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SYMBIO INTERNATIONAL WORKSHOP

2014

The 17th International Workshop on Nuclear Safety & Simulation Technology (IWNSST17) Date: January 21-22, 2014

Place: Kyoto, Japan

Co-organizers: Symbio Community Forum, Kyoto, Japan Harbin Engineering University, Harbin, China January 21, Tue., 2014、 9:00~17:00 Place: Conference room3, Clock Tower Centennial Hall,

Yoshida campus, Kyoto University, Yoshida-Honmachi, Sakyo-ku, Kyoto, Japan

Program: Opening address by the organizer 9:30~9:40 *Morning session 9:40~12:20 Chair: Prof. Morten Lind

Paper 1. 9:40~10:20 "Corrosion and Oxide Layer Growth Modeling Using Deterministic and Stochastic Methods", by Prof. Yitung Chen, ASME Fellow, Professor and Graduate Program Coordinator, Department of Mechanical Engineering, Co-Director, Center for Energy Research, University of Nevada Las Vegas. Paper 2. 10:20~11:00 ”Reliability Assessment for Detecting and Sizing Pipe Wall Thinning and its

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Application to Risk Management", by Prof. Fumio Kojima, Professor and Research Coordinator, Organization of Advanced Science and Technology and Graduate School of System Informatics, Kobe University. Paper 3. 11:00~11:40 ” Corrosion of Structural Materials and Electrochemistry in High Temperature Water of Nuclear

Power Systems”, by Dr. Shunsuke Uchida, Research Fellow, Institute of Applied Energy.

Paper 4 11:40~12:00

“Introduction of AR Research Activities for Decommissioning of NPP”, by Professor Hiroshi Shimoda, Professor, Department of Socio-environmental Energy Science, Graduate School of Energy Science, Kyoto University *Lunch break 12:00~13:00 “La Tour”

*Afternoon session (I ) 13:00~14:40 Chair: Prof. Takeshi Matsuoka Paper 5 13:00~13:40 ”Short Introduction of Automation Researches at DTU”, by Prof. Ole Ravn, Professor , Department of Electrical Engineering, Danish Technical University, Denmark Paper 6 13:40~14:20 “Functional Modeling of Control Systems”, by Prof. Morten Lind, Professor Emeritus, Danish Technical University, Denmark and Visiting Professor of Harbin Engineering University Coffee break 14:20~14:40

*Afternoon session (II) 14:40~17:20 Chair: Mr. Takashi Nitta (JAPC) Paper 7 14:40~15:20 “Post-facta analysis of Fukushima Daiichi Accident by Simple Physical Model”, by Dr. Fumiya Tanabe, President, Research Institute of Sociotechnical Safety (SOCTEX) Paper 8 15:20~16:00

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”Reliability analyses of PWR safety systems by the GO-FLOW methodology”, by Prof. Takeshi Matsuoka, Professor Emeritus Utsunomiya University and Visiting Professor of Harbin Engineering University Paper 9 16:00~16:40 ” The Automatic Control Design and Simulation of Reactor Control System in Small Modular Reactor", by Mr. Longtao Liao, Nuclear Power Institute of China, Chengdu, Sichuan, China Paper 10 16:00~17:20 “Development of Mitsubishi Computerized Human Machine Interface and Digital I&C system for PWR Plants”, by Mr. Koji Ito, MHI

Closing remarks and miscellaneous message by the organizer 17:20~17:30

Group photo of all participants

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Scene of conference room

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Summary of presentations

Paper 1 Corrosion and Oxide Layer Growth Modeling Using Deterministic and Stochastic Methods Presenter: Yitung Chen, Professor, Department of Mechanical Engineering, University of Nevada Las Vegas, U.S.A. also Visiting Professor, Harbin Engineering University, China

Abstract: A decade ago it was roughly estimated that a total cost of corrosion of $276 billion has been reported in the U.S. industry. This value shows that the impact of corrosion is approximately 3.1 percent of United States’ GNP in that time. This cost is considered to be a conservative estimate since only available well-documented costs were used by the U.S. government. The indirect cost of corrosion was conservatively estimated to be

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equal to the direct cost, giving a total direct plus indirect cost of $552 billion or 6% of the GNP in 2001. The World Corrosion Organization has reported that the annual cost of corrosion worldwide is about $2.2 trillion which is over 3% of the world's GDP in 2010. The National Materials Crosscut Program (NMCP) in the U.S. expects the candidate materials for the nuclear energy applications meet the following design objectives: (1) acceptable dimensional stability including void swelling, thermal creep, irradiation creep, stress relaxation, and growth; (2) acceptable strength, ductility, and toughness; (3) acceptable resistance to creep rupture, fatigue cracking, creep-fatigue interactions, and helium embrittlement; and (4) acceptable chemical compatibility and corrosion resistance (including stress corrosion cracking and irradiation-assisted stress corrosion cracking) in the presence of coolants and process fluids. The fundamentals of corrosion mechanism and oxide layer growth need to be carefully studied in order to understand how it happens and how we can protect the engineering systems and extend the material life span. It is especially extremely important in the application of nuclear energy because corrosion could lead to the loss of coolant accidents (LOCAs) and the reactor damages. The research efforts in studying the corrosion phenomena have been used the different microscopic, meso-scopic, and atomic levels to understand the corrosion mechanism of how and why it occurs under the different fluid flow conditions. The deterministic method has been firstly sought and studied. The simplified analytical solutions have been derived and compared to the numerical and experimental results. The stochastic method has also been used and developed in order to simulate the oxide layer growth. The advantages and disadvantages of using deterministic and stochastic methods will be explained. The developed models have been successfully used in the lead-bismuth eutectic system and the supercritical water system for the different materials and super alloys. The reasonable results and good agreements have been obtained. Presentation PPT

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Paper 2 Reliability Assessment for Detecting and Sizing Pipe Wall Thinning and its Application to Risk Management Presenter: Fumio Kojima, Professor, Organization of Advanced Science and Technology and Graduate School of System Informatics, Kobe University.

Abstract: Pipe wall inspection is aimed at providing a life management process ensuring replacement or repair prior to in-service failure. The application of condition monitoring (CM) to pipe wall inspection plays essential roles in developing existence instrumentation of measurement equipment and together with better performance for optimizing maintenance procedures of piping system. In my research, CM is applied to pipe wall thinning monitoring with hybrid use of ultrasonic guided wave (GW) and electromagnetic acoustic transducer (EMAT). GW enables long-range inspection and is an efficient technique for screening position and direction of any defect on pipe. EMAT has advantages on continuous surveillance on the pipe wall thickness. Both techniques

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allow the remote capabilities of inspection performance and, as a result, the proposed inspection technique is feasible for condition based monitoring of pipe wall inspection. The main part of the research is concerned with the reliability assessment of the proposed inspection technique using probability of detection (POD). In GW, the inspection process involves randomness due to variability in inspection conditions, including inspection strategies for test signals, difficulties associated with the location, and size of pipe wall pinning. Taking into account for these factors, we develop a simple inspection model with the appropriate threshold values that are obtained by maximizing the matching detection events between the model and the inspection data. In GW test, the threshold value of detecting model can be given by POD curve based on hit/miss analysis. In EMAT, a reliability assessment method for the pipe wall thinning measurements using EMAT-EMAR is discussed. The POD function is then evaluated within the common versus a approach. The final part of the report is devoted to applicability and the validity of the POD based assessment method considered here. The structural integrity and safety margins are maintained for the piping systems by providing the acceptance criteria for wall thinning. Such safety margin can be derived by means of the hybrid use of predicting and monitoring. A predictive model for pipe wall thinning rate is formulated with the uncertainty qualification. The prescribed model reveals quite complicated manner that depends on the piping system geometry, operational records in service, environmental conditions, etc. The verification and validation tests are usual procedures in the proper modeling issues. Consequently, the safety margin can be evaluated by following the logarithmic normal distribution. Presentation PPT

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Paper 3 Corrosion of Structural Materials and Electrochemistry in High Temperature Water of Nuclear Power Systems Presenter: Shunsuke Uchida, Research Fellow, Institute of Applied Energy

Abstract: The latest experiences with corrosion in the cooling systems of nuclear power plants are reviewed. High temperature cooling water causes corrosion of structural materials, which often leads to adverse effects in the plants, e.g., increasing shutdown radiation, generating defects in materials of major components and fuel claddings, and increasing the volume of radwaste sources. Corrosion behaviors are much affected by water qualities and differ according to the values of water qualities and the materials themselves. In order to establish reliable operation, each plant requires its own unique optimal water chemistry control based on careful consideration of its system, materials and operational history. Electrochemistry is one of key issues that determine corrosion related problems but it is

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not the only issue. Most phenomena for corrosion related problems, e.g., flow-accelerated corrosion (FAC), intergranular stress corrosion cracking (IGSCC), primary water stress corrosion cracking (PWSCC) and thinning of fuel cladding materials, can be understood based on an electrochemical index, e.g., electrochemical corrosion potential (ECP), conductivities and pH. The most important electrochemical index, ECP, can be measured at elevated temperature and applied to in situ sensors of corrosion conditions to detect anomalous conditions of structural materials at their very early stages. In the presentation, theoretical models based on electrochemistry to estimate wall thinning rate of carbon steel piping due to flow-accelerated corrosion and corrosive conditions determining IGSCC crack initiation and growth rate are introduced. Presentation PPT

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Paper 4 Introduction of AR Research Activities for Decommissioning of NPP Presenter: Hiroshi Shimoda, Professor, Graduate School of Energy Science, Kyoto University

Abstract: This presentation starts with the explanation of “what is AR“ and “Decommission project of Fugen plant in Japan” and then go to where AR technology can be effectively applied for the decommission work of NPP. The presentation will proceed to the plausible technologies of AR application for NPP dismantling procedure and introduce the R&D products by the author’s group for the development of two types of tracking methods, circular marker method and natural feature point extraction. The presentation will finally introduce several field experiments to validate the authors’ development AR methods for the decommissioning work at Fugen plant. Presentation PPT

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Paper 5 Short Introduction of Automation Research at DTU Presenter: Ole Ravn, Head of Group, Automation and Control, Technical University of Denmark

Abstract: Automation and control are important elements in modern engineering and in developing our standard of living in a sustainable manner. The robotics industry is projected to have the same impact on society in the 21st century as the automotive industry had in the 20th century. Mobile robot navigation is a key technology in realizing this vision. The presentation describes the work done at DTU (Technical University of Denmark) in robotics focusing on the technologies enabling robust mobile robot navigation and real-time sensor based robot control. Examples are given of applications in agriculture and meat processing industry. Furthermore the status of recent and current projects will be presented and an overview of current research at Automation and Control is outlined.

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Presentation PPT

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Paper 6 Functional modeling of control systems Presenter: Morten Lind, Professor Emeritus, Department of Electrical Engineering, Technical University of Denmark also Visiting Professor at Harbin Engineering University.

Abstract: Previous research on Multilevel Flow Modeling has developed concepts and tools for representation and reasoning about goals and functions of complex automated processes. MFM is presently applied for modeling nuclear power plants and processes within the oil and gas sector for risk assessment and supervisory control. One of the powerful features of MFM is the ability to represent relations between the physical process and the control systems on multiple levels of functional abstraction. These relations are of importance in analyzing the causes and consequences of disturbances in complex automated processes. MFM concepts have foundations in logical and semantic theories of action. Current MFM research deal with the following three challenges:

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1. Using MFM for reliability assessment of NPP control systems. Of particular interest is the study of type I and type II interactions.

2. Formalizing the representation of operation modes. The challenge is here both to represent the modes themselves and the control actions which provide transition between modes.

3. Development of principles for reasoning about control. “Traditional” techniques to reason about control is based on models of behavior. MFM offer a complementary framework for reasoning about means and ends of control. The presentation will give a short introduction to MFM and its foundation in theories of action. The explanation on how the action theory has contributed to formalization of control functions and how it also will contribute to extensions of MFM addressing with some of the challenges mentioned above.

Presentation PPT

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Paper 7 Post-Facta Analysis of Fukushima Daiichi NPP Accident by Simple Physical Model Presenter: Fumiya Tanabe, Director, Sociotechnical Systems Safety Research Institute (SCTEX) Co. Ltd.

Abstract: Analyses are performed of the first core melt behavior of the Unit 1, Unit 2 and Unit 3 reactors of Fukushima Daiichi Nuclear Power Station on 11-15 March 2011. The analyses are based on a measured data investigation and a simple physical model calculation. Estimated are time variation of core water level behavior, core material temperature and hydrogen generation rate. The analyses reveal characteristics of accident process of each reactor. Analyses are performed also of the re-melt (melt again) behavior in another chaotic period of 19-31 March 2011. Presentation PPT

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Paper 8 Reliability analyses of PWR safety systems by the GO-FLOW methodology Presenter: Takeshi Matsuoka, Lecturer, Utsunomiya University and Visiting Professor of Harbin Engineering University,

Abstract: Several advanced Pressurized Water Reactors (PWRs) as AP1000 have been proposed. The AP1000 employs passive safety systems, in order to provide significant improvement in plant safety design. A passive safety system is defined as a safety system which operation is only relied on passive components. A passive component does not require any external input or energy for its operation, and only relies on natural physical laws (gravity, natural circulation, conduction, etc ). In the present paper, the reliability or availability of AP1000 passive safety systems are evaluated by the GO-FLOW methodology, and compared with active safety systems of conventional PWR plants. The AP1000 safety systems have two main systems, passive core cooling system (PXS)

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and passive containment cooling system (PCCS). The conventional PWR safety systems consist of containment spray system (CSS) and emergency core cooling system (ECCS) which are active safety systems, that is, they have active components as water injection pump, motor operated valve, and so on. The PXS consists of the following subsystems, Passive safety injection system (PSIS), four stages of an automatic depressurization system (ADS), passive residual heat removal system (PRHRS). There are two modes of operation in the PCSS system, five modes in the PXS, and four stages of ADS operations. The operations of these safety systems have dynamical characteristics, and make phased mission problems. The GO-FLOW methodology is well adopted to analyze dynamical system behavior and phased mission problems. Also important information could be obtained by the common mode failure and uncertainty analyses. These two kinds of analyses are also performed in the comparisons of AP1000 and conventional PWR Safety systems. Presentation PPT

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Paper 9 The Automatic Control Design and Simulation of Reactor Control System in Small Modular Reactor Presenter: Longtao Liao, Engineer, Nuclear Power Institute of China

Abstract: In China, the development and application of Small Modular Reactor (SMR) aims at electricity generation, heat supply and seawater desalination, etc. The SMR is a pressurized water reactor, and is designed to integrate the once-through steam generator (OTSG) into the reactor pressure vessel. Based on the characteristic of the reactor and the OTSG, the automatic control design of the reactor power and feed water control system are discussed. The simulation results are presented to illustrate the performance of the control scheme. Presentation PPT

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Paper 10 Development of Mitsubishi Computerized Human Mashine Interface and Digital I&C system for PWR Plants Presenter: Koji Ito, Chief Engineer, Mitsubishi Heavy Industries, ltd

Abstract: Fully computerized Human Machine Interface (HMI) system and digitalized Instrumentation and Control (I&C) System of Mitsubishi Heavy Industries, Ltd. (MHI) are to be introduced. This design is currently being applied to the latest Japanese PWR plant and to nuclear power plant I&C modernization program in Japan. Brief introduction of the facilities to be presented and the schedule of the facility visit to MHI on the 22th January are also explained.

Presentation PPT