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Journal Articles

Promoting nuclear fuel cycle and ensuring nuclear non-proliferation/nuclear security

Tamai, Hiroshi; Mochiji, Toshiro; Senzaki, Masao*; Iwamoto, Tomonori*; Ishiguro, Yuzuru*; Kitade, Yuta; Sato, Heigo*; Suehiro, Rie*; Taniguchi, Tomihiro*; Fukasawa, Tetsuo*; et al.

Dai-41-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2020/11

In light of recent delay of plutonium use in Japan and the increasing criticism of nuclear non-proliferation and nuclear security in the nuclear fuel cycle, the validity of these criticisms will be examined for the sustainable development of the nuclear fuel cycle policy. Issues on the view point of nuclear non-proliferation and nuclear security are examined.

Journal Articles

Next generation reactor development and current status

Kamide, Hideki

Karyoku Genshiryoku Hatsuden, 71(11), p.638 - 648, 2020/11

Development status of next generation reactors is outlined mainly for sodium cooled fast reactor developed in Japan Atomic Energy Agency (JAEA). Development strategy in Japan, status of the advanced reactor developments including SMR over the world, and also results of research and development in JAEA are explained.

Journal Articles

Validation study of finite element thermal-hydraulics analysis code SPIRAL to a large-scale wire-wrapped fuel assembly at low flow rate condition

Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Gerschenfeld, A.*

Proceedings of Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo 2020 (SNA + MC 2020), p.73 - 80, 2020/10

A finite element thermal-hydraulics simulation code SPIRAL has been developed in Japan Atomic Energy Agency (JAEA) to analyze the detailed thermal-hydraulics phenomena in a fuel assembly (FA) of Sodium-cooled Fast Reactors (SFRs). The numerical simulation of a large-scale sodium test for 91-pin bundle (GR91) at low flow rate condition was performed for the validation of SPIRAL with the hybrid k-e turbulence model to take into account the low Re number effect near the wall in the flow and temperature fields. Through the numerical simulation, specific velocity distribution affected by the buoyancy force was shown on the top of the heated region and the temperature distribution predicted by SPIRAL agreed with that measured in the experiment and the applicability of the SPIRAL to thermal-hydraulic evaluation of large-scale fuel assembly at low flow rate condition was indicated.

Journal Articles

Kinetic study on eutectic melting process between boron carbide and stainless steel in sodium-cooled fast reactor

Kikuchi, Shin; Sakamoto, Kan*; Takai, Toshihide; Yamano, Hidemasa

Nippon Kikai Gakkai 2020-Nendo Nenji Taikai Koen Rombunshu (Internet), 4 Pages, 2020/09

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (B$$_{4}$$C) as control rod element and stainless steel (SS) as control rod cladding or related structure may occur. Thus, behavior of B$$_{4}$$C-SS eutectic melting is one of the phenomena to evaluate the core disruptive accidents in SFR. In order to clarify the kinetic feature of B$$_{4}$$C-SS eutectic melting process in the interface, the thinning test for SS crucibles using the pellets of B$$_{4}$$C or SS with low B$$_{4}$$C concentration were performed to obtain the rate constant with dependence of B$$_{4}$$C concentration against SS. It was found that the rate constants of eutectic melting between SS and SS low B$$_{4}$$C concentration were smaller than that of B$$_{4}$$C-SS in the high temperature range. Besides, the rate constant of eutectic melting between SS and B$$_{4}$$C containing SS became small when decreasing the B$$_{4}$$C concentration against SS.

Journal Articles

Oxide dispersion strengthened steels

Ukai, Shigeharu*; Ono, Naoko*; Otsuka, Satoshi

Comprehensive Nuclear Materials, 2nd Edition, Vol.3, p.255 - 292, 2020/08

Fe-Cr-based oxide dispersion strengthened (ODS) steels have a strong potential for high burnup (long-life) and high-temperature applications typical for SFR fuel cladding. Current progress in the development of Fe-Cr-based ODS steel claddings is reviewed, including their relevant mechanical properties, e.g. tensile and creep rupture strengths in the hoop directions. In addition, this paper reviewed the current research status on corrosion resistant Fe-Cr-Al-based ODS steel claddings, which are greatly paid attention recently as the accident tolerant fuel claddings for the light water reactor (LWR) and also as the claddings of the lead fast reactors (LFR) utilizing Pb-Bi eutectic (LBE) coolant.

Journal Articles

Measurement of thermal decomposition temperature and rate of sodium hydride

Kawaguchi, Munemichi

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

In decommissioning sodium-cooled fast reactors, the operators can be exposed to radiation during dismantling of the cold trap equipment (C/T). The C/T has higher dose equipment because the C/T trapped the tritium of the fission product during the operation to purify the sodium coolant. In this research, the thermal decomposition temperature and rate of sodium hydride were measured as the fundamental research for improvement of the thermorlysis method prior to the dismantling. We measured the thermal decomposition temperature and rate using sodium hydride powder (95.3%, Sigma-Aldrich) in Al$$_2$$O$$_3$$ crucible with TG-DTA (STA2500, NETASCH Japan). The heating rates were set to $$beta$$ = 2.0, 5.0, 10.0, 20.0 K/min, and the weight decrease was measured. The thermal decomposition temperature and rate were obtained from the temperature of the onset of the weigh decrease and the Kissinger plot, respectively. Furthermore, we set to the thermal decomposition temperature of around 600 K, and the weight decreasing rates were measured. The change of the chemical composition of the sodium hydride with heating (from NaH to Na) was measured with X-Ray Diffraction (XRD) analysis. As a result, the thermal decomposition occurred at around 600 K, and the almost all hydrogen in the sodium hydride released within 1 h. The thermal decomposition rate strongly depended on the heating temperature.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 1; Project overview and progress until 2018

Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08

One of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors is eutectic reactions between boron carbide (B$$_{4}$$C) and stainless steel (SS) as well as its relocation. Such behaviors have never been simulated in CDA numerical analyses in the past, therefore it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study focuses on B$$_{4}$$C-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in a range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies conducted until 2018. Specific results in this paper are boron concentration distributions of solidified B$$_{4}$$C-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 3; Kinetic study of boron carbide-stainless steel eutectic melting by differential thermal analysis

Kikuchi, Shin; Yamano, Hidemasa; Nakamura, Kinya*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 9 Pages, 2020/08

In a postulated severe accidental condition of sodium-cooled fast reactor (SFR), eutectic melting between boron carbide (B$$_{4}$$C) and stainless steel (SS) may take place. Thus, kinetic behavior of B$$_{4}$$C-SS eutectic melting is one of the important phenomena to be considered when evaluating the core disruptive accidents in SFR. In this study, for the first step to obtain the fundamental information on kinetic feature of B$$_{4}$$C-SS eutectic melting, the thermal analysis using the pellet type samples of B$$_{4}$$C and Type 316L SS as different experimental technique was performed. The differential thermal analysis endothermic peaks for the B$$_{4}$$C-SS eutectic melting appeared from 1483K to 1534K and systematically shifted to higher temperatures when increasing heating rate. Based on this kinetic feature, apparent activation energy and pre-exponential factor for the B$$_{4}$$C-SS eutectic melting were determined by Kissinger method. It was found that the kinetic parameters obtained by thermal analysis were comparable to the literature values.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 4; Validation of a multi-phase model for eutectic reaction between stainless steel and boron carbide

Liu, X.*; Morita, Koji*; Yamano, Hidemasa

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 9 Pages, 2020/08

In our previous study, a two-dimensional fast reactor safety analysis code, SIMMER-III, was extended to include a physical model to simulate the eutectic reaction between stainless steel (SS) and B$$_{4}$$C. Based on experimental knowledge on eutectic reaction, the growth of eutectic material was modeled according to a parabolic rate law. Heat and mass transfer behaviors among reactor materials including a eutectic composition in solid and liquid phases were also modeled considering both equilibrium and non-equilibrium processes in phase change. Physical properties of the eutectic composition were also formulated based on experimental measurements for 5 mass% B$$_{4}$$C-SS composition. In this study, we extended the eutectic reaction model to SIMMER-IV, a three-dimensional counterpart of SIMMER-III. We performed validation analysis using SIMMER-III and SIMMER-IV with the developed model based on an experiment, where a B$$_{4}$$C pellet was immersed into a molten SS pool. Boron concentration in the pool was measured at several time points and the boron concentration after solidification of the molten pool was compared with the experiment post analysis result. Simulation results of boron distribution are comparable to the experimental results.

Journal Articles

Numerical validation of AQUA-SF in SNL T3 sodium spray fire experiment

Sonehara, Masateru; Aoyagi, Mitsuhiro; Uchibori, Akihiro; Takata, Takashi; Ohshima, Hiroyuki; Clark, A. J.*; Louie, D. L. Y.*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 4 Pages, 2020/08

In order to investigate the multi-dimensional effects of sodium combustion, a benchmark analysis of the SNL Surtsey spray combustion experiment (SNL T3 experiments) using AQUA-SF and SPHINCS is conducted in JAEA. As a best estimate analysis, the spray burning duration is adjusted in the computation in order to take into account the temporary suppression of the spray combustion observed in the experiment. Furthermore, droplet size of SPHINCS and AQUA-SF are optimized to represent the T3 experimental results. The best estimate of AQUA-SF results in the droplet diameter of 2.5 mm, which agrees quite well with the spatial temperature measurements, and the sodium droplet diameter measurement with a high speed camera.

Journal Articles

Development of neutronics and thermal-hydraulics coupled analysis method on platform for design optimization in fast reactor

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.

Journal Articles

Oxygen potential and self-irradiation effects on fuel temperature in Am-MOX

Ikusawa, Yoshihisa; Hirooka, Shun; Uno, Masayoshi*

2018 GIF Symposium Proceedings (Internet), p.321 - 327, 2020/05

Research and development of Minor actinides (MAs) bearing MOX fuel for fast reactor has been proceeding from the viewpoint of reducing radioactive waste. In order to develop, MA bearing MOX, it is indispensable to clarify the influence of MA addition on irradiation behavior. The addition of Americium (Am) to MOX affects vapor pressure and thermal conductivity, which are important properties from the perspective of evaluating fuel temperature. This is because vapor pressure affects fuel restructuring, and thermal conductivity affects fuel temperature distribution. Focusing on these physical properties, this study evaluates the influence of Am on fuel temperature using irradiation behavior analysis code to contribute to the development of MA-bearing MOX fuel. An increase in Am content decreases the thermal conductivity and increases the oxygen potential of oxide fuel. Because vapor pressure increases with increasing Am content, pore migration is accelerated, and the central void diameter increases with increasing Am content. As a result, after formation of the central void, the influence of Am content on the fuel center temperature is mild. Alpha particles generated by radioactive decay of transuranium elements cause lattice defects in the oxide fuel pellets. It is well known that this phenomenon, which is called self-irradiation, affects thermal conductivity. Since americium is the typical alpha radioactive nucleus, to evaluate fuel temperature of Am-MOX is necessary to take account of the influence of self-irradiation damage on thermal conductivity. Self-irradiation decreases thermal conductivity, and as the Am content increases, the rate of decrease in thermal conductivity is accelerated. Because it recovers with temperature rise, the decrease in thermal conductivity due to self-irradiation damage has very little effect on fuel center temperature. These results suggest that Am-MOX fuel could be irradiated under the same conditions as conventional MOX fuel.

Journal Articles

A Study on sodium-concrete reaction in presence of internal heating

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$$^{circ}$$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 800$$^{circ}$$C 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.

Journal Articles

Computer code analysis of irradiation performance of axially heterogeneous mixed oxide fuel elements attaining high burnup in a fast reactor

Uwaba, Tomoyuki; Yokoyama, Keisuke; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*; Pelletier, M.*

Nuclear Engineering and Design, 359, p.110448_1 - 110448_7, 2020/04

 Times Cited Count:0 Percentile:100(Nuclear Science & Technology)

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.

Journal Articles

Post-test material analysis of eutectic melting reaction of boron carbide and stainless steel

Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro

Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00360_1 - 19-00360_13, 2020/03

It is necessary to simulate a eutectic melting reaction and relocation behavior of boron carbide (B$$_{4}$$C) as a control rod material and stainless steel (SS) during a core disruptive accident in an advanced sodium-cooled fast reactor designed in Japan because the B$$_{4}$$C-SS eutectic relocation behavior has a large uncertainty in the reactivity history based on a simple calculation. A physical model simulating the eutectic melting reaction and relocation was developed and implemented into a severe accident simulation code. The developed model must be validated by using test data. To validate the physical model, therefore, the visualization tests of SS-B$$_{4}$$C eutectic melting reaction was carried out by contacting SS melts of several kg with a B$$_{4}$$C pellet heated up to about 1500 $$^{circ}$$C. The tests have shown the eutectic reaction visualization as well as freezing and relocation of the B$$_{4}$$C-SS eutectic in upper part of the solidified test piece due to the density separation. Post-test material analyses by using X-ray diffraction and transmission electron microscope techniques have indicated that FeB appeared at the B$$_{4}$$C-SS contact interface and (Fe,Cr)$$_{2}$$B at the top surface of the test piece. Glow discharge optical emission spectrometry has been applied to quantitative analysis of boron concentration distributions. The boron concentration was high at the upper surface and near the original position of the B$$_{4}$$C pellet.

Journal Articles

Development of numerical analysis code LEAP-III for tube failure propagation

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Ohshima, Hiroyuki

Nippon Kikai Gakkai Rombunshu (Internet), 86(883), p.19-00353_1 - 19-00353_6, 2020/03

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.

Journal Articles

National policy process on energy security; Case study on restart process of the fast breeder reactor Monju

Koito, Yuko

Rinsho Seiji Kenkyu, (10), p.1 - 23, 2019/12

no abstracts in English

Journal Articles

Outline of the R&D plan for the fast reactor cycle system development in JAEA

Hayafune, Hiroki; Maeda, Seiichiro; Ohshima, Hiroyuki

Nippon Genshiryoku Gakkai-Shi, 61(11), p.798 - 803, 2019/11

In the "Strategic Roadmap" of Fast Reactor Development decided at the Inter-Ministerial Council for Nuclear Power in December 2018, the development works for the around next 10 years were identified, and the role of JAEA was presented. In response, JAEA has prepared a framework for R&D plans for about 5 years on the fast reactor technology and the fuel cycle technology (reprocessing, fuel manufacturing, fuel and material development). In the future, JAEA will promote independent R&D works based on these plans, and provide the obtained R&D results together with various testing functions of JAEA to the activities of the private sector, etc. Through these actions, JAEA will actively contribute to the future fast reactor development. This article outlines JAEA's policy and the R&D items (development of ARKADIA; Advanced Reactor Knowledge- and AI-Aided Design Integration Approach through the whole Plant Life Cycle, development of standards and standards system, development of safety improvement technology, research in the fuel cycle technology), the policy of international cooperation, the human resource development, and the future perspective were explained.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 1; Project overview

Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.418 - 427, 2019/09

Eutectic reactions between boron carbide (B$$_{4}$$C) and stainless steel (SS) as well as its relocation are one of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors. Since such behaviors have never been simulated in CDA numerical analyses, it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study is focusing on B$$_{4}$$C-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in the range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies by 2017. Specific results in this paper is boron concentration distributions of solidified B$$_{4}$$C-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.

Journal Articles

Verification of detailed core-bowing analysis code ARKAS_cellule with IAEA benchmark problems

Ota, Hirokazu*; Ohgama, Kazuya; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.30 - 39, 2019/09

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