoto, Taku; Hirooka, Shun; Morimoto, Kyoichi; Kato, Masato
2018 GIF Symposium Proceedings (Internet), p.315 - 320, 2020/05
Recently, a research group studying at Plutonium Fuel Development Facility (PFDF) in Japan Atomic Energy Agency has systematically measured vast amounts of physical properties in the non-stoichiometric (U, Pu)O. Lattice parameter, elastic modulus, thermal expansion, oxygen potential, oxygen chemical diffusion coefficient and thermal conductivity were successfully measured as function of Pu content, O/M ratio and temperature, and the effects of Pu content and O/M ratio on their physical properties were evaluated. In this work, those experimental data are reviewed, and latest experimental data set on the non-stoichiometric (U, Pu)O are presented. The data set would be available in development of a fuel performance code.
Cappia, F.*; Tanaka, Kosuke; Kato, Masato; McClellan, K.*; Harp, J.*
Journal of Nuclear Materials, 533, p.152076_1 - 152076_14, 2020/05
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 (800C), 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.
Kudo, Hideyuki*; Otani, Yuichi*; Hara, Masahide*; Kato, Atsushi; Otaka, Masahiko; Ide, Akihiro*
Journal of Nuclear Science and Technology, 57(4), p.408 - 420, 2020/04
In a fuel handling system of sodium-cooled fast reactors (SFRs), it is necessary to remove the sodium remaining on spent fuel assemblies (FAs) before storing them in a spent fuel water pool (SFP) in order to minimize plant operating loads. A next-generation SFR in Japan has adopted an advanced dry cleaning process which consists of the following steps, argon gas blowing to remove the metallic residual sodium on the FA, moist argon gas blowing to deactivate the residual sodium, and direct storage in the SFP. This three-step process increases economic competitiveness and reduces waste products thanks to a waterless process. In this R&D work, performance of the dry cleaning process has been investigated.
Uwaba, Tomoyuki; Yokoyama, Keisuke; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*; Pelletier, M.*
Nuclear Engineering and Design, 359, p.110448_1 - 110448_7, 2020/04
Coupled computer code analyses of irradiation performance of axially heterogeneous mixed oxide (MOX) fuel elements with high burnup in a fast reactor were conducted. Post-irradiation experiments revealed local concentration of Cs near the interfaces between MOX fuel and blanket columns including the internal blanket of the fuel elements as well as an increase in their cladding diameters. The analyses indicated that the local Cs concentration occurred as a result of Cs axial migration from the MOX fuels toward the blanket pellets near the interfaces. Swelling of the blanket pellets induced by the formation of low-density Cs-U-O compound was not sufficient to cause pellet-to-cladding mechanical interaction (PCMI). The PCMI analyzed in the MOX fuel column regions was insignificant, and the cladding diameter increases were caused mainly by void swelling in cladding and irradiation creep due to fission gas pressure.
Mitsumoto, Rika; Hazama, Taira; Takahashi, Keita; Kondo, Satoru
JAEA-Technology 2019-020, 167 Pages, 2020/03
The prototype fast breeder reactor Monju has produced valuable technological achievements through design, construction, operation and maintenance over half a century since 1968. This report compiles the reactor technologies developed for Monju, including the areas: history and major achievements, design and construction, commissioning, safety, reactor physics, fuel, systems and components, sodium technology, materials and structures, operation and maintenance, and accidents and failures.