数値流体力学の原子炉熱流動への応用に関する国際シンポジウムの

数値流体力学の原子炉熱流動への応用に関する国際シンポジウムの開催案内
㈱原子力安全システム研究所
原子力安全システム研究所では標記シンポジウムを下記の要領で開催いたします。
（別添プログラムを参照ください）
お招きした講演者は世界的にも著名な原子炉熱流動の専門家であり、休憩時間を含め
て相互交流の場を提供できると期待しておりますので、是非ご参加いただくようご案内
いたします。
開催日：
会場：
主催：
共催：
後援：
協賛：
2015 年 5 月 22 日（金）13:10-17:30
（株）原子力安全システム研究所３階ロードマーシャルメモリアルホール
（株）原子力安全システム研究所（INSS）
福井大学附属国際原子力工学研究所
福井県、美浜町、日本原子力研究開発機構、若狭湾エネルギー研究センター
日本機械学会、日本機械学会動力エネルギーシステム部門、日本機械学会
北陸信越支部、日本混相流学会、日本原子力学会熱流動部会
参加費：無料
使用言語：英語（通訳なし）
申込締切：5 月 18 日(月)
無料バス：敦賀駅発 12:40 (INSS 行き)
INSS 発 17:40 (敦賀駅行き)
参加ご希望の方は、予めメールもしくは FAX にてお申し込み下さい。
以上
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[送り状]
（株）原子力安全システム研究所技術システム研究所
歌野原陽一宛て (E-mail: [email protected]) FAX:0770-37-2009：申込締切 5 月 18 日(月)
数値流体力学の原子炉熱流動への応用に関する国際シンポジウムへの参加申込
2015 年 5 月 22 日（金）に INSS で開催される「数値流体力学の原子炉熱流動への応用に関する国
際シンポジウム」に参加希望します。
氏名：
所属：
E-mail：
無料バス乗車：
有(出発時間：行
, 帰り
)
無
問合せ先：（株）原子力安全システム研究所技術システム研究所
歌野原陽一
E-mail: [email protected], Tel: 050-7105-0099
プログラム
時
間
内容
12:40
敦賀駅発・INSS 行き
12:50-13:10
受
13:10-13:20
開会挨拶
13:20-13:35
INSS 技術システム研究所の概要紹介
13:35-14:00
INSS の研究紹介 T 字合流管での熱疲労評価法を改善するための熱応力特性
三好弘二准主任研究員（INSS）
14:00-15:00
招待講演１原子炉安全解析への CFD の応用
Dr. Thomas Höhne, Helmholtz-Zentrum Dresden-Rossendorf (ドイツ)
休憩
15:00-15:30
15:30-16:20
16:20-17:20
17:20-17:30
17:40
付
三島嘉一郎所長（INSS）
福井大学附属国際原子力工学研究所の研究紹介
機器での熱流動問題への CFD の応用
望月弘保特命教授
軽水炉安全問題への CFD の応用
渡辺正教授
招待講演２原子力システム熱流体挙動に対する数値シミュレーションの妥当性確認
Professor Yassin A. Hassan, Texas A&M University (米国)
閉
会
INSS 発・敦賀駅行き
Summary and Brief Biography
Invited Lecture 1: Application of computational fluid dynamics in nuclear reactor safety
analysis
Dr. Thomas Höhne, Helmholtz-Zentrum Dresden-Rossendorf (HZDR)
Dr. Höhne studied Thermal Engineering at the Department of Mechanical Engineering at
Technical University of Dresden/Germany. He graduated 1995 in collaboration with ABB Power
Generation Switzerland. In 1996 he worked as a CFD expert at the ABB Research Centre in
Switzerland on the numerical simulation of the flow in low pressure steam turbines. Since 1997
he is working at the CFD department in the Institute of Fluid Dynamics at HZDR in the field of
the application of computational fluid dynamics in nuclear reactor safety analysis and obtained
in this field his PhD in 2003. His newer activities comprise CFD-analyses in buoyancy driven
natural circulation phenomena and stratified horizontal two-phase flow processes. Nowadays he
is senior scientist and project leader. He wrote over 180 publications in the field of single and
multiphase CFD, fluid dynamics applications in NRS and in the design and performance of
experimental facilities.
Abstract: The last decade has seen an increasing use of three-dimensional CFD codes to predict
steady state and transient flows in nuclear reactors because a number of important phenomena
such as pressurized thermal shocks, coolant mixing, and thermal striping cannot be predicted by
traditional one-dimensional system codes with the required accuracy and spatial resolution. CFD
codes contain models for simulating turbulence, heat transfer, multi-phase flows, and chemical
reactions. For this reason the long-term objective of the activities of the Helmholtz-Zentrum
Dresden-Rossendorf Germany (HZDR) R&D program lies in the development of theoretical
models for basic phenomena of transient, three-dimensional single and multiphase systems. Such
models must be validated before they can be used with sufficient confidence in nuclear reactor
safety (NRS) applications. The necessary validation is performed by comparing model results
against measured data. However, in order to obtain a reliable model assessment, CFD
simulations for validation purposes must satisfy strict quality criteria given in the Best Practice
Guidelines (BPG). The following topical issues which are related to PWR, where CFD
calculations have been performed, will be briefly discussed: Coolant mixing of the primary
circuit, horizontal stratified flow phenomena in the hot leg and fibre material transport in a core
under loss of coolant conditions.
Invited Lecture 2: Validation in Computational Simulation of Nuclear System Thermal
Fluids Behavior
Professor Yassin A. Hassan, Texas A&M University
Dr. Yassin Hassan is Head of the Department of Nuclear Engineering, Sallie and Don Davis ’61
Professor of Engineering and Professor of the Department of Mechanical Engineering at Texas
A&M University. Prior to joining Texas A&M in September 1986, he worked for seven years at
Nuclear Power Division, Babcock & Wilcox Company, Lynchburg, Virginia. His research is in
computational and experimental thermal hydraulics, reactor safety, laser-based flow
visualization and diagnostic imaging techniques, system modeling, multiphase flow, transient
and accident analyses and advanced nuclear reactors. He received his master's and Ph.D.
degrees from the University of Illinois at Urbana-Champaign in nuclear engineering. He also
has a master's degree in mechanical engineering from the University of Virginia.
Dr. Hassan's professional recognitions include selection as a fellow of the American Association
for the Advancement of Science (AAAS), the American Nuclear Society (ANS), and the American
Society of Mechanical Engineers (ASME). He was awarded the 2008 ANS Seaborg Medal for
outstanding research contributions, the 2003 George Westinghouse Gold Medal for achievements
in power field of mechanical engineering, the 2004 Thermal Hydraulics Technical Achievement
Award by the Thermal Hydraulic Division of the ANS, the 2003 ANS Arthur Holly Compton
Award for contributions to nuclear engineering education and research, and the 2001 Glenn
Murphy Award of the American Association for Engineering Education.
He is the editor-in-chief of Nuclear Engineering and Design, the premier technical journal of the
nuclear engineering field. Dr. Hassan was sworn in as a part-time technical judge to the Atomic
Safety and Licensing Board Panel of the U.S. Nuclear Regulatory Commission.
Abstract: This lecture is intended to provide an overview of efforts of and some insights into the
validation of the nuclear system thermal fluids behavior. In addition to general CFD validation
requirements, we stress out the importance of considering the physics in specific problems for
the validation procedure. We provide specific examples to support the respective main points
addressed using the previous CFD validation experiences from our group’s and other groups’.
For the last few decades, the broad applications of the computational simulation to predict
nuclear system thermal fluid behaviors have been increased. Traditionally, system codes and
sub-channel analysis codes have been successful to a degree for the nuclear safety analysis.
However, because their main role is to combine numerous empirical correlations based on very
simple governing equations to evaluate the synthetic effect, the expected level of accuracy was
relatively low. The validation is performed against global variables (pressure, temperature, void
fraction, liquid level, and so on) in a relatively large control volume (components, sub-channels,
and so on). On the other hand, in recent years, with the advances in numerical algorithms and the
computing power increase, the CFD has become a viable tool to predict the detailed local
thermal fluids behavior in the nuclear system with a higher accuracy than system and
sub-channel codes.
However, because the nuclear system, in general, involves very complex phenomena
characterized as high turbulence and/or multi-phase during not only transient events and
accidents but also normal operating conditions, the models used in the CFD code for the nuclear
system problems still involves empirical closure models, such as RANS equations with
turbulence closure models for turbulence calculations and two-fluid model equations with
closure models for two-phase predictions. These approaches are used in industrial/practical
applications, rather than less-empirical models, such as wall-resolved LES/DNS turbulence
models and interfacial tracking models. Once some assumptions and related modeling are
involved in the CFD code, there is a need to understand the physical phenomena to capture the
flow behavior phenomena. This is to assure that the assumptions and modeling have to have
capability of reflecting key phenomenon. Therefore a need of high-fidelity CFD grade data. This
presentation will address some of the experiments.
Presentation 1: An investigation of thermal stress characteristics to improve evaluation
method for thermal fatigue at a T–junction pipe
Dr. Koji Miyoshi, Institute of Nuclear Safety System, Inc. (INSS)
Dr. Koji Miyoshi graduated Kobe University in 1999. He entered the Kansai Electric Power
Company. He worked at the Maintenance and Repair Department in a nuclear power plant.
Since 2009, he have worked on thermal striping phenomena at Thermal Hydraulics and
Mechanics Group in INSS. He obtained his PhD in this field March, 2015.
Abstract: Thermal fatigue cracking may initiate at a T-junction pipe where high and low
temperature fluids flow in from different directions and mix. Thermal stress is caused by a
temperature gradient in a structure and by its variation. It is possible to obtain stress distributions
if the temperature distributions at the pipe inner surface are obtained by experiments. The wall
temperature distributions at a T-junction pipe were measured by experiments. A total of 148
thermocouples were installed in the pipe inner surface in the downstream region. The thermal
stress distributions were calculated using the experimental data by FEM analysis. The
circumferential and axial stress fluctuations were larger than the radial stress fluctuation range.
The stress fluctuation at the position of the maximum stress fluctuation had 10sec period. The
distribution of the stress fluctuation was similar to that of the temperature fluctuation. The large
stress fluctuations were caused by the time variation of the heating region by the hot jet flow.
The obtained data of three-dimensional stress will be available to validate the results of the
fluid-structure coupled numerical simulations.
Presentation 2: CFD application to thermal hydraulic problems in components
Professor Hiroyasu Mochizuki, Research Institute of Nuclear Engineering, University of Fukui
Prof. Mochizuki started his education with B.S. degree in 1972, followed by M.S. and pursued
Ph.D. in 1979 all at Tokyo Institute of Technology, Tokyo, Japan. Professional carrier started in
1978 and continued at Japan Atomic Energy Agency (JAEA) till 2000 as a scientist in the area of
reactor thermal hydraulics. He became the Director of Paris Office and served between 2000 and
2003 to manage the liaison office between CEA and JAEA. From 2006 onwards, he dedicated
himself towards education and worked as Visiting Professor of Graduate School of Nuclear
Power and Energy Safety Engineering, University of Fukui till 2007 and became Professor by
2008 at the same venue. From 2009 onwards, he is the Professor of Research Institute of Nuclear
Engineering, Univ. of Fukui.
Abstract: The present presentation introduces a conjugate analysis of fluid heat transfer and
stress at a T-junction and a thermal-hydraulic analysis in an intermediate heat exchanger (IHX)
using an ANSYS/CFX code. The former calculation was conducted in order to analyze the
benchmark problem proposed by OECD/NEA. The latter one was conducted in order to solve the
problem of liquid metal heat transfer degradation in the low Péclet number region.
A calculation model of the T-junction simulates the test equipment used for the experiments and
additionally takes into account piping thickness (3, 5, 10 mm), although the test section was
made of plastic. The various parameters are calculated using the CFX code, i.e., temperature
distributions around the T-junction, flow detachment and attachment, structure of eddies
produced by the cold and hot water, temperature fluctuation frequencies of liquid and structure
inside. The calculated results in terms of fluid are compared to the measured results. Calculated
temperatures of the pipe coincide with measured results. However, goof agreement is not
obtained for the temperatures coincide with measured temperatures in general. However, the
wall side temperatures are over predicted. In terms of velocity, good agreement is obtained
near the wall but there is a discrepancy at the center region of the pipe.
As for the heat transfer coefficient of an IHX at 50MW Steam Generator Facility which
simulated an IHX of Monju plant, the Nusselt number evaluated by measured data indicated that
the values were far below the normal heat transfer coefficient under the low Pe number
conditions. An IHX of the “Monju” reactor was calculated using the CFX code. The calculated
temperatures in the IHX are compared with the measured results, and good agreement has been
obtained. The calculated results show that the heat transfer in the lower plenum of the IHX
becomes large as the flow rate is lowered. Inspired by the CFD calculation, one
counter-current-type heat exchanger is analyzed using a model based on three separated
countercurrent heat exchanger models. The Nusselt number calculated from measured
temperature and flow rate data in the 50MW SG is correctly reproduced by the calculation which
uses the well-known Nu-Pe correlation, when the calculated result was processed in the same
way as the experiment using the heat exchanger theory. Finally it was clarified that the
deviation of the Nu number under the low Pe number region is a superficial phenomenon which
is caused by the heat transfer in the plena of the heat exchanger.
Presentation 3: Application of CFD for LWR safety problems
Professor Tadashi Watanabe, Research Institute of Nuclear Engineering, University of Fukui
Ph. D. degree in Nuclear Engineering at Tokyo Institute of Technology, Japan, in 1985. Research
engineer in the reactor safety division of Japan Atomic Energy Agency since 1985, and Professor
in the research institute of nuclear engineering, University of Fukui, since 2012. The major
research fields are nuclear reactor thermal hydraulics, reactor safety analysis, and two-phase
flow simulations.
Abstract: The application of CFD codes for Light Water Reactor (LWR) safety problems is
presented. One example is the coupling of a CFD code and a system analysis code for studying
the instability of single-phase natural circulation in steam generator U tubes. The other is the
usage of CFD results in structural analysis as the boundary conditions for studying the
pressurized thermal shock.

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