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Journal Articles

Validation practices of multi-physics core performance analysis in an advanced reactor design study

Doda, Norihiro; Kato, Shinya; Hamase, Erina; Kuwagaki, Kazuki; Kikuchi, Norihiro; Ohgama, Kazuya; Yoshimura, Kazuo; Yoshikawa, Ryuji; Yokoyama, Kenji; Uwaba, Tomoyuki; et al.

Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.946 - 959, 2023/08

An innovative design system named ARKADIA is being developed to realize the design of advanced nuclear reactors as safe, economical, and sustainable carbon-free energy sources. This paper focuses on ARKADIA-Design for design studies as a part of ARKADIA and introduces representative verification methods for numerical analysis methods of the core design. ARKADIA-Design performs core performance analysis of sodium-cooled fast reactors using a multiphysics approach that combines neutronics, thermal-hydraulics, core mechanics, and fuel pin behavior analysis codes. To confirm the validity of these analysis codes, validation matrices are identified with reference to experimental data and reliable numerical analysis results. The analysis models in these codes and the representative practices for the validation matrices are described.

Journal Articles

Verification of fuel assembly bowing analysis model for core deformation reactivity evaluation

Doda, Norihiro; Uwaba, Tomoyuki; Ohgama, Kazuya; Yoshimura, Kazuo; Nemoto, Toshiyuki*; Tanaka, Masaaki; Yamano, Hidemasa

Nihon Kikai Gakkai Kanto Shibu Dai-29-Ki Sokai, Koenkai Koen Rombunshu (Internet), 5 Pages, 2023/03

An evaluation method for reactivity feedback due to core deformation during reactor power increase in sodium-cooled fast reactors is being developed for realistic core design evaluation. In this evaluation method, fuel assembly bowing was modeled with a beam element of the finite element method, and the assembly's pad contact between adjacent assemblies was modeled with a dedicated element which could consider the wrapper tube cross-sectional distortion and the pad stiffness depending on pad contact conditions. This fuel assembly bowing analysis model was verified for thermal bowing of a single assembly and assembly pad contact between adjacent assemblies in a core as past benchmark problems. The calculation results by this model showed good agreement with those of reference solutions of theoretical solutions or results by participating institutions in the benchmark. This study confirmed that the analysis model was able to calculate thermal assembly bowing appropriately.

Journal Articles

Development of ARKADIA-Design for design optimization support; Application of coupling method using multi-level simulation technique for plant thermal-hydraulics analysis

Doda, Norihiro; Yoshimura, Kazuo; Hamase, Erina; Yokoyama, Kenji; Uwaba, Tomoyuki; Tanaka, Masaaki

Proceedings of Technical Meeting on State-of-the-art Thermal Hydraulics of Fast Reactors (Internet), 3 Pages, 2022/09

ARKADIA-Design is being developed to support the optimization of sodium-cooled fast reactors in the conceptual design stage. Design optimization requires various types of numerical analysis: 1-D plant dynamics analysis for efficient evaluation of various design options and multi-dimensional analysis for a detailed evaluation of local phenomena, including multi-physics. For those analyses, ARKADIA-Design performs whole plant analyses based on the multi-level simulation (MLS) technique in which the analysis codes are coupled to simulate the phenomena in an intended degree of resolution. This paper describes an outline of the coupling analysis methods in the MLS of the ARKADIA-Design and the numerical simulations of the experimental fast breeder reactor EBR-II tests by the coupled analysis.

Journal Articles

Sodium-cooled Fast Reactors

Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.

JAEA Reports

Experimental study on velocity distribution in the subchannels of a fuel pin bundle with wrapping wire; Evaluation of the characteristics of flow field in 3-pin bundle

Hiyama, Tomoyuki; Aizawa, Kosuke; Nishimura, Masahiro; Kurihara, Akikazu

JAEA-Research 2021-009, 29 Pages, 2021/11

JAEA-Research-2021-009.pdf:2.25MB

In sodium-cooled fast reactors, high burnup of fuel is required for practical use. It is important to predict and evaluate the flow behavior in a fuel assembly because there is a concern that the heat removal capacity of the fuel assembly with high burnup will be locally reduced due to swirling and thermal deformation of the fuel rods. In this study, flow field measurement tests were conducted using a 3-pin bundle system test specimen for the purpose of elucidating the phenomenon and constructing a verification database for thermal hydraulics analysis code. The viewpoints of the experiment for elucidating the phenomenon are as follows; (1) Overall flow behavior in the subchannel including near the wrapping wire, (2) Relationship between Reynolds number including laminar flow region and flow field, and (3) Evaluation of the effect of the presence or absence of wrapping wire on the flow field. As a result, detailed flow field data in the subchannel was obtained by PIV measurement. It was found that when the wrapping wire crossed the subchannel, the flow occurred toward adjacent subchannel and the flow occurred that follows the winding direction of the wrapping wire. It was confirmed that the tendency of the flow velocity distribution of the Reynolds number in the laminar flow region is significantly different from that of the transition region and the turbulent region under the condition. The test was conducted using a same 3-pin bundle system without the wrapping wire, and it was confirmed that mixing by the wrapping wire occurred even in the laminar flow region.

Journal Articles

Velocity distribution in the subchannels of a pin bundle with a wrapping wire; Evaluation of the Reynolds number dependence in a three-pin bundle

Aizawa, Kosuke; Hiyama, Tomoyuki; Nishimura, Masahiro; Kurihara, Akikazu; Ishida, Katsuji*

Mechanical Engineering Journal (Internet), 8(4), p.20-00547_1 - 20-00547_11, 2021/08

A sodium-cooled fast reactor has been designed to attain a high burn-up core in commercialized fast reactor cycle systems. The sodium-cooled fast reactor adopts a wire spacer between fuel pins. The wire spacer performs functions of securing the coolant channel and the mixing between subchannels. In high burn-up fuel subassemblies, the fuel pin deformation due to swelling and thermal bowing may decrease the local flow velocity in the subassembly and influence the heat removal capability. Therefore, understanding the flow field in a wire-wrapped pin bundle is important. This study performed particle image velocimetry (PIV) measurements using a wire-wrapped three-pin bundle water model to grasp the flow field in the subchannel under conditions, including the laminar to turbulent regions. In the region away from the wrapping wire, the maximum flow velocity was increased by decreasing the Re number. Accordingly, the PIV measurements using the three-pin bundle geometry without the wrapping wire were also conducted to understand the effect of the wrapping wires on the flow field in the subchannel. The results confirmed that the mixing due to the wrapping wire occurred, even in the laminar condition. These experimental results are useful not only for understanding the pin bundle thermal hydraulics, but also for the code validation.

Journal Articles

Effect of water vapor on re-saturation process in EBS performance of re-saturation process by Thermo-Hydro-Mechanical coupled analysis

Sato, Shin*; Ono, Hirokazu; Tanai, Kenji; Yamamoto, Shuichi*; Fukaya, Masaaki*; Shimura, Tomoyuki*; Niunoya, Sumio*

Jiban Kogaku Janaru (Internet), 15(3), p.529 - 541, 2020/09

no abstracts in English

Journal Articles

Investigation on velocity distribution in the subchannels of pin bundle with wrapping wire; Evaluation of Reynolds number dependence in 3-pin bundle

Aizawa, Kosuke; Hiyama, Tomoyuki; Nishimura, Masahiro; Kurihara, Akikazu; Ishida, Katsuji*

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 8 Pages, 2020/08

A sodium-cooled fast reactor is designed to attain a high burn-up core in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, the deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the flow velocity distribution in a wire wrapped pin bundle. In this study, the detailed flow velocity distribution in the subchannel has been obtained by PIV (Particle Image Velocimetry) measurement using a wire-wrapped 3-pin bundle water model. Flow velocity conditions in the pin bundle were set from 0.036 m/s ($$Re$$ = 270) to 1.6m/s ($$Re$$ = 13,500). From the PIV results, the maximum flow velocity was increased by decreasing the $$Re$$ number in the region away from the wrapping wire. Moreover, the PIV measurements by using the 3-pin bundle geometry without the wrapping wire were conducted. From the results, the effect of the wrapping wire on the flow field in the subchannel was understood. There experimental results useful not only for understanding of pin bundle thermal hydraulics but also code validation.

Journal Articles

Effect of copper addition on precipitation behavior near grain boundary in Al-Zn-Mg alloy

Matsuda, Kenji*; Yasumoto, Toru*; Bendo, A.*; Tsuchiya, Taiki*; Lee, S.*; Nishimura, Katsuhiko*; Nunomura, Norio*; Marioara, C. D.*; Lervik, A.*; Holmestad, R.*; et al.

Materials Transactions, 60(8), p.1688 - 1696, 2019/08

 Times Cited Count:13 Percentile:63.77(Materials Science, Multidisciplinary)

no abstracts in English

Journal Articles

Enhancement of element production by incomplete fusion reaction with weakly bound deuteron

Wang, H.*; Otsu, Hideaki*; Chiga, Nobuyuki*; Kawase, Shoichiro*; Takeuchi, Satoshi*; Sumikama, Toshiyuki*; Koyama, Shumpei*; Sakurai, Hiroyoshi*; Watanabe, Yukinobu*; Nakayama, Shinsuke; et al.

Communications Physics (Internet), 2(1), p.78_1 - 78_6, 2019/07

 Times Cited Count:7 Percentile:56.64(Physics, Multidisciplinary)

Searching for effective pathways for the production of proton- and neutron-rich isotopes through an optimal combination of reaction mechanism and energy is one of the main driving forces behind experimental and theoretical nuclear reaction studies as well as for practical applications in nuclear transmutation of radioactive waste. We report on a study on incomplete fusion induced by deuteron, which contains one proton and one neutron with a weak binding energy and is easily broken up. This reaction study was achieved by measuring directly the cross sections for both proton and deuteron for $$^{107}$$Pd at 50 MeV/u via inverse kinematics technique. The results provide direct experimental evidence for the onset of a cross-section enhancement at high energy, indicating the potential of incomplete fusion induced by loosely-bound nuclei for creating proton-rich isotopes and nuclear transmutation of radioactive waste.

Journal Articles

Characterisation of structural similarities of precipitates in Mg-Zn and Al-Zn-Mg alloys systems

Bendo, A.*; Maeda, Tomoyoshi*; Matsuda, Kenji*; Lervik, A.*; Holmestad, R.*; Marioara, C. D.*; Nishimura, Katsuhiko*; Nunomura, Norio*; Toda, Hiroyuki*; Yamaguchi, Masatake; et al.

Philosophical Magazine, 99(21), p.2619 - 2635, 2019/07

 Times Cited Count:26 Percentile:83.26(Materials Science, Multidisciplinary)

Journal Articles

Numerical analysis of flow field around simulated wire-wrapped fuel pins of fast reactor

Kikuchi, Norihiro; Ohshima, Hiroyuki; Imai, Yasutomo*; Hiyama, Tomoyuki; Nishimura, Masahiro; Tanaka, Masaaki

Nihon Kikai Gakkai Kanto Shibu Ibaraki Koenkai 2015 Koen Rombunshu, p.179 - 180, 2015/08

In an economically improved sodium-cooled fast reactor, a narrower gap is considered among the fuel pins so as to achieve a high burn-up. Therefore, it is needed to evaluate thermal-hydraulic characteristics in case of a change of the gap geometry due to deformation of fuel pin caused by such as a swelling and a thermal bowing. For this purpose, a FEM analysis code, SPIRAL has been being developed in JAEA and the code validations using water or sodium experimental results have also being performed. In this study, a numerical analysis of a flow field around wire-wrapped fuel pins based on a 3-pin bundle water experiment was carried out as a validation study of SPIRAL. As a result, it was demonstrated that the hybrid-type turbulent model incorporated in SPIRAL has a high applicability to investigate the flow structure of the narrow gap in the fuel assembly.

Journal Articles

Study on flow in the subchannels of pin bundle with wrapping wire

Nishimura, Masahiro; Hiyama, Tomoyuki; Kamide, Hideki; Ohshima, Hiroyuki; Nagasawa, Kazuyoshi*; Imai, Yasutomo*

Proceedings of 9th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-9) (CD-ROM), 7 Pages, 2014/11

Journal Articles

Planning of in-situ experiment for understanding of gas migration behaviour in sedimentary rock, 1; Setting of gas injection procedure

Tanai, Kenji; Fujita, Tomoo; Noda, Masaru*; Yamamoto, Shuichi*; Shimura, Tomoyuki*; Sato, Shin*

Dai-13-Kai Iwa No Rikigaku Kokunai Shimpojiumu Koen Rombunshu (CD-ROM), p.167 - 172, 2013/01

Japan Atomic Energy Agency has been planning in-situ gas migration test in Horonobe URL, Hokkaido. This paper discusses the optimum gas injection procedure for the test to understand gas migration behaviour in surrounded rock. The stepwise constant gas injection was selected, taking into account domestic and overseas gas related research results. Hydro-mechanical-gas coupling analysis which is able to consider the dissolved methane in Horonobe groundwater was applied to evaluate the gas behaviour. The results have indicated no significant mechanical damages to the rock and have supported the sppropriateness of selected gas injection procedure for the test.

Journal Articles

Planning of in-situ experiment for understanding of gas migration behaviour in sedimentary rock, 2; Affects of dissolved methane to the gas migration in bedrock

Tanai, Kenji; Fujita, Tomoo; Sato, Shin*; Noda, Masaru*; Yamamoto, Shuichi*; Shimura, Tomoyuki*

Dai-13-Kai Iwa No Rikigaku Kokunai Shimpojiumu Koen Rombunshu (CD-ROM), p.173 - 178, 2013/01

Japan Atomic Energy Agency has been planning ${it in-situ}$ gas migration test in Horonobe URL, Hokkaido. It is expected that dissolved methane in Horonobe groundwater might have an effect on gas migration behaviour in bedrock. A series of two-phase multi-component analyses by use of GETFLOWS were conducted to understand the influence of dissolved methane. The increase of total gas pressure has been shown due to the existence of dissolved methane. The results also indicated that the injected nitrogen gas volume was influenced by dissolved methane.

Journal Articles

FIREX project and effects of self-generated electric and magnetic fields on electron-driven fast ignition

Mima, Kunioki*; Sunahara, Atsushi*; Shiraga, Hiroyuki*; Nishimura, Hiroaki*; Azechi, Hiroshi*; Nakamura, Tatsufumi; Jozaki, Tomoyuki*; Nagatomo, Hideo*; Garcia, C.*; Veralde, P.*

Plasma Physics and Controlled Fusion, 52(12), p.124047_1 - 124047_6, 2010/12

 Times Cited Count:9 Percentile:34.11(Physics, Fluids & Plasmas)

Fast ignition is a new scheme in laser fusion, in which higher energy gain with a smaller laser pulse energy is expected. At Osaka University, a laser with four beams and a total output of 10 kJ ps-1, laser for fast ignition experiment (LFEX), has been constructed and we have carried out an integrated experiment with one beam of the LFEX. Through experiments it was found that the coupling efficiency is degraded when the laser pre-pulse is not sufficiently small. Furthermore, the distance between the hot electron source and the core plasma is large. In this paper it is proposed that a thin foil covers the laser entrance of the cone to mitigate the pre-plasma and a double cone reduces the loss of high energy electrons from the side wall of the cone. The simulations indicate that a higher coupling efficiency is expected for the double cone target with a thin foil at the laser entrance.

Journal Articles

Effect of oxygen-to-metal ratio on melting temperature of uranium and plutonium mixed oxide fuel for fast reactor

Kato, Masato; Morimoto, Kyoichi; Nakamichi, Shinya; Sugata, Hiromasa*; Konashi, Kenji*; Kashimura, Motoaki; Abe, Tomoyuki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 7(4), p.420 - 428, 2008/12

The melting temperatures of MOX for fast reactor fuel were investigated as functions of Pu content, Am content and oxygen-to-metal (O/M) ratio using thermal arrest technique. Rhenium inner was used for the measurement to prevent the reaction between the sample and capsule materials. The solidus temperatures decreased with increasing Pu and Am content and increased with decreasing O/M ratio. It is considered that the maximum temperature in U-Pu-O system varies in hypostoichiometric composition region. The melting temperatures were evaluated by ideal solid solution model in UO$$_{2}$$-PuO$$_{2}$$-AmO$$_{2}$$-PuO$$_{1.7}$$ system, and the model was derived for calculating solidus and liquidus temperature. The derived model reproduced the experimental data with $$pm$$25 K.

Journal Articles

Thermal conductivities of (U,Pu,Am)O$$_{2}$$ solid solutions

Morimoto, Kyoichi; Kato, Masato; Ogasawara, Masahiro*; Kashimura, Motoaki; Abe, Tomoyuki

Journal of Alloys and Compounds, 452(1), p.54 - 60, 2008/03

 Times Cited Count:28 Percentile:77.26(Chemistry, Physical)

Plutonium and uranium mixed oxide (MOX) fuel with high Pu content have been developed as a fuel of fast reactor (FR). As the storage time of Pu raw material between reprocessing and fabrication increases, americium content of the fabricated MOX fuel increases up to a few percent. In this work, the thermal conductivity of MOX fuel containing Am was investigated as a part of clarifying the effect of Am content on thermal physical properties. The pellets of (Am$$_{0.007}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$, (Am$$_{0.02}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$ and (Am$$_{0.03}$$ Pu$$_{0.3}$$ U)O$$_{2.00}$$ were prepared. The oxygen to metal ratio (O/M ratio) of sintered pellet was adjusted to 2.00. The thermal diffusivity measurement was carried out in the range of temperature from 900 K to 1700 K by the laser flash method, and thermal conductivity of these pellets was evaluated. The heat capacity for evaluating thermal conductivity was derived from heat capacity of UO$$_{2}$$, PuO$$_{2}$$ and AmO$$_{2}$$ by using the Kopp-Neumann rule.

Journal Articles

Solidus and liquidus of plutonium and uranium mixed oxide

Kato, Masato; Morimoto, Kyoichi; Sugata, Hiromasa*; Konashi, Kenji*; Kashimura, Motoaki; Abe, Tomoyuki

Journal of Alloys and Compounds, 452(1), p.48 - 53, 2008/03

 Times Cited Count:29 Percentile:77.26(Chemistry, Physical)

Plutonium and uranium mixed oxide has been developed as a fuel of a fast reactor. The maximum temperature of the fuel pellet is limited within a design criterion to prevent fuel melting. So, the melting points of the mixed oxide have been investigated since the development of fast reactor started. However the measured data are limited. In this work, the melting points of (U1-yPuy)O$$_{2-x}$$ (y: 0, 0.12, 0.2, 0.3, 0.4) were measured by the thermal arrest method. The evaluated melting point of this study underestimates in case of MOX with high Pu contents of 30% and 40%. The solidus of UO$$_{2}$$, (Pu$$_{0.12}$$U$$_{0.88}$$)O$$_{2.00}$$ and (Pu$$_{0.2}$$U$$_{0.8}$$)O$$_{2.00}$$ were determined to be 3128K, 3077K and 3052K, respectively. The solidus temperature of hypostoichiometric MOX slightly increased with decreasing O/M.

Journal Articles

Solidus and liquidus temperatures in the UO$$_{2}$$-PuO$$_{2}$$ system

Kato, Masato; Morimoto, Kyoichi; Sugata, Hiromasa*; Konashi, Kenji*; Kashimura, Motoaki; Abe, Tomoyuki

Journal of Nuclear Materials, 373(1-3), p.237 - 245, 2008/02

 Times Cited Count:59 Percentile:96.07(Materials Science, Multidisciplinary)

The melting of plutonium and uranium mixed oxide (MOX) containing Pu of more than 30% was investigated using a tungsten capsule and a rhenium inner capsule. In the conventional measurement of MOX in the tungsten capsule, a liquid phase of tungsten and plutonium oxide appeared in the MOX during melting. This liquid phase was found to have an effect on the measurement of melting point. Therefore the rhenium inner capsule was used to avoid the effect. The solidus and liquidus temperatures in the UO$$_{2}$$-PuO$$_{2}$$ system were decided from the MOX data measured using the rhenium capsule, and the effect of the Am content on the solidus temperature was evaluated. The variation of the solidus and liquidus temperatures in the UO$$_{2}$$-PuO$$_{2}$$-AmO$$_{2}$$ ternary system was represented to an accuracy of $$sigma$$=$$pm$$9K and $$sigma$$=$$pm$$16K, respectively, by the ideal solution model.

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