2024/04/17 更新

写真a

ヤマジ アキフミ
山路 哲史
所属
理工学術院 大学院先進理工学研究科
職名
教授
学位
博士(工学) ( 東京大学 )

経歴

  • 2022年04月
    -
    継続中

    早稲田大学   理工学術院 先進理工学研究科 共同原子力専攻   教授

  • 2017年04月
    -
    2022年03月

    早稲田大学   共同原子力専攻   准教授

  • 2014年09月
    -
    2017年03月

    早稲田大学   共同原子力専攻   講師(専任)

  • 2011年09月
    -
    2014年08月

    経済協力開発機構原子力機関   Data Bank   Nuclear Scientist

  • 2006年04月
    -
    2014年08月

    日本原子力研究開発機構(JAEA)   研究員

学歴

  • 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  

  • 1997年04月
    -
    2001年03月

    東京大学   工学部   システム量子工学科  

  • 1996年09月
    -
    1997年03月

    Imperial College London   Department of Physics   Department of Physics  

委員歴

  • 2019年09月
    -
    2023年07月

    日本原子力学会  標準委員会基盤応用・廃炉技術専門部会 委員

  • 2019年04月
    -
    2022年03月

    原子力規制委員会  技術評価検討会 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
    -
    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

  • 2019年09月
    -
     

    Generation IV International Forum  SCWRシステム運営委員会(SSC)委員及びGIF国内連絡会 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

  • 2015年04月
    -
    2019年03月

    日本原子力学会海外情報連絡会  委員

  • 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

  • 2018年07月
    -
     

    International Journal of Advanced Nuclear Reactor Designand Technology (JANDT)  Editorial Board Member

  • 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

  • 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

▼全件表示

所属学協会

  •  
     
     

    日本原子力学会

研究分野

  • 原子力工学

研究キーワード

  • 原子炉物理学、燃料ふるまい、新型炉、原子炉過酷事故

受賞

  • 早稲田大学ティーチングアワード(2019年度秋学期)

    2020年08月   早稲田大学   原子力理工学概論  

    受賞者: 鷲尾 方一, 山路 哲史, 古谷 正裕

  • 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   日本原子力学会   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: 山路 哲史, 古谷 正裕, 大石 佑治, 段 广涛

  • 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  

 

論文

  • Development of FEMAXI-ATF for analyzing PCMI behavior of SiC cladded fuel under power ramp conditions

    Yoshihiro Kubo, Akifumi Yamaji

    Nuclear Engineering and Technology   56 ( 3 ) 846 - 854  2024年03月  [査読有り]

    DOI

    Scopus

  • Validating ground-based aerodynamic levitation surface tension measurements through a study on Al<inf>2</inf>O<inf>3</inf>

    Yifan Sun, Guangtao Duan, Akifumi Yamaji, Tomoya Takatani, Hiroaki Muta, Yuji Ohishi

    npj Microgravity   8 ( 1 )  2022年12月

     概要を見る

    The surface tension of a molten sample can be evaluated based on its resonant frequency with various levitation techniques. Under a 1-G condition, the use of levitation forces to counteract gravity will cause the levitated sample’s resonant frequency to differ from that under microgravity. A mathematical relationship to correct for this deviation is not available for a sample levitated with aerodynamic levitation (ADL), which raises issues on the validity of surface tension measurements done with ADL. In this study, we compared the surface tension of molten Al2O3 obtained using the front tracking (FT) simulation method, the drop-bounce method with ADL, and the oscillating drop method with ADL. The drop-bounce method simulates microgravity by allowing the sample to free-fall over a period of tens of milliseconds. Based on the results of this comparison, we determined that the surface tension of molten materials measured with ground-based ADL with the oscillating drop method, calculated using the resonant frequency of the l=2 m=0 mode, only shows a small deviation from that obtained under microgravity.

    DOI

    Scopus

    3
    被引用数
    (Scopus)
  • A new potential interface tension model for MPS method avoiding unphysical particle cohesion

    Takanari Fukuda, Xin Li, Akifumi Yamaji

    Progress in Nuclear Energy   150   104311 - 104311  2022年08月

    DOI

    Scopus

  • Estimation of long-term ex-vessel debris cooling behavior in Fukushima Daiichi Nuclear Power Plant unit 3

    Ikken SATO, Akifumi YAMAJI, Xin LI, Hiroshi MADOKORO

      9 ( 2 ) 21 - 00436  2022年02月  [査読有り]

    DOI

  • A multiphase MPS method coupling fluid–solid interaction/phase-change models with application to debris remelting in reactor lower plenum

    Guangtao Duan, Akifumi Yamaji, Mikio Sakai

    Annals of Nuclear Energy   166   108697 - 108697  2022年02月  [査読有り]

    DOI

    Scopus

    15
    被引用数
    (Scopus)
  • Estimation of debris relocation and structure interaction in the pedestal of Fukushima Daiichi Nuclear Power Plant Unit-3 with Moving Particle Semi-implicit (MPS) method

    Xin Li, Akifumi Yamaji, Guangtao Duan, Ikken Sato, Masahiro Furuya, Hiroshi Madokoro, Yuji Ohishi

    Annals of Nuclear Energy   169   108923 - 108923  2021年12月  [査読有り]

    DOI

    Scopus

    2
    被引用数
    (Scopus)
  • Analysis of the localized metallic phase solidification in VULCANO VF-U1 with MPS method

    Takanari Fukuda, Akifumi Yamaji, Xin Li, Jean-François Haquet, Anne Boulin

    Nuclear Engineering and Design   385   111537 - 111537  2021年12月  [査読有り]

    担当区分:最終著者

    DOI

    Scopus

    4
    被引用数
    (Scopus)
  • Development of accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan

    K. Sakamoto, Y. Miura, S. Ukai, N.H. Oono, A. Kimura, A. Yamaji, K. Kusagaya, S. Takano, T. Kondo, T. Ikegawa, I. Ioka, S. Yamashita

    Journal of Nuclear Materials   557   153276 - 153276  2021年12月  [査読有り]

    DOI

    Scopus

    40
    被引用数
    (Scopus)
  • Estimation of the fuel debris thermal energy at the time of the major core slumping of Fukushima Daiichi Nuclear Power Plant Unit-3 with MELCOR-2.2

    Mariko Regalado, Xin Li, Akifumi Yamaji, Ikken Sato

    Annals of Nuclear Energy   160   108430 - 108430  2021年09月  [査読有り]

    DOI

    Scopus

    2
    被引用数
    (Scopus)
  • 2D MPS method analysis of ECOKATS-V1 spreading with crust fracture model

    Jubaidah, Yuki Umazume, Nozomu Takahashi, Xin Li, Guangtao Duan, Akifumi Yamaji

    Nuclear Engineering and Design   379   111251 - 111251  2021年08月  [査読有り]

    DOI

    Scopus

    5
    被引用数
    (Scopus)
  • Development of MPS method and analytical approach for investigating RPV debris bed and lower head interaction in 1F Units-2 and 3

    Nozomu Takahashi, Guangtao Duan, Akifumi Yamaji, Xin Li, Ikken Sato

    Nuclear Engineering and Design   379   111244 - 111244  2021年08月  [査読有り]

    DOI

    Scopus

    10
    被引用数
    (Scopus)
  • Improvement of solidification model and analysis of 3D channel blockage with MPS method

    Reo Kawakami, Xin Li, Guangtao Duan, Akifumi Yamaji, Isamu Sato, Tohru Suzuki

    Frontiers in Energy   15 ( 4 ) 946 - 958  2021年06月  [査読有り]

    DOI

    Scopus

    5
    被引用数
    (Scopus)
  • Sensitivity analysis of core slumping and debris quenching behavior of Fukushima Daiichi Unit-3 accident

    Xin Li, Ikken Sato, Akifumi Yamaji, Mariko Regalado, Jun Wang

    Annals of Nuclear Energy   150   107819 - 107819  2021年01月  [査読有り]

    DOI

    Scopus

    3
    被引用数
    (Scopus)
  • An incompressible–compressible Lagrangian particle method for bubble flows with a sharp density jump and boiling phase change

    Guangtao Duan, Akifumi Yamaji, Mikio Sakai

    Computer Methods in Applied Mechanics and Engineering   372   113425 - 113425  2020年12月  [査読有り]

    DOI

    Scopus

    17
    被引用数
    (Scopus)
  • Preliminary Core Design Study of Small Supercritical Fast Reactor with Single-Pass Cooling

    Kyota Uchimura, Akifumi Yamaji

      1 ( 1 ) 46 - 53  2020年11月  [査読有り]

    DOI

  • Imposing accurate wall boundary conditions in corrective‐matrix‐based moving particle semi‐implicit method for free surface flow

    Guangtao Duan, Takuya Matsunaga, Akifumi Yamaji, Seiichi Koshizuka, Mikio Sakai

      93 ( 1 ) 148 - 175  2020年08月  [査読有り]

    DOI

    Scopus

    24
    被引用数
    (Scopus)
  • A review on MPS method developments and applications in nuclear engineering

    Gen Li, Jinchen Gao, Panpan Wen, Quanbin Zhao, Jinshi Wang, Junjie Yan, Akifumi Yamaji

    Comput. Methods Appl. Mech. Engrg   367  2020年08月  [査読有り]

  • Conceptual design of Super FR for MA transmutation with axially heterogeneous core

    Takanari Fukuda, Akifumi Yamaji

    Nuclear Engineering and Design, Vol.363(2020)   363  2020年06月  [査読有り]

     概要を見る

    Part of this work was conducted under “Understanding Mechanisms of Severe Accidents and Improving Safety of Nuclear Reactors by Computer Science” of Institute for Advanced Theoretical and Experimental Physics and Waseda Research Institute for Science and Engineering.

  • Recent progress in development of accident tolerant fecral-Ods fuel claddings for BWRs in Japan

    K. Sakamoto, Y. Miura, S. Ukai, A. Kimura, A. Yamaji, K. Kusagaya, S. Yamashita

    GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water Reactor Fuel Performance Conference     197 - 205  2020年

     概要を見る

    Copyright © GLOBAL 2019 - International Nuclear Fuel Cycle Conference and TOP FUEL 2019 - Light Water Reactor Fuel Performance Conference.All rights reserved. This paper introduces some highlights of the current progress of development of accident tolerant FeCrAl-ODS (oxide dispersion strengthened) fuel claddings for BWRs (boiling water reactors) in the program sponsored and organized by the Ministry of Economy, Trade and Industry (METI) of Japan. Both the experimental and the analytical studies in JFY2018 resulted in a successful step to develop the technical basis to introduce the FeCrAl-ODS fuel claddings to the current BWRs.

  • Investigation on corium spreading over ceramic and concrete substrates in VULCANO VE-U7 experiment with moving particle semi-implicit method

    Jubaidah, Guangtao Duan, Akifumi Yamaji, Christophe Journeau, Laurence Buffe, Jean-Francois Haquet

    Annals of Nuclear Energy   141  2019年12月  [査読有り]

     概要を見る

    The potential reasons for the corium spreading difference over inert ceramic and reactive concrete channels of VULCANO VE-U7 experiment are investigated using Moving Particle Semi-implicit (MPS) method. A new thermal contact resistance model has been developed for MPS so that influence of the subscale heat transfer between the melt/crust and the substrate on spreading can be considered. The results indicate that the spreading difference is not much influenced by heat loss of the melt to different substrates, but more likely due to gas bubbles in the concrete channel. The most likely responsible gas bubble effect could not be well identified with the single channel analysis, because it could not consider the inflow mass interactions at the stabilization pool. The double-channel analysis with such consideration indicated enhancement of the effective thermal conductivity of the melt as the key influence of the gas bubbles that led to the difference.

    DOI

    Scopus

    24
    被引用数
    (Scopus)
  • Analysis of hemispherical vessel ablation failure involving natural convection by MPS method with corrective matrix

    Nozomu Takahashi, Guangtao Duan, Masahiro Furuya, Akifumi Yamaji

    International Journal of Advanced Nuclear Reactor Design and Technology   1   19 - 29  2019年10月  [査読有り]

     概要を見る

    In a severe accident of a light water reactor, the reactor pressure vessel (RPV) lower head may fail due to ablation at the vessel wall boundary involving natural convection of molten core materials. Accurate prediction of RPV lower head failure is essential for assessing severe accident progression and improving accident management because it greatly influences the subsequent ex-vessel accident progressions. However, there have been still large uncertainties about RPV lower head failure mode in the Fukushima Daiichi Nuclear Accident in 2011. The Lagrangian based MPS (moving particle semi-implicit) method has advantage of analyzing such phenomena involving complex interfaces and liquid-solid phase changes over other Eulerian mesh-based method. In the preceding study, small-scale Pb–Bi hemisphere vessel ablation experiment, with silicone oil as simulated molten core, was reproduced qualitatively by original MPS method. However, ablation mechanism associated with natural convection of the high temperature liquid could not be discussed because of significant influence of numerical discretizing error. In this study, the improved MPS method coupling corrective matrix in the particle interaction model which largely suppress the numerical fluctuation was adopted to analyze the experiment. The results show that the ablated metal relocation may enhance convective heat transfer in the downstream. As a result, ablation of the vessel wall extends from the level, close to the silicone oil surface down to the bottom of the vessel rather than previously simulated localized ablation near the silicone oil surface.

    DOI

    Scopus

    14
    被引用数
    (Scopus)
  • Improving breeding performance of Super FR with fuel shuffling in multi-axial layers

    Sukarman, Shogo Noda, Takanari Fukuda, Akifumi Yamaji

    Journal of Nuclear Engineering and Design   355 ( 110323 ) 110323 - 110323  2019年09月  [査読有り]

     概要を見る

    Super FR is the fast reactor version of SuperCritical Light Water Reactor (SCWR), which operates under supercritical condition of light water, so that designing the core with high coolant outlet temperature above the pseudo-critical point is possible. To raise the core outlet temperature of Super FR, axially heterogeneous core concept was proposed by the preceding study. It consisted of alternatively arranged two seed layers and two blanket layers (i.e., four-layers core). Furthermore, the core was axially divided to two sections and fuel shuffling was considered in each of the two axial sections independently to improve breeding performance. However, the study assumed the same fuel depletion history along the axial direction of the core, despite the large coolant density change. The core outlet temperature was only 387 C and the Compound System Doubling Time (CSDT) was about 98 years, which was not as short as desired. Hence, this study aims to reveal more comprehensive understanding of the concept of improving the performance of Super FR with fuel shuffling in multi-axial layers through design analyses. A five-layers core is considered, which consists of the lower blanket, lower seed, inner blanket, upper seed, and the upper blanket layers. Influences of the lower blanket/upper blanket discharge burnup on the core characteristics, such as the CSDT, void reactivity, and core outlet temperature have been studied with different coolant inlet temperatures for the first time. Different fuel depletion histories are considered for each of the axial layers with consideration of the axial coolant density change (which had not been considered in the preceding study). The present results show that it is preferable to keep the lower blanket discharge burnup relatively low, compared with that of the upper blanket to improve breeding performance with respect to achieving short CSDT. However, it leads to slight reduction in the core outlet temperature. Based on the sensitivity analyses, a representative design is proposed, which achieves CSDT of 74 years and average core outlet temperature of 492 C, which show significant improvement from the preceding design.

    DOI

    Scopus

    2
    被引用数
    (Scopus)
  • The truncation and stabilization error in multiphase moving particle semi-implicit method based on corrective matrix: Which is dominant?

    Guangtao Duan, Akifumi Yamaji, Seiichi Koshizuka, Bin Chen

    Computers & Fluids   190   254 - 273  2019年08月  [査読有り]

     概要を見る

    The Lagrangian nature of the moving particle semi-implicit (MPS) method brings two challenges: disordered particle distribution and particle clumping. The former can cause large random discretization error for the original MPS models while corrective matrix can effectively reduce such large error to the high-order truncation error. The latter can trigger instability easily and thus some adjustment strategies for stability are indispensable, thereby causing non-negligible stabilization error. The purpose of this paper is to compare the relative magnitude of the truncation and stabilization error, which is of great significance for future improvements. An indirect approach is developed because of the difficulty of separating different error from total error in dynamic simulations. The basic idea is to check whether the total error decreases significantly after the truncation error is further reduced. First, a second order corrective matrix (SCM) is proposed for MPS to reduce the truncation error further, as demonstrated by theoretical error analysis. Second, error analysis reveals that the first order gradient model produces less numerical diffusion than the second order gradient model in interpolation after particle shifting. Then, several numerical examples, including Taylor-Green vortex, elliptical drop deformation, excited pressure oscillation flow and continuous oil spill flow, are simulated to test the variance of total error after SCM is applied. It is found that the SCM schemes basically did not remarkably decrease the total error for incompressible free surface flow, implying that truncation error is not dominant compared to the stabilization error. Therefore, reducing the stabilization error is of more significance in future.

    DOI

    Scopus

    54
    被引用数
    (Scopus)
  • Density and viscosity of liquid ZrO2 measured by aerodynamic levitation technique

    Toshiki Kondo, Hiroaki Muta, Ken Kurosaki, Florian Kargl, Akifumi Yamaji, Masahiro Furuya, Yuji Ohishi

    Heliyon   5 ( 7 )  2019年07月  [査読有り]

     概要を見る

    Liquid ZrO2 is one of the most important materials involved in severe accident analysis of a light-water reactor. Despite its importance, the physical properties of liquid ZrO2 are scarcely reported. In particular, there are no experimental reports on the viscosity of liquid ZrO2. This is mainly due to the technical difficulties involved in the measurement of thermo-physical properties of liquid ZrO2, which has an extremely high melting point. To address this problem, an aerodynamic levitation technique was used in this study. The density of liquid ZrO2 was calculated from its mass and volume, estimated based on the recorded image of the sample. The viscosity was measured by a droplet oscillation technique. The density and viscosity of liquid ZrO2 at temperatures ranging from 2753 K to 3273 K, and 3170 K–3471 K, respectively, were successfully evaluated. The density of liquid ZrO2 was found to be 4.7 g/cm 3 at its melting point of 2988 K and decreased linearly with increasing temperature, and the viscosity of liquid ZrO2 was 13 mPa at its melting point.

    DOI

    Scopus

    45
    被引用数
    (Scopus)
  • Ablation analysis with MPS for proposing ex-vessel corium spreading management in light water reactors

    Masafumi Katta, Guangtao Duan, Akifumi Yamaji, Masahiro Furuya

    International Conference on Nuclear Engineering, Proceedings, ICONE   2019-May  2019年05月

     概要を見る

    In a postulated sever accident of a light water reactor (LWR), molten core debris (corium) may breach the reactor pressure vessel and be released to the ex-vessel containment floor. A core catcher manages ex-vessel corium cooling by uniformly spreading the corium into a large space, but it requires a dedicated plant design. In contrast, corium shields have been back-fitted to some boiling water reactors to prevent excessive amount of corium to flow into sump pits, where effective corium cooling may be difficult. However, corium shields can only block the corium flow and cannot contribute to uniform spreading of the ex-vessel corium. This study proposes a preliminary concept of “Debris Spreading Floor”, which can be applied to any types of reactor plants including back-fitting to the existing plants. More specifically, the existing containment floor is overlaid with sacrificial material and refractory material is placed around sump pits. It is intended to allow the original function of sump pits to collect leaking water under normal, abnormal transient and design basis accident conditions. However, under postulated severe accident condition, spreading of ex-vessel corium is promoted by ablating itself with the hot corium and guiding corium spreading away from sump pits. To develop the concept, mechanistic analysis of corium spreading, which can consider influence of substrate ablation, is needed. The Moving Particle Semi-implicit (MPS) method is a Lagrangian particle method and thus suitable for mechanistic simulation of free-surface spreading flow involving solid / liquid phase change and interactions. In this research, effect of the proposed concept is firstly presented with the MPS simulations. Preliminary simulations in 2D show that, amount of corium flowing into sump pits is reduced by the concept. Secondly, validity of the MPS simulations is quantitatively discussed by simulating the experiments with simulant. The experiment was carried out by Central Research Institute of Electric Power Industry (CRIEPI) by pouring liquid Pb-Bi onto a Pb-Bi block so that the inflow liquid spreads on the block surface while it also ablates the block. Sensitivity analyses have been carried out with different initial conditions, calculation resolutions, subscale models and parameters of the MPS simulations to identify the key models and parameters for quantitative prediction of the melt / substrate interactions.

  • Sensitivity analysis of in-vessel accident progression behavior in Fukushima Daiichi Nuclear Power Plant Unit 3

    Xin Li, Ikken Sato, Akifumi Yamaji

    Annals of Nuclear Energy   133   21 - 34  2019年05月  [査読有り]

     概要を見る

    The Great East Japan earthquake and the subsequent tsunami which occurred on March 11th 2011 put
    the operating Units 1–3 at Fukushima Daiichi Nuclear Power Plant (NPP) in severe accident conditions
    resulting from loss of offsite power and AC power. It is believed that the Station Blackout (SBO) and loss
    of heat sink led to core meltdown in all Units 1–3. Despite past research efforts on the severe accident
    progression in Fukushima NPP Units 1–3, there are still knowledge gaps and uncertainties existing in
    understanding of the severe accident scenarios and consequences.
    Hence, this study aims at identifying the modeling uncertainties and addressing the sensitivity parameters
    in Fukushima NPP Unit 3. A more detailed Control Volume (CV) division model of the reactor core
    region has been developed to better simulate the thermal-hydraulic behavior of liquid water and steam,
    which is considered to be crucial in simulating the core uncovery and degradation process. The boundary
    conditions such as the water injection rates by the Reactor Core Isolation Cooling (RCIC) system, the High
    Pressure Core Injection (HPCI) system and Alternative Water Injection (AWI) to the reactor core were
    determined based on the available reactor water level and pressure measurement data. The current study
    suggested that the local vapor heatup behavior could influence the core melting and relocation behavior,
    which can lead to different core degradation scenarios. With the current modeling assumptions in
    MELCOR, the best estimate conditions for RPV pressure history of Unit 3 suggested that 6 SRVs could have
    remained open when the major core slumping took place at ca. 45:20 h (ca. 12:00, March 13) with 50 to
    80 tons of water inventory in the lower plenum. The current analysis also suggested that the efficiency of
    the AWI to the reactor core could have been only 15% as of reported by TEPCO with the current modeling
    conditions if debris dryout was assumed to have occurred at around ca. 54.0 h (20:40 h, March 13th). As
    for lower head failure, there is still large uncertainty in predicting lower head failure time with Larson-
    Miller creep rupture model in the current MELCOR modeling.

    DOI

    Scopus

    6
    被引用数
    (Scopus)
  • 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   343   218 - 231  2019年01月  [査読有り]

     概要を見る

    Corium (lava-like mixture of fissile material) spreading prediction is of great significance in the severe accidents
    of nuclear power plants. Crust formation due to solidification distinguishes corium spreading from common
    isothermal spreading. The Lagrangian moving particle semi-implicit (MPS) method is potential for such
    spreading flow with both free surface and crust-melt interface. Crust formation is usually represented by viscosity
    escalation, but crust creeping is an associated problem. In the original MPS algorithm, creeping velocity
    cannot be reduced steadily by the continuous increase of viscosity, owing to the numerical creeping. A new
    solution algorithm is proposed for particle methods to eliminate such numerical creeping, so that creeping
    velocity decreases proportionally with viscosity rise. In this situation, high enough viscosity can effectively
    represent crust behaviors. Three numerical examples, leakage flow with high viscosity, dam break flow with low
    viscosity and the VULCANO VE-U7 corium spreading experiment with both high and low viscosities simultaneously,
    are investigated to contrast the performance difference between the original and new algorithms. It is
    demonstrated that the current algorithm is suitable for crust formation in corium spreading.

    DOI

    Scopus

    57
    被引用数
    (Scopus)
  • Core design of PWR-type seed-blanket core breeder reactor with tightly packed fuel assembly

    Tetuso Takei, Akifumi Yamaji

    Nuclear Engineering and Design   333   45 - 54  2018年07月  [査読有り]

     概要を見る

    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.

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  • 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   115 ( 10 ) 1287 - 1314  2018年05月  [査読有り]

     概要を見る

    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.

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    87
    被引用数
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  • 複雑流路内の三次元流動観察と数値混相流体力学解析

    古谷 正裕, 山路 哲史, 大石 佑治

    年次大会   2018   S0530306  2018年

     概要を見る

    <p>A pedestal-floor structure of a nuclear-reactor containment-vessel was additive manufactured with two sump-pits and a doorway at a reduced scale of 1:100 and 1:50. A silicone-oil jet impinged and spread on the pedestal floor. Such three-dimensional flow was visualized from three cameras. A volume of fluid (VOF) simulation was successfully reproduced this three-dimensional spreading behavior, silicone-oil flow-thickness evolution and drainage weights.</p>

    DOI CiNii

  • 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   4 ( 1 )  2018年01月  [査読有り]

     概要を見る

    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.

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    1
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  • Flexible core design of Super FBR with multi-axial fuel shuffling

    Shogo Noda, Takayuki Someya, Akifumi Yamaji

    NUCLEAR ENGINEERING AND DESIGN   324   45 - 53  2017年12月  [査読有り]

     概要を見る

    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 pseudocritical 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.

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    8
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  • Investigation on influence of crust formation on VULCANO VE-U7 corium spreading with MPS method

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

    ANNALS OF NUCLEAR ENERGY   107   119 - 127  2017年09月  [査読有り]

     概要を見る

    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. (C) 2017 Elsevier Ltd. All rights reserved.

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    19
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  • Three-dimensional numerical study on the mechanism of anisotropic MCCI by improved MPS method

    Xin Li, Akifumi Yamaji

    NUCLEAR ENGINEERING AND DESIGN   314   207 - 216  2017年04月  [査読有り]

     概要を見る

    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, 3-5 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. (C) 2017 Elsevier B.V. All rights reserved.

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    25
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  • Numerical analysis of the melt behavior in a fuel support piece of the BWR by MPS

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

    ANNALS OF NUCLEAR ENERGY   102   422 - 439  2017年04月  [査読有り]

     概要を見る

    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 perfornied 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. (C) 2017 Elsevier Ltd. All rights reserved.

    DOI

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    29
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  • Development of MPS Method for Analyzing Melt Spreading Behavior and MCCI in Severe Accidents

    Akifumi Yamaji, Xin Li

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

     概要を見る

    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.

    DOI

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    15
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  • Melting Penetration Simulation of Fe-U System at High Temperature Using MPS-LER

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

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

     概要を見る

    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.

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    2
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  • A numerical study of isotropic and anisotropic ablation in MCCI by MPS method

    Xin Li, Akifumi Yamaji

    PROGRESS IN NUCLEAR ENERGY   90   46 - 57  2016年07月  [査読有り]

     概要を見る

    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. (C) 2016 Elsevier Ltd. All rights reserved.

    DOI

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    38
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  • Core design of a high breeding fast reactor cooled by supercritical pressure light water

    Takayuki Someya, Akifumi Yamaji

    NUCLEAR ENGINEERING AND DESIGN   296   30 - 37  2016年01月  [査読有り]

     概要を見る

    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. (C) 2015 Elsevier B.V. All rights reserved.

    DOI

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    11
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  • Analysis of accidents and abnormal transients of a high breeding fast reactor cooled by supercritical-pressure light water

    Rui Guo, Akifumi Yamaji, Yoshiaki Oka

    NUCLEAR ENGINEERING AND DESIGN   295   228 - 238  2015年12月  [査読有り]

     概要を見る

    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. (C) 2015 Elsevier B.V. All rights reserved.

    DOI

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    1
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  • 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 Engineering   11   944 - 954  2015年11月  [査読有り]

     概要を見る

    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.

    DOI

  • Multi-physics and multi-scale benchmarking and uncertainty quantification within OECD/NEA framework

    M. Avramova, K. Ivanov, T. Kozlowski, I. Pasichnyk, W. Zwermann, K. Velkov, E. Royer, A. Yamaji, J. Gulliford

    Annals of Nuclear Energy   84   178 - 196  2015年10月  [査読有り]

    DOI

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    14
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  • Analysis of metal vessel wall ablation experiment with high temperature liquid by MPS method

    Daisuke Masumura, Yoshiaki Oka, Akifumi Yamaji, Masahiro Furuya

    International Topical Meeting on Nuclear Reactor Thermal Hydraulics 2015, NURETH 2015   9   7401 - 7413  2015年

     概要を見る

    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.

  • DESIGN STUDY TO INCREASE PLUTONIUM CONVERSION RATIO OF HC-FLWR CORE

    Akifumi Yamaji, Yoshihiro Nakano, Sadao Uchikawa, Tsutomu Okubo

    NUCLEAR TECHNOLOGY   179 ( 3 ) 309 - 322  2012年09月  [査読有り]

     概要を見る

    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 similar 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 similar to 0.95 under the framework of UO2 and U-Pit 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 UO2 fuel 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 U-235 enrichment in the UO2 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 similar to 0.95 has been revealed. Together with the standard HC-FLWR design, the concept covers a wide range of needs on fissile Pit 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.

    DOI

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    6
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  • THE IMPACT OF AMERICIUM TARGET IN-CORE LOADING ON REACTIVITY CHARACTERISTICS AND ULOF RESPONSE OF SODIUM-COOLED MOX FBR

    Akifumi Yamaji, Katsuyuki Kawashima, Shigeo Ohki, Tomoyasu Mizuno, Tsutomu Okubo

    NUCLEAR TECHNOLOGY   171 ( 2 ) 142 - 152  2010年08月  [査読有り]

     概要を見る

    The idea of recycling minor actinides (MAs) with fast breeder reactors (FBRs) is an effective way to potentially reduce environmental burdens associated with nuclear energy production. For such FBR cores, it is necessary to find one or more promising MA loading methods that can effectively transmute MAs while minimizing deterioration of the core performance and reducing the overall fuel fabrication cost. In this study, the homogeneous MA loading core with 3 wt% MAs is used as a reference design to evaluate the impact of the americium (Am) target in-core loading on reactivity characteristics and unprotected loss-of-flow (ULOF) response of sodium-cooled mixed-oxide FBR.
    The Am target loading core of this study is designed by roughly preserving the MA inventory of the homogeneous MA loading core while placing Am and curium (Cm) to the ring-shaped target region between the inner and the outer core regions with 20 wt% content.
    This design can flatten core radial reactivity worth distributions and effectively reduce reactivity insertion into the core during ULOF compared with the homogeneous MA loading core. It also has relatively flat and stable radial power distributions, which allow a relatively large coolant flow rate to be distributed to the target region.
    During ULOF, the power increase of the Am target loading core of this study is slower than that of the homogeneous MA loading core. The maximum fuel temperature of the target region does not become particularly high compared with that of the inner core, and it is much lower than the melting point. Hence, the proposed Am target in-core loading method does not have a significant influence on ULOF response of the core. It is promising from the viewpoints of the reactivity characteristics and ULOF response.

    DOI

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    2
    被引用数
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  • Super light water reactors and super fast reactors: Supercritical-pressure light water cooled reactors

    Yoshiaki Oka, Seiichi Koshizuka, Yuki Ishiwatari, Akifumi Yamaji

    Super Light Water Reactors and Super Fast Reactors: Supercritical-Pressure Light Water Cooled Reactors     1 - 651  2010年  [査読有り]

     概要を見る

    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.

    DOI

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    154
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  • FEMAXI-6 Code Verification with MOX Fuels Irradiated in Halden Reactor

    Akifumi Yamaji, Motoe Suzuki, Tsutomu Okubo

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   46 ( 12 ) 1152 - 1161  2009年12月  [査読有り]

     概要を見る

    The advanced reactor concept innovative water reactor for flexible fuel cycle (FLWR) is being studied to achieve effective and flexible utilization of uranium and plutonium resources based on well-developed light water reactor (LWR) technology. In order to design and evaluate FLWR fuel rod behavior, uncertainties in FEMAXI-6 calculations and key models and parameters for predicting LWR MOX fuel rod behavior need to be evaluated. In this study, the test fuel data bases (TFDBs) obtained from the Halden reactor experiments (IFA-597.4 rod-10, rod-11, and IFA-514 rod-1) were used for the evaluations. The maximum discharge burnup was about 40 GWd/tMOX (IFA-514 rod-1). Based on the evaluation results, fission gas release (FGR), pellet densification, swelling, and relocation models were found to be particularly important. The FGR model has a relatively large uncertainty for predicting MOX fuel rod behavior. However, the uncertainties in the other models are within the range expected by the property variations of MOX fuels. Hence, the densification. swelling, and relocation models of FEMAXI-6 can be applied to MOX fuel analyses.

    DOI

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

    A. Yamaji, Y. Oka, Y. Ishiwatari, J. Liu, M. Suzuki

    ANNALS OF NUCLEAR ENERGY   33 ( 11-12 ) 984 - 993  2006年07月  [査読有り]

     概要を見る

    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 rationalizing the criteria for abnormal transients of the Super LWR is developed. The fuel rod integrities can be assured by preventing plastic strains on the cladding, preventing the cladding buckling collapse, and keeping the pellet centerline temperature below its melting point. The FEMAXI-6 fuel. analysis code is used to evaluate the fuel rod integrities in abnormal 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 abnormal transients. (c) 2006 Elsevier Ltd. All rights reserved.

    DOI

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    16
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  • Fuel and core design of Super Light Water Reactor with low leakage fuel loading pattern

    K Kamei, A Yamaji, Y Ishiwatari, Y Oka, J Liu

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   43 ( 2 ) 129 - 139  2006年02月  [査読有り]

     概要を見る

    An equilibrium core for the High Temperature Supercritical-pressure Light Water Reactor, now called the Super Light Water Reactor (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. This flow scheme is compatible with a low leakage fuel loading pattern (LLLP) in which 3(rd) cycle fuel assemblies are loaded in the core peripheral region. Stainless steel is used for fuel rod claddings and for structural materials. Watts correlations are used for predicting heat transfer ill the core. They take into account the improved heat transfer for downward flow. 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 and to reduce the thermal stress in water rod walls. Ail average coolant core outlet temperature of 500 degrees C is achieved. The effects of various heat transfer correlations oil the cladding surface temperature are evaluated.

    DOI

  • Development of statistical thermal design procedure to evaluate engineering uncertainty of Super LWR

    J Yang, Y Oka, J Liu, Y Ishiwatari, A Yamaji

    JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY   43 ( 1 ) 32 - 42  2006年01月  [査読有り]

     概要を見る

    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.

    DOI

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

    A Yamaji, K Kamei, Y Oka, S Koshizuka

    ANNALS OF NUCLEAR ENERGY   32 ( 7 ) 651 - 670  2005年05月  [査読有り]

     概要を見る

    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 &DEG; 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 &DEG; 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. &COPY; 2005 Elsevier Ltd. All rights reserved.

    DOI

    Scopus

    88
    被引用数
    (Scopus)
  • Three-dimensional core design of high temperature supercritical-pressure light water reactor with neutronic and thermal-hydraulic coupling

    Akifumi Yamaji, Yoshiaki Oka, Seiichi Koshizuka

    Journal of Nuclear Science and Technology   42 ( 1 ) 8 - 19  2005年  [査読有り]

     概要を見る

    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 39 kW/m, and a positive water density reactivity coefficient. © 2005 Taylor &amp
    Francis Group, Ltd.

    DOI

    Scopus

    87
    被引用数
    (Scopus)
  • Safety of Super LWR, (II)

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

    Journal of Nuclear Science and Technology   42 ( 11 ) 935 - 948  2005年  [査読有り]

     概要を見る

    This paper describes safety analysis of the high-temperature supercritical water-cooled thermal reactor with downward-flow water rods (called Super LWR) at supercritical pressure. Eleven transients and four accidents are chosen for the safety analysis considering types of abnormalities. The cladding temperature is taken as the important transient criterion instead of the heat flux ratio. The once-through cooling system and the downward-flow water rod system characterize safety of the Super LWR. “Loss of feedwater” is important because it is the same as “loss of reactor coolant flow” unlike BWR and PWR. However, the downward-flow water rods mitigate core heat-up before startup of the auxiliary feedwater system because they remove heat from the fuel channels by heat conduction and supply their water inventory to the fuel channels by volume expansion. During pressurization transients, the reactor power does not increase significantly unlike BWR due to no void collapse in single-phase flow and decrease in coolant density by flow stagnation in the once-through cooling system. All the transients and the accidents satisfy the criteria. Increases in the hottest cladding temperatures are about 50°C at transients and 250°C at accidents at maximum. The period of the high cladding temperature is very short at transients. © 2002 Taylor &amp
    Francis Group, Ltd.

    DOI

    Scopus

    52
    被引用数
    (Scopus)
  • Safety of super LWR, (I) safety system design

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

    Journal of Nuclear Science and Technology   42 ( 11 ) 927 - 934  2005年  [査読有り]

     概要を見る

    This paper describes design concept of safety system of the high-temperature supercritical pressure light water cooled reactor with downward-flow water rods (Super LWR). Since this reactor is once-through cooling system without water level and coolant circulation, the fundamental safety requirement is keeping core coolant flow rate while that of light water reactors (LWR) is keeping coolant inventory. “Coolant supply from cold-leg” and “coolant outlet at hot-leg” are needed for it. The advantage of the once-through cooling system is that reactor depressurization induces core coolant flow and cools the core. The downward-flow water rod system enhances this effect because the top dome and the water rods supply its water inventory to the core like an “in-vessel accumulator.” The safety system of the Super LWR is designed referring to those of LWR in consideration of its characteristics and safety principle. “Coolant supply” is kept by high-pressure auxiliary feedwater system and low-pressure core injection system. “Coolant outlet” is kept by safety relief valves and automatic depressurization system. The Super LWR is equipped with two independent shutdown systems: reactor scram system and standby liquid control system. The capacities and the actuation conditions determined in this study are to be used in safety analysis. © 2002 Taylor &amp
    Francis Group, Ltd.

    DOI

    Scopus

    40
    被引用数
    (Scopus)

▼全件表示

書籍等出版物

  • NUCLEAR POWER PLANT DESIGN AND ANALYSIS CODES

    ( 担当: 分担執筆,  担当範囲: Noval CFD methods, 18. Moving Particle Semi-implicit method)

    Elsevier  2021年

  • (連載講座)第4世代原子炉の開発動向 第3回 超臨界圧軽水冷却炉

    山路 哲史( 担当: 単著)

    2018年05月

  • Super Light Water Reactors and Super Fast Reactors

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

    Springer  2010年 ISBN: 9781441960351

  • 山路 哲史

    (解説)スーパー軽水炉, 炉心設計( 担当: 単著)

    日本原子力学会誌ATOMOΣ  2007年09月

講演・口頭発表等

  • MPS Method Simulation for Estimating Fuel Debris Distributions Under the Damaged Reactor Pressure Vessel of 1F Unit-2

    Yamato Bando, Akifumi Yamaji, Takuya Yamashita

    2023 International Conference on Environmental Remediation and Radioactive Waste Management (ICEM2023)  

    開催年月:
    2023年10月
     
     
  • Accident-Tolerant Fuel R&D Program in Japan

    S. Yamashita, A. Mohamad, I.Ioka, Y.Nemoto, T.Kawanishi, Y.Kaji, M.Osaka, N.Murakami, M.Owaki, M.Sasaki, K.Sakamoto, J.Matsunaga, A.Yamaji, H.Ohta

    2023 Water Reactor Fuel Performance Meeting (WRFPM2023)  

    開催年月:
    2023年06月
     
     
  • Preliminary Conceptual Development of the Super LWR Spectral Shift Core

    Akira Hirose, Takanari Fukuda, Akifumi Yamaji

    International Conference on Nuclear Engineering(ICONE30)  

    開催年月:
    2023年05月
     
     
  • Super FR Core Design Option with High Inlet Temperature for MA Transmutation

    T. Fukuda, A. Yamaji

    2021 International Congress on Advances in Nuclear Power Plants(ICAPP2021)  

    発表年月: 2021年10月

    開催年月:
    2021年10月
     
     
  • Development of MPS Method for Analysing Convection and Solidification of Multi-Component Corium in Severe Accident of a Light Water Reactor

    T. Fukuda, A. Yamaji, X. Li, J.F. Haquet, A. Boulin

    International Conference on Particle-Based Methods(PARTICLES 2021)  

    発表年月: 2021年10月

    開催年月:
    2021年10月
     
     
  • Design Study of SMR Class Super FR Core for In-Vessel Retention

    R. Sasaki, A. Yamaji, K. Uchimura

    28th International Conference on Nuclear Engineering (ICONE28)  

    発表年月: 2021年08月

    開催年月:
    2021年08月
     
     
  • Preliminary Core Design of The Solid Moderator Reactor for Investigation of The In-Depth Europa Ice Layer

    S. Fukizaki, A. Yamaji, T. Fukuda

    28th International Conference on Nuclear Engineering (ICONE28)  

    発表年月: 2021年08月

    開催年月:
    2021年08月
     
     
  • Analyses of wet and dry cavity strategies for BWR severe accident management with MELCOR-2.2

    A. Takashima, A. Yamaji, X. Li, D. Fujiwara, H. Shirai, T. Noju

    28th International Conference on Nuclear Engineering (ICONE28)  

    発表年月: 2021年08月

    開催年月:
    2021年08月
     
     
  • Preliminary Evaluation on the Relocation Phase of Ex-Vessel Unit-3

    Xin Li, Akifumi Yamaji, Masahiro Furuya, Ikken Sato, Hiroshi Madokoro, Yuji Ohishi

    28th International Conference on Nuclear Engineering (ICONE28)  

    発表年月: 2021年08月

    開催年月:
    2021年08月
     
     
  • LOCA analysis of super FR with RELAP/SCDAPSIM and FEMAXI-7

    S.Nakamoto, A.Yamaji, T.Okui

    The 10th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-10)  

    発表年月: 2021年03月

    開催年月:
    2021年03月
     
     
  • Core design of SUPER FR-MIX for improving neutronics and thermal-hydraulics performances

    T. Horiguchi, A.Yamaji, T.Fukuda, K.Uchimura

    The 10th International Symposium on Supercritical Water-Cooled Reactors (ISSCWR-10)  

    発表年月: 2021年03月

    開催年月:
    2021年03月
     
     
  • Numerical Study of Collision Behavior of Melt Droplets During Fuel-Coolant Interaction

    Panpan Wen, Gen Li, Jinchen Gao, Yupeng Li, Akifumi Yamaji, Junjie Yan

    International Conference on Nuclear Engineering (ICONE2020)  

    発表年月: 2020年08月

    開催年月:
    2020年08月
     
     
  • Preliminary Investigation on Improvement of FP Management during BWR Severe Accident with MELCOR-2.2

    ICONE2020 Anaheim, USA (Virtual, Online)  

    開催年月:
    2020年08月
     
     
  • Preliminary Power Transient Analysis of the Super FR with Axially Heterogeneous Core

    T. Okui, A. Yamaji

    PHYSOR2020  

    発表年月: 2020年04月

  • Preliminary Core Design Study of Small Supercritical Fast Reactor with Single Pass Cooling

    K. Uchimura, A. Yamaji

    PHYSOR2020   (ケンブリッジ) 

    発表年月: 2020年04月

  • Recent Progress in Development of Accident Tolerant FeCrAl-ODS Fuel Claddings for BWRs in Japan

    K. Sakamoto, Y. Miura, S. Ukai, A. Kimura, A. Yamaji, K. Kusagaya, S. Yamashita

    Top Fuel 2019   (シアトル)  米国原子力学会  

    発表年月: 2019年09月

  • Overview of Accident-Tolerant Fuel R&D Program in Japan

    S. Yamashita, I. Ioka, Y. Nemoto, T. Kawanishi, M. Kurata, Y. Kaji, T. Fukahori, T.Nozawa, D. Sato, N. Murakami, H. Sato, T. Kondo, K. Sakamoto, K. Kusagaya, S. Ukai, A. Kimura, A. Yamaji

    Top Fuel 2019   (シアトル)  米国原子力学会  

    発表年月: 2019年09月

  • Benchmark of Fuel Performance Codes for FeCrAl Cladding Behevior Analysis

    G. Pastore, K.A. Gamble, M. Cherubini, C. Giovedi, A. Yamaji, Y. Kaji, P. Van Uffelen, M.S. Veshchunov

    Top Fuel 2019   (シアトル)  米国原子力学会  

    発表年月: 2019年09月

  • Analysis of FeCrAl-ODS Cladded Fuel Performance during BWR Power Ramp with FEMAXI-7

    Yoshihiro Fujiwara, Akifumi Yamaji, Shigeharu Ukai, Kan Sakamoto, Shinichiro Yamashita

    Top Fuel 2019   (シアトル)  米国原子力学会  

    発表年月: 2019年09月

     概要を見る

    In Japan, as one of the candidates for the accident tolerant fuel cladding, Oxide Dispersion Strengthened (ODS) type of FeCrAl (FeCrAl-ODS) steel is being developed for BWR. Compared with Zircaloy (Zry), it has higher thermal expansion rate and is expected to have lower irradiation creep rate and higher yield stress. The purpose of this study is to perform FEMAXI-7 analyses of FeCrAl-ODS cladded fuel rod under BWR power ramp test conditions to reveal influence of such cladding properties on PCMI behavior. Sensitivity analyses have been conducted with different cladding creep and plastic deformation models / parameters. The evaluated range of pellet temperature and PCMI contact pressure of FeCrAlODS cladded fuel rod is comparable or less than those with Zry cladded fuel rod. Hence, the results indicate that the current evaluations with FEMAXI-7 is well within the experience of the code with typical light water reactor fuel pellet modeling with respect to evaluations of stress-strain interactions between pellet and cladding. More investigations may be necessary as the irradiation data are gained with FeCrAl-ODS and incorporated into mechanical models of the cladding.

  • Insights on in-vessel core degradation behavior from sensitivity analysis of Fukushima Daiichi nuclear power plant unit3 by MELCOR

    Xin Li, Ikken Sato, Akifumi Yamaji

    福島廃炉研究国際会議2019   (福島県)  日本機械学会  

    発表年月: 2019年05月

     概要を見る

    In order to highlight the knowledge gaps and uncertainties existing in understanding of the severe accident scenarios and consequences, this study focuses on identifying the modeling uncertainties and addressing the sensitivity parameters in Fukushima NPP Unit 3 with MELCOR code. A more detailed Control Volume (CV) division model of the reactor core region has been developed to better simulate the thermal-hydraulic behavior of liquid water and vapor, which is considered to be crucial in simulating the core uncovery and degradation process. The boundary conditions such as the water injection rates by the Reactor Core Isolation Cooling (RCIC) system, the High Pressure Core Injection (HPCI) system and Alternative Water Injection (AWI) to the reactor core were determined based on the available reactor water level and pressure measurement data. The current study suggested that the local vapor heatup behavior could influence the core melting and relocation behavior, which can lead to different core degradation scenarios. With the current modeling assumptions in MELCOR, the best estimate conditions for RPV pressure history of Unit 3 suggested that 6 Safety Relief Valves (SRVs) could have remained open when the major core slumping is assumed to occur at ca. 45:20 h (ca. 12:00, March 13) with 50 to 80 tons of water inventory in the lower plenum. The current analysis also suggested that the efficiency of the AWI to the reactor core could have been only 15% as of reported by TEPCO estimated amount of water discharged by fire engines with the current modeling conditions if debris dryout was assumed to have occurred at around ca. 54:00 h (20:40 h, March 13th). As for lower head failure, there is still large uncertainty in predicting lower head failure time with Larson-Miller creep rupture model in the current MELCOR modeling.

  • Preliminary sensitivity analysis for estimating core thermal energy at the time of core slumping of Fukushima Daiichi unit3 with MELCOR-2.2

    Mariko Regalado, Akifumi Yamaji

    福島廃炉研究国際会議2019   (福島県)  日本機械学会  

    発表年月: 2019年05月

     概要を見る

    To provide supportive information to understand the current debris status in Fukushima Daiichi Nuclear Power Plant Unit-3, sensitivity analyses have been carried out with MELCOR-2.2 with different control volume divisions of the core region in the CVH package. The particular focus of the analyses is the estimated debris thermal energy up to and at the time of the core slumping event, which may be the key to determine the following debris cooling and failure mechanism of the lower head of the reactor pressure vessel (RPV). When the core region is treated by a single CVH control volume in the axial direction, the gas temperature and compositions (i.e., steam and hydrogen) are assumed to be uniform. This assumption leads to rapid decrease of the RPV water level after the loss of HPCI water injection. The water level decreases significantly below the bottom of the active fuel (BAF) level, whereas in the actual accident, the water level may have been maintained just below BAF. By dividing the corresponding CVH control volume to 5 – 10 divisions, local gas composition changes due to steam-zirconium reaction and the consequent local heat-up of the core can be captured and the water level decreased more gradually. As the result, the estimated core thermal energy has increased. However, the simulated core thermal energy is still decreased before the core slumping as part of the energy is transferred to the suppression chamber by the steam from the lower plenum. The authors discuss that such discrepancy may be due to the uncertainties with macroscopic gas permeability of the damaged core. In this study, such uncertainty is qualitatively considered and discussed for estimating the core thermal energy up to and at the time of the core slumping, based on the MELCOR analysis results. With such considerations, the authors expect significant melting of the fuel debris at the time of the major core slumping

  • Estimation of thermal status of the fuel debris at the time of core slumping of 1F2 with MELCOR-2.2

    Kodai Wadayama, Akifumi Yamaji

    福島廃炉研究国際会議2019   (福島県)  日本機械学会  

    発表年月: 2019年05月

     概要を見る

    For decommissioning Fukushima Daiichi Nuclear Power Plants, estimating the current debris status is important. One way to support such attempt is to estimate thermal status of the fuel debris at the time of core slumping, when major part of the degraded core may have relocated to the bottom of the reactor pressure vessel (RPV). The characteristic three RPV pressure peaks recorded for the unit-2 (1F2) have been discussed with different integrated analysis codes as signs of such core degradation, but so far, evaluations regarding the fuel debris thermal status up to and at the time of the core slumping have been limited. This study aims to reveal potential issues for such evaluations with MELCOR-2.2. The key plant reference data are the histories of recorded RPV pressure, primary containment vessel (PCV) pressure, and the RPV water level. To estimate the fuel debris thermal status, the following evaluations were considered in details: Natural circulation of gas (steam and hydrogen) in the core, oxidation of fuel cladding, heat transport from RPV to suppression chamber (S/C) and the consequent changes to the RPV and PCV pressures and RPV water level. The following two cases were investigated: (1)1CV case: The core is represented by a single control volume, which assumes uniform gas composition and temperature in the axial direction; (2) 10CV case: The core control volume is divided into five volumes. In 10CV case, the core oxidation reaction and heat-up were observed locally and metallic melt pool is formed in the core region after the RPV water level recovered and reached the bottom of the active fuel (BAF). However, UO2 was almost unmolten in all cases. The authors discuss that macroscopic gas permeability of the damaged core may have been overestimated in the current MELCOR simulation, but such uncertainty may not significantly change the overall view of the core thermal status as largely unmolten at the time of the core slumping.

  • Ablation Analysis with MPS for Proposing Ex-Vessel Corium Spreading Management in Light Water Reactors

    M. Katta, A. Yamaji, G. Duan, M. Furuya

    第27回原子力工学国際会議   (つくば市)  日本機械学会  

    発表年月: 2019年05月

     概要を見る

    In a postulated sever accident of a light water reactor (LWR), molten core debris (corium) may breach the reactor pressure vessel and be released to the ex-vessel containment floor. A core catcher manages ex-vessel corium cooling by uniformly spreading the corium into a large space, but it requires a dedicated plant design. In contrast, corium shields have been back-fitted to some boiling water reactors to prevent excessive amount of corium to flow into sump pits, where effective corium cooling may be difficult. However, corium shields can only block the corium flow and cannot contribute to uniform spreading of the ex-vessel corium. This study proposes a preliminary concept of “Debris Spreading Floor”, which can be applied to any types of reactor plants including back-fitting to the existing plants. More specifically, the existing containment floor is overlaid with sacrificial material and refractory material is placed around sump pits. It is intended to allow the original function of sump pits to collect leaking water under normal, abnormal transient and design basis accident conditions. However, under postulated severe accident condition, spreading of exvessel corium is promoted by ablating itself with the hot corium and guiding corium spreading away from sump pits. To develop the concept, mechanistic analysis of corium spreading, which can consider influence of substrate ablation, is needed. The Moving Particle Semi-implicit (MPS) method is a Lagrangian particle method and thus suitable for mechanistic simulation of free-surface spreading flow involving solid / liquid phase change and interactions. In this research, effect of the proposed concept is firstly presented with the MPS simulations. Preliminary simulations in 2D show that, amount of corium flowing into sump pits is reduced by the concept. Secondly, validity of the MPS simulations is quantitatively discussed by simulating the experiments with simulant. The experiment was carried out by Central Research Institute of Electric Power Industry (CRIEPI) by pouring liquid Pb-Bi onto a Pb-Bi block so that the inflow liquid spreads on the block surface while it also ablates the block. Sensitivity analyses have been carried out with different initial conditions, calculation resolutions, subscale models and parameters of the MPS simulations to identify the key models and parameters for quantitative prediction of the melt / substrate interactions.

  • SENSITIVITY ANALYSIS OF CORE SLUMPING AND ALTERNATIVE WATER INJECTION IN FUKUSHIMA NUCLEAR POWER PLANT UNIT 3ACCIDENT

    Xin Li, Ikken Sato, Akifumi Yamaji

    European Review Meeting on Severe Accident Research 2019   (プラハ)  UJV Rez  

    発表年月: 2019年03月

  • Overview of the Heterogeneous Core Design Studies of Super Fast Reactor at Waseda University

    A. Yamaji, S. Noda, T. Fukuda, Sukarman

    International Symposium on SCWRs  

    開催年月:
    2019年03月
     
     
  • 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   (パリ)  Generation IV Forum(GIF)  

    発表年月: 2018年10月

     概要を見る

    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.

  • A Novel Approach for Crust Behaviors in Corium Spreading Based on Multiphase MPS Method

    G.Duan, A.Yamaji, S.Koshizuka

    12th International Topical Meeting on Reactor Thermal-Hydraulics, Operation   (Qingdao)  CNS  

    発表年月: 2018年10月

  • Numerical Investigation on VULCANO VE-U7 Corium Spreading Over Ceramic/Concrete Substrates with MPS Method

    Jubaidah, G.Duan, A.Yamaji

    12th International Topical Meeting on Reactor Thermal-Hydraulics, Operation   (Qingdao)  CNS  

    発表年月: 2018年10月

  • Improvement of Numerical Analysis of Molten Core-Concrete Interaction with Top Water Injection by MPS Method

    J.Murata, X.Li, A.Yamaji

    12th International Topical Meeting on Reactor Thermal-Hydraulics, Operation   (Qingdao)  CNS  

    発表年月: 2018年10月

  • Numerical Investigation of the Stop-and-Go Mechanism in FARO L26S Spreading Experiment by MPS Method

    K.Uchida, G.Duan, A.Yamaji

    12th International Topical Meeting on Reactor Thermal-Hydraulics, Operation, and Safety   (Qingdao)  CNS  

    発表年月: 2018年10月

  • Analysis of Irradiation Matrix for the Japanese FeCrAl-ODS Test Fuel Rods Irradiations at the Halden Reactor using FEMAXI-7 code

    N.Susuki, A.Yamaji, K.Kusagaya, K.Sakamoto, S.Yamashita

    TopFuel 2018   (プラハ)  European Nuclear Society  

    発表年月: 2018年09月

  • Three-Dimensional Numerical Study on Pool Stratification Behavior in Molten Corium-Concrete Interaction (MCCI) With MPS Method

    Xin Li, Ikken Sato, Akifumi Yamaji, Guangtao Duan

    26th International Conference on Nuclear Engineering (ICONE-26)  

    開催年月:
    2018年07月
     
     
  • 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 Engineering   (ロンドン) 

    発表年月: 2018年07月

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

    Yudai Tasaki, Akifumi Yamaj

    2018 International Congress on Advances in Nuclear Power Plants   (シャーロット、ノースカロライナ) 

    発表年月: 2018年04月

  • 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 Meeting   (済州道) 

    発表年月: 2017年09月

  • 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 Meeting   (済州島) 

    発表年月: 2017年09月

  • A core design of innovative breeder BWR

    Rui G, Yamaji A, Cai Y, Peng X

    25th International Conference on Nuclear Engineering (ICONE25)  

    開催年月:
    2017年07月
     
     
  • 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 Engineering   (上海) 

    発表年月: 2017年05月

  • CONCEPTUAL CORE DESIGN OF BREEDING BWR

    Rui Guo, Akifumi Yamaji

    25th International Conference on Nuclear Engineering   (上海) 

    発表年月: 2017年05月

  • 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 Plants   (福井、京都) 

    発表年月: 2017年04月

  • 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 Plants   (福井、京都) 

    発表年月: 2017年04月

  • 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 Plants   (福井、京都) 

    発表年月: 2017年04月

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

    Y.Goto, A.Yamaji

    (ウィーン) 

    発表年月: 2017年02月

  • 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 Safety   (慶州) 

    発表年月: 2016年10月

  • 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 Safety   (慶 州) 

    発表年月: 2016年10月

  • Sensitivity Study of 1F1 Type Accident by MELCOR code

    Kenta Saitoa, Akifumi Yamaji

    Transaction of ANS Winter meeting 2015   (ワシントン) 

    発表年月: 2015年11月

  • 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 2015   (シカゴ) 

    発表年月: 2015年09月

  • 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,   (シカゴ) 

    発表年月: 2015年09月

     概要を見る

    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.

  • 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 2014   (京都) 

    発表年月: 2014年09月

  • 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 2014   (京都) 

    発表年月: 2014年09月

  • 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 2014   (京都) 

    発表年月: 2014年09月

  • 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 2014   (京都) 

    発表年月: 2014年09月

  • Development of IDAT: IRPhE database and analysis tool

    Hill I, Gulliford J, Soppera N, Bossant M, Yamaji A

    2012 ANS Annual Meeting and Embedded Topical Meeting: Nuclear Fuels and Structural Materials for the Next Generation Nuclear Reactors, NFSM 2012  

    開催年月:
    2012年06月
     
     
  • 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 2009   (東京) 

    発表年月: 2009年05月

  • 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 2009   (東京) 

    発表年月: 2009年05月

     概要を見る

    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

  • 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 2009   (東京) 

    発表年月: 2009年05月

     概要を見る

    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

  • スーパー軽水炉(超臨界圧軽水炉)の炉心設計

    山路哲史  [招待有り]

    日本原子力学会  

    発表年月: 2007年09月

  • スーパー軽水炉の炉心・燃料設計

    山路哲史  [招待有り]

    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  

    発表年月: 2006年09月

  • スーパー軽水炉の炉心・燃料設計

    山路哲史  [招待有り]

    革新的水冷却炉研究会(第9回)  

    発表年月: 2006年03月

  • 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 Sustainability   (筑波) 

    発表年月: 2005年10月

     概要を見る

    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.

  • 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 Sustainability   (筑波) 

    発表年月: 2005年10月

     概要を見る

    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.

  • 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 Sustainability   (筑波) 

    発表年月: 2005年10月

  • 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 2005   (ソウル) 

    発表年月: 2005年05月

     概要を見る

    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

  • 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 2005   (ソウル) 

    発表年月: 2005年05月

     概要を見る

    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.

  • 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 Plants   (ペンシルバニア州) 

    発表年月: 2004年06月

     概要を見る

    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.

  • 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 Energy   (ロサンゼルス) 

    発表年月: 2003年11月

     概要を見る

    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.

  • 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 Energy   (ロサンゼルス) 

    発表年月: 2003年11月

     概要を見る

    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

  • 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 Plants   (京都) 

    発表年月: 2003年09月

  • 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 Plants   (京都) 

    発表年月: 2003年09月

  • 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 Plants   (京都) 

    発表年月: 2003年09月

  • 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 Plants   (コルドバ) 

    発表年月: 2003年05月

  • 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 Plants   (フロリダ) 

    発表年月: 2002年06月

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

    Akifumi YAMAJI, Yoshiaki OKA, Seiichi KOSHIZUKA

    2001 ANS/HPS Student Conference   (テキサス) 

    発表年月: 2001年04月

▼全件表示

共同研究・競争的資金等の研究課題

  • 社会に受け入れられる事故復旧性スーパー高速炉概念の研究

    研究期間:

    2020年04月
    -
    2024年03月
     

  • Multi-Physicsモデリングによる福島2・3号機ペデスタル燃料デブリ深さ方向の性状同定

    日本原子力研究開発機構  CLADS英知を結集した原子力科学技術・人材育成推進事業

    研究期間:

    2019年10月
    -
    2022年03月
     

  • 沸騰水型軽水炉の過酷事故進展解析モデルの研究

    株式会社テプコシステムズ  共同研究

    研究期間:

    2017年06月
    -
    2021年03月
     

    山路哲史

  • コリウム広がり解析のMPS-THEMAクロスウォーク

    日本学術振興会  二国間交流事業

    研究期間:

    2018年04月
    -
    2019年12月
     

  • Multi-physicsモデリングによるEx-Vessel溶融物挙動理解の深化

    文部科学省  英知を結集した原子力科学技術・人材育成推進事業

    研究期間:

    2016年10月
    -
    2019年03月
     

  • 事故耐性燃料棒のふるまいと溶融時の挙動解析研究

    日本学術振興会  科学研究費助成事業

    研究期間:

    2015年04月
    -
    2018年03月
     

    山路 哲史

     概要を見る

    原子炉過酷事故の防止・抑制を目標に、燃料被覆管に炭化ケイ素(SiC)等を用いる事故耐性燃料(ATF)が検討されている。本研究では沸騰水型軽水炉(BWR)にこれら候補ATFを導入した場合の通常運転時、異常な過渡変化時、溶融事故時のふるまいを明らかにし、ATFが取りえる主要な状態(通常時、異常な過渡変化時、溶融事故時)のマルチスケールな挙動理解からATF導入のための要求や課題を明らかにした

  • 高速・熱中性子結合炉心の炉物理的研究

    日本学術振興会  科学研究費助成事業

    研究期間:

    2010年04月
    -
    2013年03月
     

    岡 芳明, 石渡 祐樹, 山路 哲史

     概要を見る

    高速・熱中性子結合炉心は高速炉心に熱中性子を発生する領域が分散配置された高速炉心である。高速・熱中性子結合炉心の核的な特性を原子炉物理学の観点から明らかにするとともに水冷却高速炉の高増殖性について研究した。まず、高速・熱中性子結合炉心は炉心の非均質性が高いため、隣接する集合体間の核的非均質性の影響をセル均質化マクロ断面積に反映させた炉心解析法を開発した。次に、燃料棒を密に束ねた新燃料集合体を考案し、軽水冷却高速炉で、複合システム増倍時間が40 年程度にできることを核的に世界で初めて明らかにした。熱水力設計を含む炉の成立性は今後の課題である

▼全件表示

Misc

  • Multi-Physicsモデリングによる福島2・3号機ペデスタル燃料デブリ深さ方向の性状同定(9)福島第一原子力発電所3号機デブリのペデスタル移行時に着目したプラントデータの分析

    佐藤一憲, 山路哲史, 古谷正裕, 大石佑治, LI Xin, 間所寛, 深井尋史

    日本原子力学会秋の大会予稿集(CD-ROM)   2021  2021年

    J-GLOBAL

  • Multi-Physicsモデリングによる福島2・3号機ペデスタル燃料デブリ深さ方向の性状同定(6)全体概要と3号機ペデスタルのデブリ臨界性の試評価

    山路哲史, 岸本和真, LI Xin, 古谷正裕, 佐藤一憲, 間所寛, 大石佑治

    日本原子力学会春の年会予稿集(CD-ROM)   2021  2021年

    J-GLOBAL

  • Multi-Physicsモデリングによる福島2・3号機ペデスタル燃料デブリ深さ方向の性状同定(1)ねらいと全体計画

    山路哲史, 古谷正裕, 大石佑治, 佐藤一憲, 深井尋史, LI Xin, 間所寛

    日本原子力学会春の年会予稿集(CD-ROM)   2020  2020年

    J-GLOBAL

  • Multi-Physicsモデリングによる福島2・3号機ペデスタル燃料デブリ深さ方向の性状同定(2)2,3号機燃料デブリ状態に係る論点

    佐藤一憲, 山路哲史, 古谷正裕, 大石佑治, LI Xin, 間所寛, 深井尋史

    日本原子力学会春の年会予稿集(CD-ROM)   2020  2020年

    J-GLOBAL

  • Multi-Physicsモデリングによる福島2・3号機ペデスタル燃料デブリ深さ方向の性状同定(3)ねらいと全体計画及び一年目の進捗

    山路哲史, 古谷正裕, 大石佑治, 佐藤一憲, 深井尋史, LI Xin, 間所寛

    日本原子力学会秋の大会予稿集(CD-ROM)   2020  2020年

    J-GLOBAL

  • Multi-Physicsモデリングによる福島2・3号機ペデスタル燃料デブリ深さ方向の性状同定(4)2号機RPVバウンダリー破損モードの検討

    佐藤一憲, 山路哲史, 古谷正裕, 大石佑治, LI Xin, 間所寛, 深井尋史

    日本原子力学会秋の大会予稿集(CD-ROM)   2020  2020年

    J-GLOBAL

  • Multi-physicsモデリングによるEx-Vessel溶融物挙動理解の深化(6)全体概要とMPS法によるspreading解析の高度化(3)

    山路哲史, 古谷正裕, 大石佑治, JUBAIDAH, DUAN Guangtao

    日本原子力学会春の年会予稿集(CD-ROM)   2019  2019年

    J-GLOBAL

  • 高温粘性流体の三次元流動とペデスタル床のアブレーション

    古谷正裕, 古谷正裕, 山路哲史, 大石佑治

    混相流シンポジウム講演論文集(Web)   2019  2019年

    J-GLOBAL

  • 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

    山路哲史, 古谷正裕, 大石佑治, DUAN Guangtao

    日本原子力学会春の年会予稿集(CD-ROM)   2018  2018年

    J-GLOBAL

  • Multi-physicsモデリングによるEx-Vessel溶融物挙動理解の深化(4)全体概要とMPS法によるSpreading解析の高度化(2)

    山路哲史, 古谷正裕, 大石佑治, DUAN Guangtao

    日本原子力学会秋の大会予稿集(CD-ROM)   2018  2018年

    J-GLOBAL

  • Multi-physicsモデリングによるEx-Vessel溶融物挙動理解の深化(5)ペデスタル床複雑構造に拡がる溶融物の三次元流動

    古谷正裕, 山路哲史, 大石佑治

    日本原子力学会秋の大会予稿集(CD-ROM)   2018  2018年

    J-GLOBAL

  • 複雑流路内の三次元流動観察と数値混相流体力学解析

    古谷正裕, 山路哲史, 大石佑治

    日本機械学会年次大会講演論文集(CD-ROM)   2018  2018年

    J-GLOBAL

  • Multi-physicsモデリングによるEx-Vessel溶融物挙動理解の深化(1)全体計画

    山路哲史, 古谷正裕, 大石佑治, DUAN Guangtao

    日本原子力学会秋の大会予稿集(CD-ROM)   2017  2017年

    J-GLOBAL

  • スーパー軽水炉 (超臨界圧軽水炉) の炉心設計

    山路 哲史

    日本原子力学会誌 = Journal of the Atomic Energy Society of Japan   49 ( 9 ) 607 - 612  2007年09月

    CiNii

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他学部・他研究科等兼任情報

  • 理工学術院   先進理工学部

学内研究所・附属機関兼任歴

  • 2022年
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    2024年

    理工学術院総合研究所   兼任研究員

  • 2022年
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    2024年

    カーボンニュートラル社会研究教育センター   兼任センター員

特定課題制度(学内資金)

  • MPS法-VOF法による先行流下物凝固後の後続溶融物挙動解析Crosswalk

    2023年   吉田 啓之, 山下 晋

     概要を見る

    軽水炉の過酷事故では金属溶融物が炉内の燃料集合体中等の構造物群を流下し、先行流下物の凝固が後続溶融物を他の経路に選択的に流下させ、炉心損傷進展や事故後のデブリ堆積分布に影響する。このような挙動を解明するために実施された従来の実験は試験部への溶融物の流入境界の不確かさがその後の溶融物挙動に及ぼす影響が大きかったため解析モデルの妥当性確認が困難であった。申請者らはこのような課題を解決するために2022年度に日本原子力研究開発機構(JAEA)と共同で試験部への流入境界の不確かさを低減した実験を提案、実施した。本申請研究では、その結果得られた実験データを活用し、JAEAの有するオイラー法に基づくVOF法解析と申請者らが開発するラグランジュ法に基づくMPS法解析の相互比較(Crosswalk)を実施し、先行流下物の凝固が後続溶融物挙動に及ぼす影響を明らかにした。オイラー法による解析では気相による冷却効果を小さくすると実験結果と同様に、中央部では溶融物が薄く、壁面近傍部で厚い分布となった。そこで、MPS法による解析においても気相による冷却効果は無視した。流動先端の接触角を仮想的に大きくしたケースや、流動先端部と床の接触熱抵抗を大きくしたケースでは、流動先端部と床の粘性相互作用が後続バルク流体のそれに比べて小さくなり、実験結果と同様な流動挙動を再現できた。すなわち、前者では流動先端部と床との間に微小なギャップが生じ、流動先端部の流動に対する床の抵抗が低下した。また、同様に、流動先端部と床の接触熱抵抗を大きくしたケースでは、流動先端部の溶融物の粘性が低く、床による流動抵抗が低下した。以上の実験、VOF法による解析、MPS法による解析から、低融点合金の流下挙動の模擬には、溶融物と床との伝熱が及ぼす影響が支配的であることが分かり、そのモデリングが重要であることを明らかにした。

  • 粒子法による溶融物流下時の凝固に伴う非閉塞流路への選択的流下挙動の解明

    2022年   吉田 啓之, 山下 晋

     概要を見る

    軽水炉の過酷事故では金属溶融物などの先行流下物の凝固に伴う流路閉塞が後続流下物を他の流路に選択的に流下させ、事故進展に影響する可能性がある。流体を計算点(粒子)でラグランジュ的に離散化し、固液界面の追跡が容易なMPS法を用いた先行研究では、計算精度と安定性を損なわずに凝固粒子の計算負荷を軽減する改良MPS法が提案されたが、実験との比較による妥当性は示されていない。本研究では上述の7本ピンバンドル実験を改良MPS法で解析し、溶融物の選択的流下挙動を解析するための課題を明らかにした。さらに、実験に伴う不確かさを低減する新たな溶融物挙動解析ベンチマーク用実験を日本原子力研究開発機構と連携し提案した。

  • 固体減速炉による深宇宙衛星氷層沈降探査機の概念設計

    2021年  

     概要を見る

    NASA等による月面利用の新規プロジェクトの次の展開には、太陽光の届かない深宇宙探査の新たな展開も考えられ、深宇宙探査機の電源には原子炉システムを用いることが検討されている。本研究では、日本原子力研究開発機構等による先行研究で検討された小型原子炉を木星の衛星エウロパに投下し、その熱出力で氷層を溶かしながら自重で沈降する探査炉に継続利用するシナリオを構築した。そのために、先行研究の炉心設計を小型原子炉の炉心核計算により最適化した。さらに、MPS法による炉心の氷層溶融沈降解析により、炉心熱設計の成立性を示した。

  • 改良ステンレス被覆燃料のExtended LOCA時ふるまい解析

    2020年  

     概要を見る

    燃料ふるまい解析コードFEMAXI-7の改良と解析により、改良ステンレス鋼被覆管を用いた燃料の冷却喪失が持続するExtended LOCA時ふるまいを明らかにし、改良ステンレス鋼の高温クリープ挙動特性データ取得の必要性や必要とされるデータ範囲を明らかにした。9×9型BWR燃料を例に解析した結果、Zry被覆管の予測破裂温度は約1200K, 改良ステンレス鋼被覆管は同約1400Kから1470Kとなった。改良ステンレス鋼のZryに対する優位性を示すためには約1200Kから1470Kの歪量を測定した実験データが必要であることが明らかになった。

  • 軸方向非均質スーパー高速炉のLOCA特性の解明

    2019年  

     概要を見る

    第四世代の原子炉の一つであるスーパー高速炉は原子炉冷却材に超臨界水を用いる。その冷却材喪失事故(LOCA)時には高温高圧の超臨界水単相流冷却から減圧後の水蒸気二層流冷却及び過熱蒸気単相流冷却までの広い範囲で燃料被覆管温度を評価する必要がある。本研究では、軸方向に複数のMOX燃料層とブランケット燃料層から構成するスーパー高速炉のLOCA時炉心冷却特性を一点近似動特性モデルと一次元熱水力モデルに基づくRELAP/SCDAPSIMコードを用いて明らかにした。従来と異なり、上部MOX層-中間ブランケット層境界で大きな温度勾配が生じ、スクラム時間遅れの増大に伴い温度勾配も増大することが明らかになった。

  • 新型軽水炉燃料の冷却材喪失事故時ふるまい解析

    2018年  

     概要を見る

    福島事故後、原子炉事故時の安全性を向上する新型燃料の研究開発が急務となっており、本研究では、軽水炉燃料ふるまい解析コードFEMAXI-7の改良と解析により、改良ステンレス鋼(FeCrAl-ODS)被覆燃料と従来燃料の原子炉冷却材喪失事故(LOCA)時ふるまいの違いを明らかにした。原子力国際機関(IAEA)が、各国の燃料ふるまい解析コードを用いたFeCrAl-ODS被覆燃料のLOCA時のふるまい解析のベンチマークを実施し、研究者(早大)は日本を代表して同国際ベンチマークにも参加し、研究成果を共有した。通常運転時とLOCA時の両ケースの解析に成功したのは参加した5カ国のうち、米国と日本(早大)の2カ国のみで、日本の優れた研究力の発信にも貢献した。

  • 機能的デブリ分散床の基礎概念研究

    2017年  

     概要を見る

    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。

  • 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  

     概要を見る

    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。

  • 高温スーパー高速炉の燃料棒ふるまい設計研究

    2015年  

     概要を見る

    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。

  • スーパー高速炉の炉心高温化の研究

    2014年  

     概要を見る

    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。

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