氏名

ヤマジ アキフミ

山路 哲史

職名

准教授 (https://researchmap.jp/7000010159/)

所属

(大学院先進理工学研究科)

連絡先

メールアドレス

メールアドレス
akifumi.yamaji@waseda.jp

住所・電話番号・fax番号

住所
〒169-8555新宿区大久保3-4-1 西早稲田キャンパス51号館11階09B
電話番号
03-5286-8225
fax番号
03-5286-8225

URL等

WebページURL

http://www.f.waseda.jp/akifumi.yamaji/(研究室)

研究者番号
00571704

本属以外の学内所属

兼担

理工学術院(先進理工学部)

学内研究所等

理工学術院総合研究所(理工学研究所)

兼任研究員 2018年-

学歴・学位

学歴

1996年09月-1997年03月 Imperial College London Department of Physics
1997年04月-2001年03月 東京大学 工学部 システム量子工学科
2001年04月-2006年03月 東京大学大学院 工学系研究科 システム量子工学専攻

学位

博士(工学) 課程 東京大学 原子力学

経歴

2006年04月-2014年08月日本原子力研究開発機構(JAEA)研究員
2011年09月-2014年08月経済協力開発機構原子力機関Data BankNuclear Scientist
2014年09月-2017年03月早稲田大学共同原子力専攻講師(専任)
2017年04月-現在:早稲田大学共同原子力専攻准教授

所属学協会

日本原子力学会

委員歴・役員歴(学外)

2014年12月-2017年03月日本原子力学会「社会と共存する魅力的な軽水炉の展望」調査専門委員会委員
2015年04月-2019年03月日本原子力学会海外情報連絡会委員
2016年06月-2018年03月日本原子力学会「燃料デブリ」研究専門委員会委員
2017年12月-2019年03月日本原子力学会熱流動部会「熱水力安全評価基盤技術高度化戦略マップ検討」ワーキンググループ安全評価サブワーキンググループ委員
2018年07月-2020年03月日本原子力学会原子力発電部会「次期軽水炉の技術要件」WG委員
2018年07月-2020年06月日本原子力学会国際活動委員会委員
2018年07月-International Journal of Advanced Nuclear Reactor Designand Technology (JANDTEditorial Board Member
2018年10月-日本原子力研究開発機構1F事故進展基盤研究に関わる分科会主査

受賞

日本原子力学会英文誌最多引用論文賞(共著受賞)

2013年03月

日本原子力学会奨励賞

2007年03月

第29回日本原子力学会熱流動部会部会賞・優秀講演賞

2018年09月授与機関:日本原子力学会

タイトル:Multi-physicsモデリングによる Ex-Vessel溶融物挙動理解の深化(2)全体概要とMPS法によるSpreading解析の高度化

受賞者(グループ):山路 哲史,古谷 正裕,大石 佑治,段 广涛

研究分野

キーワード

原子炉物理学、燃料ふるまい、新型炉、原子炉過酷事故

科研費分類

工学 / 総合工学 / 原子力学

研究テーマ履歴

2015年-2017年事故耐性燃料棒のふるまいと溶融時の挙動解析研究

研究テーマのキーワード:事故耐性燃料、燃料棒ふるまい、溶融燃料挙動

個人研究

2015年-2015年高温スーパー高速炉の燃料棒ふるまい設計研究

研究テーマのキーワード:超臨界圧水冷却、高速炉、燃料ふるまい、燃料棒設計

個人研究

2014年-2014年スーパー高速炉の炉心高温化の研究

研究テーマのキーワード:超臨界圧軽水冷却、高速炉、炉心設計

個人研究

論文

Super light water reactors and super fast reactors: Supercritical-pressure light water cooled reactors

Oka, Yoshiaki; Koshizuka, Seiichi; Ishiwatari, Yuki; Yamaji, Akifumi

Super Light Water Reactors and Super Fast Reactors: Supercritical-Pressure Light Water Cooled Reactors査読有りp.1 - 6512010年12月-2010年12月 

DOIScopus

詳細

概要:Super Light Water Reactors and Super Fast Reactors provides an overview of the design and analysis of nuclear power reactors. Readers will gain an understanding of the conceptual design elements and specific analysis methods for supercritical-pressure light water cooled reactors. Nuclear fuel, reactor core, plant control, plant stand-up and stability are among the topics discussed, in addition to safety system and safety analysis parameters. Providing the fundamentals of reactor design criteria and analysis, this volume is a useful reference for engineers, industry professionals, and graduate students involved with nuclear engineering and energy technology. © Springer Science+Business Media, LLC 2010. All rights reserved.

Analysis of accidents and abnormal transients of a high breeding fast reactor cooled by supercritical-pressure light water

Guo, Rui; Yamaji, Akifumi; Oka, Yoshiaki

Nuclear Engineering and Design査読有り295p.228 - 2382015年12月-2015年12月 

DOIScopus

詳細

ISSN:00295493

概要:© 2015 Elsevier B.V. All rights reserved.A high breeding core of supercritical water cooled fast reactor (Super FBR) is designed with the tightly packed fuel assembly for obtaining a high breeding ratio with negative void reactivity. The coolant volume fraction is substantially smaller than that of the tight lattice fuel assembly in Super FRs. The present study conducted the safety analysis of this reactor for the abnormal transients and accidents at supercritical pressure. The safety system and safety criteria are similar to those of Super FRs. The accident "control rod ejection" gives the highest fuel cladding temperature and the highest peak pressure, although which are still within the limit of safety criteria. The overall results show that all the safety criteria are satisfied at the selected events.

Core design of a high breeding fast reactor cooled by supercritical pressure light water

Someya, Takayuki; Yamaji, Akifumi

Nuclear Engineering and Design査読有り296p.30 - 372016年01月-2016年01月 

DOIScopus

詳細

ISSN:00295493

概要:© 2015 Elsevier B.V. All rights reserved.A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries.

A numerical study of isotropic and anisotropic ablation in MCCI by MPS method

Li, Xin; Yamaji, Akifumi

Progress in Nuclear Energy査読有り90p.46 - 572016年07月-2016年07月 

DOIScopus

詳細

ISSN:01491970

概要:© 2016 Elsevier Ltd. All rights reserved.Anisotropic and isotropic ablation in molten corium-concrete interaction (MCCI) phenomenon was studied with the Moving Particle Semi-implicit (MPS) method by carrying out numerical simulations of CCI-2 and 3 experiments. The interaction of the fully oxidized PWR core melts with specially-designed two-dimensional limestone and siliceous concrete test sections was analyzed, focusing on investigating the two-dimensional ablation behavior with both limestone and siliceous concrete. The phase transition of molten corium and concrete was modeled based on a phase transition model for mixture. Slag film model and crust dissolution models were incorporated in the current MPS code to simulate the effect of gas generation and crust dissolution phenomena in limestone concrete. The effects of gas generation and aggregates on the concrete ablation behavior were investigated by simulating different specially designed cases. The simulation results by MPS method reproduced the isotropic and anisotropic cavity ablation profile and the overall axial and lateral ablation rates agreed well with the experimental measures. The experimental and MPS results both indicate that the crust on the corium-concrete interface can play an important part in concrete ablation process. The simulation results by MPS method also provide evidence to support the theory that aggregates are part of the cause of anisotropic ablation profile in cavity with siliceous concrete because aggregates could delay the axial basemat ablation more significantly than the lateral one and influence the power split in the melt pool.

Melting Penetration Simulation of Fe-U System at High Temperature Using MPS-LER

Mustari, A. P A; Yamaji, A.; Irwanto, Dwi

Journal of Physics: Conference Series査読有り739(1)2016年09月-2016年09月 

DOIScopus

詳細

ISSN:17426588

概要:Melting penetration information of Fe-U system is necessary for simulating the molten core behavior during severe accident in nuclear power plants. For Fe-U system, the information is mainly obtained from experiment, i.e. TREAT experiment. However, there is no reported data on SS304 at temperature above 1350°C. The MPS-LER has been developed and validated to simulate melting penetration on Fe-U system. The MPS-LER modelled the eutectic phenomenon by solving the diffusion process and by applying the binary phase diagram criteria. This study simulates the melting penetration of the system at higher temperature using MPS-LER. Simulations were conducted on SS304 at 1400, 1450 and 1500°C. The simulation results show rapid increase of melting penetration rate.

Development of MPS Method for Analyzing Melt Spreading Behavior and MCCI in Severe Accidents

Yamaji, Akifumi; Li, Xin

Journal of Physics: Conference Series査読有り739(1)2016年09月-2016年09月 

DOIScopus

詳細

ISSN:17426588

概要:Spreading of molten core (corium) on reactor containment vessel floor and molten corium-concrete interaction (MCCI) are important phenomena in the late phase of a severe accident for assessment of the containment integrity and managing the severe accident. The severe accident research at Waseda University has been advancing to show that simulations with moving particle semi-implicit (MPS) method (one of the particle methods) can greatly improve the analytical capability and mechanical understanding of the melt behavior in severe accidents. MPS models have been developed and verified regarding calculations of radiation and thermal field, solid-liquid phase transition, buoyancy, and temperature dependency of viscosity to simulate phenomena, such as spreading of corium, ablation of concrete by the corium, crust formation and cooling of the corium by top flooding. Validations have been conducted against experiments such as FARO L26S, ECOKATS-V1, Theofanous, and SPREAD for spreading, SURC-2, SURC-4, SWISS-1, and SWISS-2 for MCCI. These validations cover melt spreading behaviors and MCCI by mixture of molten oxides (including prototypic UO2-ZrO2), metals, and water. Generally, the analytical results show good agreement with the experiment with respect to the leading edge of spreading melt and ablation front history of concrete. The MPS results indicate that crust formation may play important roles in melt spreading and MCCI. There is a need to develop a code for two dimensional MCCI experiment simulation with MPS method as future study, which will be able to simulate anisotropic ablation of concrete.

Numerical analysis of the melt behavior in a fuel support piece of the BWR by MPS

Chen, Ronghua; Chen, Lie; Guo, Kailun; Yamaji, Akifumi; Furuya, Masahiro; Tian, Wenxi; Su, G. H.; Qiu, Suizheng

Annals of Nuclear Energy査読有り102p.422 - 4392017年04月-2017年04月 

DOIScopus

詳細

ISSN:03064549

概要:© 2017 Elsevier Ltd The fuel support piece in a boiling water reactor (BWR) is used to brace fuel assemblies. The channel within the fuel support piece is determined to be a potential corium relocation path from the core region to the lower head during the severe accident of BWR. In the present study, the improved ∗∗∗Moving Particle Semi-implicit (MPS) method was adopted to simulate the flow and solidification behavior of the melt in a fuel support piece. The MPS method was first validated against the Pb-Bi plate ablation test that was performed by CRIEPI. The predicted ablation mass of the plate agreed well with the experimental results. Then the flowing and freezing behaviors of molten stainless steel (SS) and zircaloy in the fuel support piece were simulated by MPS method with a three dimensional particle configuration, respectively. In this study, the flow and solidification behavior of SS was simulated first. After all the SS passed through the channel, the flowing behavior of Zr in the fuel support piece was simulated. The simulation results indicated that the crust layer formed on the inner surface of the fuel support piece during the melt discharging process. The fuel support piece was plugged by the solidified zircaloy particles in the lower initial temperature case. The fuel support piece kept intact in all the calculation that were performed under the assumed order of melt injection. The present results could help to reveal the progression of a BWR severe accident.

Three-dimensional numerical study on the mechanism of anisotropic MCCI by improved MPS method

Li, Xin; Yamaji, Akifumi

Nuclear Engineering and Design査読有り314p.207 - 2162017年04月-2017年04月 

DOIScopus

詳細

ISSN:00295493

概要:© 2017 Elsevier B.V. In two-dimensional (2-D) molten corium-concrete interaction (MCCI) experiments with prototypic corium and siliceous concrete, the more pronounced lateral concrete erosion behavior than that in the axial direction, namely anisotropic ablation, has been a research interest. However, the knowledge of the mechanism on this anisotropic ablation behavior, which is important for severe accident analysis and management, is still limited. In this paper, 3-D simulation of 2-D MCCI experiment VULCANO VB-U7 has been carried out with improved Moving Particle Semi-implicit (MPS) method. Heat conduction, phase change, and corium viscosity models have been developed and incorporated into MPS code MPS-SW-MAIN-Ver.2.0 for current study. The influence of thermally stable silica aggregates has been investigated by setting up different simulation cases for analysis. The simulation results suggested reasonable models and assumptions to be considered in order to achieve best estimation of MCCI with prototypic oxidic corium and siliceous concrete. The simulation results also indicated that silica aggregates can contribute to anisotropic ablation. The mechanisms for anisotropic ablation pattern in siliceous concrete as well as isotropic ablation pattern in limestone-rich concrete have been clarified from a mechanistic perspective.

Investigation on influence of crust formation on VULCANO VE-U7 corium spreading with MPS method

Yasumura, Yusan; Yamaji, Akifumi; Furuya, Masahiro; Ohishi, Yuji; Duan, Guangtao

Annals of Nuclear Energy査読有り107p.119 - 1272017年09月-2017年09月 

DOIScopus

詳細

ISSN:03064549

概要:© 2017 Elsevier Ltd In a severe accident of a light water reactor, the corium spreading behavior on a containment floor is important as it may threaten the containment vessel integrity. The Moving Particle Semi-implicit (MPS) method is one of the Lagrangian particle methods for simulation of incompressible flow. In this study, the MPS method is further developed to simulate corium spreading involving not only flow, but also heat transfer, phase change and thermo-physical property change of corium. A new crust formation model was developed, in which, immobilization of crust was modeled by stopping the particle movement when its solid fraction is above the threshold and is in contact with the substrate or any other immobilized particles. The VULCANO VE-U7 corium spreading experiment was analyzed by the developed MPS spreading analysis code to investigate influences of different particle sizes, the corium viscosity changes, and the “immobilization solid fraction” of the crust formation model on the spreading and its termination. Viscosity change of the corium was influential to the overall progression of the spreading leading edge, whereas termination of the spreading was primarily determined by the immobilization of the leading edge (i.e., crust formation). The progression of the leading edge and termination of the spreading were well predicted, but the simulation overestimated the substrate temperature. Further investigations may be necessary for the future study to see if thermal resistance at the corium-substrate boundary has significant influence on the overall spreading behavior and its termination.

Flexible core design of Super FBR with multi-axial fuel shuffling

Noda, Shogo; Someya, Takayuki; Yamaji, Akifumi

Nuclear Engineering and Design査読有り324p.45 - 532017年12月-2017年12月 

DOIScopus

詳細

ISSN:00295493

概要:© 2017 Elsevier B.V. To utilize the merit of supercritical water cooling, the Super FBR core concept, which is compatible with both the high breeding and the high enthalpy rise needs to be developed. One possible solution to meet such requirements may be to compose an axially heterogeneous core with MOX and blanket layers, with consideration of the large density change and specific heat of supercritical water at vicinity of the pseudocritical point. A new design concept of Super FBR has been proposed with “with multi-axial fuel shuffling”, which has flexibility in designing fuel shuffling schemes in the lower part and upper part of the core independently. Fully coupled neutronics and thermal-hydraulics core calculations were carried out to investigate impact of designing independent number of fuel batches and shuffling patterns in the upper and the lower parts of the core. Promising results were obtained, showing possibility of improving the core breeding performance with respect to the compound system doubling time (CSDT) by reducing the reactor doubling time (RDT) with designing of the independent fuel shuffling. Moreover, reduction in the ex-core factor (EF) was shown to be possible with such independent fuel shuffling. The combined effects of reductions in RDT and EF showed significant reduction in CSDT. It is the first design concept of Super FBR with coolant enthalpy rise, which covers from the liquid like state (below the pseudo-critical point) to the gas-like state (above the pseudo-critical point) of supercritical water. Further design investigations may be necessary to reduce CSDT and increase the average core outlet temperature.

Safety of Super LWR, (I) Safety System Design

Y. Ishiwatari; Y. Oka; S. Koshizuka; A. Yamaji; J. Liu

Journal of Nuclear Science and Technology査読有り42(11)p.927 - 9342005年05月-2005年05月 

DOI

詳細

ISSN:0022-3131

Safety of Super LWR, (II) Safety Analysis at Supercritical Pressure

Yuki Ishiwatari; Yoshiaki Oka; Seiichi Koshizuka; Akifumi Yamaji; Jie Liu

Journal of Nuclear Science and Technology査読有り42(11)p.935 - 9482005年05月-2005年05月 

DOIScopus

詳細

ISSN:0022-3131

Development of Statistical Thermal Design Procedure to Evaluate Engineering Uncertainty of Super LWR

Jue YANG; Yoshiaki OKA; Jie LIU; Yuki ISHIWATARI; Akifumi YAMAJI

Journal of Nuclear Science and Technology査読有り43(1)p.32 - 422006年01月-2006年01月 

DOIScopus

詳細

ISSN:0022-3131

Fuel and Core Design of Super Light Water Reactor with Low Leakage Fuel Loading Pattern

K. Kamei; A. Yamaji; Y. Ishiwatar; Y. Ok; J. Liu

Journal of Nuclear Science and Technology査読有り43(2)p.129 - 1392006年02月-2006年02月 

DOIScopus

詳細

ISSN:0022-3131

Principle of rationalizing the criteria for abnormal transients of the Super LWR with fuel rod analyses

A. Yamaji; Y. Oka; Y. Ishiwatar; J. Liu; M. Suzuki

Annals of Nuclear Energy査読有り33p.984 - 9932006年08月-2006年08月 

DOIWoS

詳細

ISSN:0306-4549

FEMAXI-6 code verification with MOX fuels irradiated in Halden reactor

Akifumi Yamaji; Motoe Suzuki; Tsutomu Okubo

Journal of Nuclear Science and Technology査読有り46(12)p.1152 - 11612009年12月-2009年12月 

DOIScopus

詳細

ISSN:0022-2131

The Impact of Americium Target In-Core Loading on Reactivity Characteristics and ULOF Response of Sodium-Cooled MOX FBR

A. Yamaji; K. Kawashima; S. Ohki; T. Mizuno; T. Okubo

Nuclear Technology査読有り171(2)p.142 - 1522010年08月-2010年08月 

DOIScopus

詳細

ISSN:0029-5450

Three-dimensional core design of high temperature supercritical-pressure light water reactor with neutronic and thermal-hydraulic

Akifumi Yamaji; Yoshiaki Oka; Seiichi Koshizuka

Journal of Nuclear Science and Technology査読有り42(1)p.8 - 192005年02月-2005年02月 

DOIScopus

詳細

ISSN:0022-2131

概要:The equilibrium core of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H) is designed by three-dimensional neutronic and thermal-hydraulic coupled core calculations. The average coolant core outlet temperature of 500°C is accurately evaluated for the first time in the development of the SCLWR-H. The average coolant core outlet temperature is one of the key parameters, which must be accurately determined in order to establish the concept of this unique reactor design. However, it can only be determined by three-dimensional core design method, taking into account the control rod patterns, fuel loading patterns, coupling of the neutronic and thermal-hydraulic calculations, and burnup distribution of each fuel assembly. The R-Z two-dimensional core design method used in previous studies could not model or evaluate such parameters with sufficient accuracy. In this study, a three-dimensional equilibrium core design method for the SCLWR-H is established. This method can accurately evaluate the average coolant core outlet temperature and has permitted a comprehensive equilibrium core to be developed, which satisfies all design criteria. The design criteria are maximum fuel rod cladding surface temperature of 650°C, maximum linear heat generation rate of 39kW/m, and a positive water density reactivity coefficient.© 2005 Taylor & Francis Group, Ltd.

Improved core design of the high temperature supercritical-pressure light water reactor

A. Yamaji; K. Kamei; Y. Oka; S. Koshizuka

Annals of Nuclear Energy査読有り32(7)p.651 - 6702005年02月-2005年02月 

DOIScopus

詳細

ISSN:0894486802

概要:A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 °C. In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 °C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained. In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies.

Design study to increase plutonium conversion ratio of hc-flwr core

Akifumi Yamaji; Yoshihiro Nakano; Sadao Uchikawa; Tsutomu Okubo

Nuclear Technology査読有り179(3)p.309 - 3222012年09月-2012年09月 

DOIScopus

詳細

ISSN:0029-5450

概要:The innovative water reactor for flexible fuel cycle (FLWR) is an advanced reactor concept based on the well-developed light water reactor (LWR) technology. It is to be introduced in two stages to achieve effective and flexible utilization of the uranium and plutonium resources. In the first stage, the high-conversion-type reactor concept (HC-FLWR) is to be introduced, with a core that achieves a fissile Pu conversion ratio of 0.84. Then, in the second stage, the reduced-moderation water reactor (RMWR) concept can be introduced, with a breeder-type core that achieves a fissile Pu conversion ratio of 1.05. From the viewpoint of effective introduction of the high-conversion-type reactor, such as the introduction capacity of the reactor, HC-FLWR is required to further raise the fissile Pu conversion ratio to ∼0.95. This study aims to develop a new core design concept for the high-conversion-type core, HC-FLWR +, to achieve the higher fissile Pu conversion ratio of ∼0.95 under the framework of UO 2 and U-Pu mixed-oxide (MOX) fuel technologies for LWRs. For raising the fissile Pu conversion ratio and controlling the void reactivity characteristics of the core, the concept of FLWR/ MIX fuel assembly, which uses MOX and enriched UO 2fuel rods, is utilized. The relationships between the main design parameters and the core performance index parameters are clarified in this study. When the fuel rod diameter and the clearance range from 1.23 to 1.28 cm and 0.25 to 0.20 cm, respectively, under the same pitch of 1.48 cm, the fissile Pu conversion ratio and the core average discharge burnup range from 0.89 to 0.94 and 53 to 49 GWd/tonne, respectively (the fissile Pu conversion ratio and the burnup are subject to a trade-off). Furthermore, when 235U enrichment in the UO 2 fuel rods is increased from 4.9 to 6 wt%, the fissile Pu conversion ratio improves to 0.97. From these relationships, two representative core designs with fissile Pu conversion ratios of 0.91 and 0.94 and one optional design with a ratio of 0.97 were obtained. Hence, the flexibility of HC-FLWR + concept to achieve a higher fissile Pu conversion ratio of ∼0.95 has been revealed. Together with the standard HC-FLWR design, the concept covers a wide range of needs on fissile Pu conversion ratio from 0.84 up to 0.97, with design variations that are expected to be within the scope of current boiling water reactor and MOX fuel technologies.

Analysis of Pb-Bi Vessel Wall Ablation Experiment with High Temperature Liquid by MPS Method

Daisuke Masumura; Akifumi Yamaji; Masahiro Furuya

Journal of Energy and Power Engineering11p.944 - 9542015年11月-2015年11月 

DOIScopus

詳細

ISSN:9781510811843

概要:In a severe accident of a light water reactor, ablation of the reactor pressure vessel (RPV) lower head by corium is a key phenomenon, which affects progression of the accident. The Moving Particle Semi- implicit (MPS) method is one of particle methods that calculate behavior of incompressible fluid by semi-implicit method. In preceding studies, MPS models have been developed to analyze phenomena such as heat conduction, phase change, natural convection, thermal stratification, and radiation heat transfer. These phenomena are expected to play key roles in the lower head ablation. This paper aims to investigate whether the MPS method is capable of analyzing the lower head ablation phenomenon, which involves complex interactions of the above mentioned phenomena. The small-scale experiment carried out at Central Research Institute of Electric Power Industry (CRIEPI) using Pb-Bi vessel and silicone oil was analyzed. The heat transfer model was modified for evaluation of heat transfer between the vessel and the oil. The results were compared both qualitatively and quantitatively with the experiment. The former is the comparison of the simulation and experiment regarding phenomena that the liquid ablates the metal vessel and discharges through the vessel wall, which showed good agreement. The latter are comparisons of the calculated liquid temperature, ablation start time and discharge start time with respect to the corresponding measurements. The analyses have shown that the MPS method is capable of analyzing ablation phenomenon qualitatively, but needs further development for quantitative prediction, including investigations on influence of the particle size used in the simulation.

An accurate and stable multiphase moving particle semi-implicit method based on a corrective matrix for all particle interaction

Guangtao Duan; Seiichi Koshizuka; Akifumi Yamaji; Bin Chen; Xin Li; Tasuku Tamai

International Journal for Numerical Methods in Engineering査読有り115(10)p.1287 - 13142018年05月-2018年05月 

DOIScopus

詳細

ISSN:0029-5981

概要:The Lagrangian moving particle semi-implicit (MPS) method has potential to simulate free-surface and multiphase flows. However, the chaotic distribution of particles can decrease accuracy and reliability in the conventional MPS method. In this study, a new Laplacian model is proposed by removing the errors associated with first-order partial derivatives based on a corrected matrix. Therefore, a corrective matrix is applied to all the MPS discretization models to enhance computational accuracy. Then, the developed corrected models are coupled into our previous multiphase MPS methods. Separate stabilizing strategies are developed for internal and free-surface particles. Specifically, particle shifting is applied to internal particles. Meanwhile, a conservative pressure gradient model and a modified optimized particle shifting scheme are applied to free-surface particles to produce the required adjustments in surface normal and tangent directions, respectively. The simulations of a multifluid pressure oscillation flow and a bubble rising flow demonstrate the accuracy improvements of the corrective matrix. The elliptical drop deformation demonstrates the stability/accuracy improvement of the present stabilizing strategies at free surface. Finally, a turbulent multiphase flow with complicated interface fragmentation and coalescence is simulated to demonstrate the capability of the developed method.

Core design of PWR-type seed-blanket core breeder reactor with tightly packed fuel assembly

TetsuoTakei; AkifumiYamaji

Nuclear Engineering and Design333p.45 - 542018年07月-2018年07月 

DOIScopus

詳細

ISSN:0029-5493

概要:Pressurized water reactor (PWR) is the reactor type with the most abundant operation experience in the world. However, studies on designing PWR-type fast reactors have been limited and there have not been any PWR-type fast breeder reactor design concepts. In this study, the concept of seed-blanket PWR-type breeder reactor with tightly packed fuel assembly (TPFA) has been developed by coupled three dimensional neutronics and thermal-hydraulics core calculations. For the seed-blanket heterogeneous core using mixed oxide (MOX) fuel and depleted uranium, it has been shown that the core height is limited to about 1.0 m or less, in order to satisfy the design criterion of negative void reactivity. Moreover, it has been shown that increasing the core power density is difficult, as it leads to substantial increase in the core pressure drop. Consequently, the core characteristics are featured by low breeding performance with compound system doubling time (CSDT) of about 150 years or more (fissile plutonium surviving ratio of below 1.01) and low thermal power of about 700 MW or less. For the core using 4.95 wt% enriched uranium for the blanket assembly, it is possible to improve the void reactivity characteristics, breeding performance and thermal power by reducing reactivity difference between the seed and the blanket fuel assemblies. By utilizing enriched uranium, the concept of breeding PWR core with CSDT of 60 years and thermal power of 1000 MW has been shown. In addition, PWR-type breeder reactor concept using enriched uranium that the fissile surviving ratio including uranium and plutonium exceeds 1 was shown for the first time.

Improved core design of a high breeding fast reactor cooled by supercritical pressure light water

Takayuki Someya; Akifumi Yamaji; Sukarman

Journal of Nuclear Engineering and Radiation Science査読有り4(1)2018年01月-2018年01月 

DOIScopus

詳細

ISSN:2332-8983

概要:The authors look for an attractive light water reactor (LWR) concept, which achieves high breeding performance with respect to the compound system doubling time (CSDT). In the preceding study, a high breeding fast reactor concept, cooled by supercritical pressure light water (Super FBR), was developed using tightly packed fuel assembly (TPFA) concept, in which fuel rods were arranged in a hexagonal lattice and packed by contacting each other. However, the designed concept had characteristics, which had to be improved, such as low power density (7.4 kW/m), large core pressure loss (1.02 MPa), low discharge burnup (core average: 8 GWd/t), and low coolant temperature rise in the core (38 C). The aim of this study is to clarify the main issues associated with improvement of the Super FBR with respect to these design parameters and to show the improved design. The core design is carried out by fully coupled three-dimensional neutronics and single-channel thermal-hydraulic core calculations. The design criteria are negative void reactivity, maximum linear heat generation rate (MLHGR) of 39 kW/m, and maximum cladding surface temperature (MCST) of 650 C for advanced stainless steel. The results show that significant improvement is possible with respect to the core thermal-hydraulic characteristics with minimal deterioration of CSDT by replacing TPFA with the commonly acknowledged hexagonal tight lattice fuel assembly (TLFA). Further design studies are necessary to improve the core enthalpy rise by reducing the radial power swing and power peaking

A novel multiphase MPS algorithm for modeling crust formation by highly viscous fluid for simulating corium spreading

Guangtao Duan; Akifumi Yamaji; Seiichi Koshizuka

Nuclear Engineering and Design査読有り343p.218 - 2312019年01月-

DOI

書籍等出版物

Super Light Water Reactors and Super Fast Reactors

Y. Oka, S. Koshizuka, Y. Ishiwatari, A. Yamaji

Springer2010年-

詳細

ISBN:978-1-4419-6035-1

講演・口頭発表等

AN UPDATE ON THE DEVELOPMENT STATUS OF THE SUPERCRITICAL WATER-COOLED REACTORS

L.K.H. Leung; Y.-P. Huang; V. Dostal; A. Yamaji; A. Sedov

4th GIF Symposium(Generation IV Forum(GIF))2018年10月16日

詳細

国際会議口頭発表(一般)開催地:パリ

概要: The Super-Critical Water-cooled Reactor (SCWR) is a high-temperature, high-pressure watercooled reactor that operates above the thermodynamic critical point of water (374°C, 22.1 MPa). Its main mission is to generate electricity efficiently, economically and safely. Furthermore, the high core outlet temperature of SCWRs (up to 625°C) facilitates co-generation, such as hydrogen production, space heating and steam production. The development of SCWRs has been advanced with the completion of three concepts and a few are being pursued within the Generation-IV International Forum. In addition, the development is being expanded to the SCW small modular reactor for deployment in small remote communities. Recent advancements and the future plan for the SCWR development are described.

Core Design Study of Super FBR with Multi-Axial Fuel Shuffling and Different Coolant Density

Shogo Noda; Sukarman; Akifumi Yamaji; Tetuo Takei; Takanari Fukuda; Arisa Ayukawa

26th International Conference on Nuclear Engineering2018年07月23日

詳細

開催地:ロンドン

Preliminary Core and Fuel Design of BWR with Multi-Axial Fuel Shuffling

Yudai Tasaki; Akifumi Yamaj

2018 International Congress on Advances in Nuclear Power Plants2018年04月

詳細

国際会議口頭発表(一般)開催地:シャーロット、ノースカロライナ

OVERVIEW OF JAPANESE DEVELOPMENT OF ACCIDENT TOLERANT FeCrAl-ODS FUEL CLADDINGSFOR BWRS

K. Sakamoto; M. Hirai; S. Ukai; A. Kimura; A. Yamaji; K. Kusagaya; T. Kondo; S. Yamashita

2017 Water Reactor Fuel Performance Meeting2017年09月

詳細

国際会議開催地:済州島

FEMAXI-7 PREDICTION OF THE BEHAVIOR OF BWR-TYPE ACCIDENT TOLERANT FUEL ROD WITH FECRAL-ODS STEEL CLADDING IN NORMAL CONDITION

Akifumi Yamaji; Daiki Yamasaki; Tomoya Okada; Kan Sakamoto; Shinichiro Yamashita

2017 Water Reactor Fuel Performance Meeting2017年09月

詳細

国際会議口頭発表(招待・特別)開催地:済州道

CONCEPTUAL CORE DESIGN OF BREEDING BWR

Rui Guo; Akifumi Yamaji

25th International Conference on Nuclear Engineering2017年05月

詳細

国際会議開催地:上海

INVESTIGATION ON ACCIDENT PROGRESSION AND MELT BEHAVIOR AT THEFUKUSHIMA DAIICHI UNITS 1&2 USING MELCOR CODE

Shan Zheng; Akifumi Yamaji; Daotong Chong; Junjie Yan; Gen Li

25th International Conference on Nuclear Engineering2017年05月

詳細

国際会議開催地:上海

Sensitivity Study of Accident Scenarios on MCCI for Fukushima Daiichi Unit-1 by MELCOR

Takumi Noju; Akifumi Yamaji; Kiyoshi Matsumoto; Xin Li

2017 International Congress on Advances in Nuclear Power Plants2017年04月

詳細

国際会議口頭発表(一般)開催地:福井、京都

Preliminary Study on Flexible Core Design of Super FBR with Multi- Axial Fuel Shuffling

Sukarman; Akifumi Yamaji; Takayuki Someya; Shogo Noda

2017 International Congress on Advances in Nuclear Power Plants2017年04月

詳細

国際会議口頭発表(一般)開催地:福井、京都

INVESTIGATION TO REDUCE MASS OF A ULTRA-LIGHT SOLID REACTORFOR ELECTRICITY SUPPLY IN ENVIRONMENTS WITHOUT HUMAN MAINTENANCE

Hiroshi Akie; Akifumi Yamaji; Teruhiko Kugo; Takamichi Iwamura; Kenya Suyama

2017 International Congress on Advances in Nuclear Power Plants2017年04月

詳細

開催地:福井、京都

Analysis of Eutectic and Metallic Melt Flow and Blockage in BWR Control Rod Guide Tube by MPS Method

Y.Goto; A.Yamaji

2017年02月

詳細

開催地:ウィーン

Analysis of the Vulcano VE-U7 Corium Spreading Experiment using MPS Method

Yusan Yasumura; Akifumi Yamaji

11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operations and Safety2016年10月

詳細

開催地:慶 州

Numerical Analysis of SURC-1 and SURC-3 MCCI Experiments by MPS Method

Emiko Kibino; Akifumi Yamaji

11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operations and Safety2016年10月

詳細

開催地:慶州

Sensitivity Study of 1F1 Type Accident by MELCOR code

Kenta Saitoa; Akifumi Yamaji

Transaction of ANS Winter meeting 20152015年11月

詳細

開催地:ワシントン

ANALYSIS OF METAL VESSEL WALL ABLATION EXPERIMENT WITH HIGH TEMPERATURE LIQUID BY MPS METHOD

Daisuke Masumura; Yoshiaki Oka; Akifumi Yamaji; Masahiro Furuya

16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015,2015年09月

詳細

国際会議開催地:シカゴ

概要: In a severe accident of a light water reactor, ablation of the reactor pressure vessel (RPV) lower head by corium is a key phenomenon, which affects progression of the accident. The Moving Particle Semi- implicit (MPS) method is one of particle methods that calculate behavior of incompressible fluid by semi-implicit method. In preceding studies, MPS models have been developed to analyze phenomena such as heat conduction, phase change, natural convection, thermal stratification, and radiation heat transfer. These phenomena are expected to play key roles in the lower head ablation. This paper aims to investigate whether the MPS method is capable of analyzing the lower head ablation phenomenon, which involves complex interactions of the above mentioned phenomena. The small-scale experiment carried out at Central Research Institute of Electric Power Industry (CRIEPI) using Pb-Bi vessel and silicone oil was analyzed. The heat transfer model was modified for evaluation of heat transfer between the vessel and the oil. The results were compared both qualitatively and quantitatively with the experiment. The former is the comparison of the simulation and experiment regarding phenomena that the liquid ablates the metal vessel and discharges through the vessel wall, which showed good agreement. The latter are comparisons of the calculated liquid temperature, ablation start time and discharge start time with respect to the corresponding measurements. The analyses have shown that the MPS method is capable of analyzing ablation phenomenon qualitatively, but needs further development for quantitative prediction, including investigations on influence of the particle size used in the simulation.

Numerical simulation of anisotropic ablation of siliceous concrete - Analysis of CCI-3 MCCI experiment by MPS method

Xin Li; Akifumi Yamaji

16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics 20152015年09月

詳細

開催地:シカゴ

Re-Evaluation and Continued Development of Shielding Benchmark Database SINBAD

I.A. Kodeli; P. Ortego; A. Milocco; G. Zerovnik; R. E. Grove; A. Yamaji; E. Sartori

The ANS Reactor Physics Topical Meeting 20142014年09月

詳細

国際会議開催地:京都

Summary and Status of OECD/NEA UAM-LWR Benchmark

M. N. Avramova; K. N. Ivanov; E. Royer; A. Yamaji; J. Gulliford

The ANS Reactor Physics Topical Meeting 20142014年09月

詳細

国際会議開催地:京都

Evaluation of Large 3600MWth Sodium-Cooled Fast Reactor OECD Neutronic Benchmarks

L. Buiron; G. Rimpault; B. Fontaine; T. K. Kim; N. E. Stauff; T. A. Taiwo; A. Yamaji; J. Gulliford

The ANS Reactor Physics Topical Meeting 20142014年09月

詳細

国際会議開催地:京都

Evaluation of Medium 1000MWth Sodium-Cooled Fast Reactor OECD Neutronic Benchmarks

N. E. Stauff; T. K. Kim; T. A. Taiwo; L. Buiron; G. Rimpault; A. Yamaji; J. Gulliford

The ANS Reactor Physics Topical Meeting 20142014年09月

詳細

開催地:京都

Evaluation of uncertainties in FEMAXI-6 calculations for predicting MOX fuel behaviors in FLWR design

Akifumi Yamaji; Motoe Suzuki; Tsutomu Okubo

International Congress on Advances in Nuclear Power Plants 20092009年05月

詳細

国際会議開催地:東京

概要: The concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been proposed and being studied at Japan Atomic Energy Agency (JAEA) to achieve effective and flexible utilization of the uranium and plutonium resources based on the well-developed LWR technology. FLWR is a Boiling Water Reactor (BWR) type concept and it is planned to be introduced by two stages. In the first stage, the MOX fuels are irradiated in a similar condition to that of the current BWR but with a harder neutron spectrum. The core average discharge burnup is about 45 GWd/tHM. In the second stage, the neutron spectrum is further hardened to achieve a multiple recycling of plutonium with higher burnups. In order to design and evaluate the integrities of FLWR fuel rods, the uncertainties in FEMAXI-6 calculations and models for predicting LWR MOX fuel behaviors need to be evaluated. As an introduction to the evaluation process, the Test-Fuel-Data-Base (TFDB) obtained from the Halden reactor experiments (IFA- 514) were used for the evaluations. The maximum discharge burnup was about 40 GWd/tMOX. Based on the present investigation, the following models were found to be particularly important. Namely the fission gas release (FGR), pellet densification, swelling, and relocation models. These models of FEMAXI-6 have been developed and the parameters have been optimized based on the past UO2 irradiation test data. For predicting MOX fuel behavior, the FGR model has a relatively large uncertainty and causes a large uncertainty in the FGR calculations. On the other hand, the uncertainties in the other models are within the range expected by the property variations of typical UO2 fuels. Hence, the densification, swelling, and the relocation models of FEMAXI-6 can be applied to MOX fuel analyses provided the corresponding MOX property variations are taken into account in the input parameters of these models. When the FGR of the MOX fuels is reasonably predicted, the calculated pellet centerline temperature can be expected to be within about plus or minus 50 K of the measurements. These uncertainties need to be taken into account when designing and evaluating the integrities of FLWR fuel rods

Design study of nuclear power systems for deep space explorers (2) electricity supply capabilities of solid cores

Akifumi Yamaji; Takakazu Takizuka; Kunihiko Nabeshima; Takamichi Iwamura; Hajime Akimoto

International Congress on Advances in Nuclear Power Plants 20092009年05月

詳細

国際会議開催地:東京

概要: This study has been carried out in series with the other study, "Criticality of Low Enriched Uranium Fueled Core" to explore the possibilities of a solid reactor electricity generation system for supplying propulsion power of a deep space explorer. The design ranges of two different systems are determined with respect to the electric power, the radiator mass, and the operating temperatures of the heat-pipes and thermoelectric converters. The two systems are the core surface cooling with heat-pipe system (CSHP), and the core direct cooling with heat-pipe system (CDHP). The evaluated electric powers widely cover the 1 to 100 kW range, which had long been claimed to be the range that lacked the power sources in space. Therefore, the concepts shown by this study may lead to a breakthrough of the human activities in space. The working temperature ranges of the main components, namely the heat-pipes and thermoelectric converters, are wide and covers down to relatively low temperatures. This is desirable from the viewpoints of broadening the choices, reducing the development needs, and improving the reliabilities of the devices. Hence, it is advantageous for an early establishment of the concept

Design study of nuclear power systems for deep space explorers (1) criticality of low enriched uranium fueled core

Teruhiko Kugo; Hiroshi Akie; Akifumi Yamaji; Kunihiko Nabeshima; Takamichi Iwamura; Hajime Akikmoto

International Congress on Advances in Nuclear Power Plants 20092009年05月

詳細

国際会議開催地:東京

スーパー軽水炉(超臨界圧軽水炉)の炉心設計

山路哲史

日本原子力学会招待有り2007年09月

スーパー軽水炉の炉心・燃料設計

山路哲史

日本原子力学会熱流動部会・計算科学技術部会Dr.フォーラム招待有り2006年09月

詳細

国内会議

スーパー軽水炉の炉心・燃料設計

山路哲史

革新的水冷却炉研究会(第9回) 招待有り2006年03月01日

詳細

国内会議

Design and Integrity Analyses of the Super LWR Fuel Rod

A.Yamaji; Y.Oka; J.Yang; J.Liu; Y.Ishiwatari; S.Koshizuka

Nuclear Energy Systems for Future Generation and Global Sustainability2005年10月

詳細

国際会議開催地:筑波

Evaluation of the Nominal Peak Cladding Surface Temperature of the Super LWR with Subchannel Analyses

A. Yamaji; T. Tanabe; Y. Oka; J. Yang; J. Liu; Y. Ishiwatari; S. Koshizuka

Nuclear Energy Systems for Future Generation and Global Sustainability2005年10月

詳細

国際会議開催地:筑波

概要: The thermal performance of a nuclear reactor core contains various engineering uncertainties. These uncertainties often arise from calculation, measurement, instrumentation, manufacture, fabrication and data processing. Statistical techniques are useful to evaluate and combine these uncertainties in the thermal design of nuclear reactors. In this paper, a statistical method is developed and employed in the thermal design of the supercritical pressure light water reactor (Super LWR) to evaluate the statistical engineering uncertainties. This method is referred as the Monte Carlo Statistical Thermal Design Procedure for Super LWR (MCSTDP). This method uses the maximum cladding surface temperature (MCST) as a crucial criterion and a sub-channel code is utilized to perform the core thermal hydraulic analysis. The engineering uncertainties are considered with strict respect to the 95/95 limit of Super LWR. The main purpose of this paper is to establish the statistical evaluation methodology. The engineering uncertain for the thermal design of Super LWR is evaluated by using this method to get an approximate quantification. The results are compared with those of the Revised Thermal Design Procedure (RTDP) of Super LWR.

Development of statistical thermal design procedure to evaluate engineering uncertainty of super lwr

Jue Yang; Yoshiaki Oka; Jie Liu; Yuki Ishiwatari; Akifumi Yamaji

Nuclear Energy Systems for Future Generation and Global Sustainability2005年10月

詳細

国際会議開催地:筑波

概要: The thermal performance of a nuclear reactor core contains various engineering uncertainties. These uncertainties often arise from calculation, measurement, instrumentation, manufacture, fabrication and data processing. Statistical techniques are useful to evaluate and combine these uncertainties in the thermal design of nuclear reactors. In this paper, a statistical method is developed and employed in the thermal design of the supercritical pressure light water reactor (Super LWR) to evaluate the statistical engineering uncertainties. This method is referred as the Monte Carlo Statistical Thermal Design Procedure for Super LWR (MCSTDP). This method uses the maximum cladding surface temperature (MCST) as a crucial criterion and a sub-channel code is utilized to perform the core thermal hydraulic analysis. The engineering uncertainties are considered with strict respect to the 95/95 limit of Super LWR. The main purpose of this paper is to establish the statistical evaluation methodology. The engineering uncertain for the thermal design of Super LWR is evaluated by using this method to get an approximate quantification. The results are compared with those of the Revised Thermal Design Procedure (RTDP) of Super LWR.

Rationalization of the Fuel Integrity and Transient Criteria for the Super LWR

Akifumi Yamaji; Yoshiaki Oka; Yuki Ishiwatari; Liu Jie; Seiichi Koshizuka; Motoe Suzuki

International Congress on Advances in Nuclear Power Plants 20052005年05月

詳細

国際会議開催地:ソウル

概要: A Detailed fuel rod design is carried out for the first time in the development of Supercritical-pressure Light Water Reactor (Super LWR). The fuel rod design is similar to that of LWR, consisting of UO2 pellets, a gas plenum and a Stainless Steel Cladding. The principle of rationalization of the criteria for anticipated transients of Super LWR is developed. The fuel rod failures can be conservatively prevented by limiting the cladding strain level within an elastic region, preventing the cladding buckling collapse, and keeping the pellet centerline temperature below its melting point. The FEMAXI-6 fuel analysis code of Japan Atomic Energy Research Institute (JAERI) is used to evaluate the fuel rod integrities in anticipated transient conditions. Detailed analyses have shown that allowable limits to the maximum fuel rod power and maximum cladding temperature can be determined to assure the fuel integrities. These limits may be useful in the plant safety analyses to confirm the fuel integrities during anticipated transients.

Fuel and Core Design of Super LWR with Stainless Steel cladding,

Kazuhiro Kamei; Akifumi Yamaji; Yuki Ishiwatari; Liu Jie; Yoshiaki Oka

International Congress on Advances in Nuclear Power Plants 20052005年05月

詳細

国際会議開催地:ソウル

概要: An equilibrium core of High Temperature Supercritical-pressure Light Water Reactor, now called Super LWR, has been designed. The fuel assemblies loaded in the peripheral region of the core are cooled with descending flow to achieve a high average coolant core outlet temperature. In the present design, Stainless Steel is used for fuel rod claddings and for structural materials. Gd 2O3 concentration and the fuel load pattern are optimized to reduce fuel enrichment. Watts correlations are used for prediction of heat transfer in the core, which takes into account an improvement of heat transfer for downward flow, different from the Oka-Koshizuka correlation used in the previous design. It is found that the water rods with their downward flow need to be thermally insulated with thin ZrO2 layer to keep the moderator temperature below the pseudo critical temperature. As a result, an average coolant core outlet temperature of 500C is achieved. In addition, the effect of various heat transfer correlations on the cladding surface temperature is evaluated

Improved Core Design of High Temperature Supercritical-Pressure Light Water Reactor

A. Yamaji; K. Kamei; Y. Oka; S. Koshizuka

2004 International Congress on Advances in Nuclear Power Plants2004年06月

詳細

国際会議開催地:ペンシルバニア州

概要: A new coolant flow scheme has been devised to raise the average coolant core outlet temperature of the High Temperature Supercritical-Pressure Light Water Reactor (SCLWR-H). A new equilibrium core is designed with this flow scheme to show the feasibility of an SCLWR-H core with an average coolant core outlet temperature of 530 °C. In previous studies, the average coolant core outlet temperature was limited by the relatively low temperature outlet coolant from the core periphery. In order to achieve an average coolant core outlet temperature of 500 °C, each fuel assembly had to be horizontally divided into four sub-assemblies by coolant flow separation plates, and coolant flow rate had to be adjusted for each sub-assembly by an inlet orifice. However, the difficulty of raising the outlet coolant temperature from the core periphery remained. In this study, a new coolant flow scheme is devised, in which the fuel assemblies loaded on the core periphery are cooled by a descending flow. The new flow scheme has eliminated the need for raising the outlet coolant temperature from the core periphery and removed the coolant flow separation plates from the fuel assemblies.

High temperature LWR operationg at supercritical pressure

Yoshiaki Oka; Seiichi Koshizuka; Yuki Ishiwatari; Akifumi Yamaji; Tin Tin Yi

Global 2003: Atoms for Prosperity: Updating Eisenhowers Global Vision for Nuclear Energy2003年11月

詳細

国際会議開催地:ロサンゼルス

概要: The elements of design of high temperature LWR operating at supercritical pressure are described. It includes conceptual design of fuel, core, plant, safety, control and start-up systems. The concept is developed at the University of Tokyo by computer analysis. The feature of the reactor such as economic improvement and hydrogen production potential are described as well as the view from the theory of boiler innovation. The technical background of the concept is LWR and supercritical fossil-fired power technologies. The concept was selected as one of the Generation 4 reactor. The research and development in Japan and in the world are underway. Potential of further design improvement exists

Three-dimensional Core Design of SCLWR-H with Neutronic and Thermal-hydraulic Coupling

A. Yamaji; Y. Oka; S. Koshizuka

Global 2003: Atoms for Prosperity: Updating Eisenhowers Global Vision for Nuclear Energy2003年11月

詳細

国際会議開催地:ロサンゼルス

概要: An SCLWR-H core is designed with 3-D coupled thermo-neutronic core calculations for the first time. The change in power distribution is reflected to the evaluation of water density distribution in the core. The thermal-hydraulic and neutronic calculations are alternatively carried out until convergence is obtained. In the 3-D core calculation, each fuel assembly is modeled and the burnup, power and temperature distributions are evaluated. Fuel enrichment, burnable poisons, reload pattern, control rod pattern and the coolant flow rate distribution are designed to give average core outlet temperature 500C.

Overview of Design Studies of High Temperature Reactor Cooled by Supercritical Light Water at the University of Tokyo

Yoshiaki Oka; Seiichi Koshizuka; Yuki Ishiwatari; Akifumi Yamaji

Global Environment and Advanced Nuclear Power Plants2003年09月

詳細

国際会議開催地:京都

Fuel Design of High Temperature Reactors cooled and Moderated by Supercritical Light Water

Akifumi YAMAJI; Yoshiaki OKA; Seiichi KOSHIZUKA

Global Environment and Advanced Nuclear Power Plants2003年09月

詳細

国際会議開催地:京都

Core Design of a High Temperature Reactor Cooled and Moderated by Supercritical Light Water

A. Yamaji; Y. Oka; S. Koshizuka

Global Environment and Advanced Nuclear Power Plants2003年09月

詳細

国際会議開催地:京都

Conceptual design of high temperature reactors cooled by supercritical light water

Yoshiaki Oka; Seiichi Koshizuka; Yuki Ishiwatari; Akifumi Yamaji

2003 International Congress on Advances in Nuclear Power Plants2003年05月

詳細

国際会議開催地:コルドバ

Elements of Design Consideration of Once-Through Cycle, Supercritical-Pressure Light Water Cooled Reactor

Y. Oka; S. Koshizuka; Y. Ishiwatari; A. Yamaji

2002 International Congress on Advances in Nuclear Power Plants2002年06月

詳細

国際会議開催地:フロリダ

Conceptual Design of a 1,000MWe Supercritical-Pressure Light Water Cooled and Moderated Reactor

Akifumi YAMAJI; Yoshiaki OKA; Seiichi KOSHIZUKA

2001 ANS/HPS Student Conference2001年04月

詳細

国際会議開催地:テキサス

外部研究資金

科学研究費採択状況

研究種別:

事故耐性燃料棒のふるまいと溶融時の挙動解析研究

2015年-0月-2018年-0月

配分額:¥4290000

研究種別:

高速・熱中性子結合炉心の炉物理的研究

2010年-0月-2013年-0月

配分額:¥12610000

学内研究制度

特定課題研究

スーパー高速炉の炉心高温化の研究

2014年度

研究成果概要:原子炉冷却材に超臨界圧水を用い、軽水炉発電技術の経済性の飛躍的な向上が可能なスーパー高速炉のさらなる性能向上のため、新たな炉内流動を検討し、その流動が炉心の核的及び熱的な特性に及ぼす影響を明らかにした。炉心の下部と上部の中間に、超...原子炉冷却材に超臨界圧水を用い、軽水炉発電技術の経済性の飛躍的な向上が可能なスーパー高速炉のさらなる性能向上のため、新たな炉内流動を検討し、その流動が炉心の核的及び熱的な特性に及ぼす影響を明らかにした。炉心の下部と上部の中間に、超臨界圧火力ボイラで用いられるような冷却材混合部を設け、冷却材温度分布の均一化による平均温度の向上を3次元核熱結合炉心計算により検討した。その結果、全発熱長240cmの炉心に対し、炉心下部より130cmの高さ位置に混合部を設ける設計で最も炉心出口温度が高くなり、従来設計に比べて50℃以上高い、554℃を達成する炉心概念を示すことができた。

高温スーパー高速炉の燃料棒ふるまい設計研究

2015年度

研究成果概要:原子炉冷却材に超臨界圧水を用い、軽水炉発電技術の経済性の飛躍的な向上が可能なスーパー高速炉は、高温高圧のため燃料被覆管に既存軽水炉のジルカロイが使えない。本研究では、高温のナトリウム冷却高速炉用に開発されたODSフェライト鋼および...原子炉冷却材に超臨界圧水を用い、軽水炉発電技術の経済性の飛躍的な向上が可能なスーパー高速炉は、高温高圧のため燃料被覆管に既存軽水炉のジルカロイが使えない。本研究では、高温のナトリウム冷却高速炉用に開発されたODSフェライト鋼および軽水炉の事故耐性燃料の候補として検討されているSiCを燃料被覆管としてスーパー高速炉用燃料棒に用いる課題を明らかにするために、軽水炉燃料棒ふるまい解析コードFEMAXI-7を用いてふるまい解析を行った。いずれの候補材料も通常運転時は優れた強度のため良好なふるまいを示したが、応力緩和効果が小さいため、出力過渡時の健全性をさらに検討する必要があることが明らかになった。

金属半球容器アブレーション現象のMPS法による解析手法の発展

2016年度

研究成果概要:原子炉過酷事故時の核燃料の溶融とそれに伴う発熱やガス発生及び周辺構造物との相変化(溶融・凝固)を伴う溶融物の流動(対流、層化)、構造物との相互作用(アブレーション)現象は複雑であり、経験式を多用する従来の手法では正確に予測できない...原子炉過酷事故時の核燃料の溶融とそれに伴う発熱やガス発生及び周辺構造物との相変化(溶融・凝固)を伴う溶融物の流動(対流、層化)、構造物との相互作用(アブレーション)現象は複雑であり、経験式を多用する従来の手法では正確に予測できない。本研究では、溶融物と金属容器プレナム構造の相互作用を機構論的に解析可能なMPS法を開発するために、MPS法によるアブレーション現象の解析手法の高度化を図った。溶融物によるアブレーションに関する実験データが豊富なコア・コンクリート反応実験(MCCI実験)のデータによるMPS法の検証計算を行い、従来用いられていた調整パラメータを使用せずに実験結果を精度よく予測することに成功した。

機能的デブリ分散床の基礎概念研究

2017年度

研究成果概要:本研究では、原子炉過酷事故時に機能的に高温デブリを分散させ、デブリの冷却性を向上する“機能的デブリ分散床”の概念を提案し、その効果を明らかにするために、伝熱・流動・相変化を機構論的に解析できるラグランジュ法に基づくMPS法を用いて...本研究では、原子炉過酷事故時に機能的に高温デブリを分散させ、デブリの冷却性を向上する“機能的デブリ分散床”の概念を提案し、その効果を明らかにするために、伝熱・流動・相変化を機構論的に解析できるラグランジュ法に基づくMPS法を用いて炉心溶融物spreading挙動の理解を深めた。仏国CEAで実施されたVULCANO VE-U7 実験の解析の結果、重力/粘性支配の流動において、流動先端に形成されるクラストと流動の固液相互作用の結果、クラストが次第に発達し、やがてバルク流動をせき止めて流動停止に至る機構が明らかになった。これらの新知見を学術論文誌に発表した。

現在担当している科目

科目名開講学部・研究科開講年度学期
応用物理学実験B先進理工学部2019通年
物理実験B先進理工学部2019通年
応用物理学実験B 【前年度成績S評価者用】先進理工学部2019通年
物理実験B  【前年度成績S評価者用】先進理工学部2019通年
卒業研究先進理工学部2019通年
卒業研究【前年度成績S評価者用】先進理工学部2019通年
原子力理工学概論先進理工学部2019秋学期
原子力理工学概論先進理工学部2019秋学期
原子力理工学概論先進理工学部2019秋学期
卒業研究先進理工学部2019通年
卒業研究  【前年度成績S評価者用】先進理工学部2019通年
原子力発電概論創造理工学部2019春学期
原子力発電概論先進理工学部2019春学期
原子力発電概論先進理工学部2019春学期
Graduation Thesis A (Physics)先進理工学部2019秋学期
Graduation Thesis A (Physics) [S Grade]先進理工学部2019秋学期
Graduation Thesis A (Applied Physics)先進理工学部2019秋学期
Graduation Thesis A (Applied Physics) [S Grade]先進理工学部2019秋学期
Graduation Thesis B (Physics)先進理工学部2019春学期
Graduation Thesis B (Physics) [S Grade]先進理工学部2019春学期
Graduation Thesis B (Applied Physics)先進理工学部2019春学期
Graduation Thesis B (Applied Physics) [S Grade]先進理工学部2019春学期
修士論文(共同原子力専攻)大学院先進理工学研究科2019通年
原子炉物理学特別研究大学院先進理工学研究科2019通年
原子炉物理学特論大学院先進理工学研究科2019春学期
原子力プラント工学・プラント制御特論大学院先進理工学研究科2019秋学期
原子炉物理学演習 I大学院先進理工学研究科2019春学期
原子炉物理学演習 II大学院先進理工学研究科2019秋学期
原子炉実習大学院先進理工学研究科2019集中(春・秋学期)
原子炉物理学特殊研究大学院先進理工学研究科2019通年