JSME Construction Standard for Superconducting Magnet of Fusion Facility “Procedure for Structural Design”

2009 ◽  
Author(s):  
Yukio Takahashi ◽  
Shigeru Tado ◽  
Kazunori Kitamura ◽  
Masataka Nakahira ◽  
Junji Ohmori ◽  
...  
Keyword(s):  
Stainless Steels ◽  
Structural Design ◽  
Structural Integrity ◽  
Superconducting Magnets ◽  
Austenitic Stainless Steels ◽  
Superconducting Magnet ◽  
Fatigue Curve ◽  
Stress System ◽  
Allowable Stress ◽  
Fatigue Assessment

Superconducting magnets are structures which have an important role in Tokamak-type fusion reactor plants. They are huge and complicated structures exposed to very low temperature, 4K and the methods for keeping their integrity need to be newly developed. To maintain their structural integrity during the plant operation, a procedure for structural design was developed as a part of JSME Construction Standard for Superconducting Magnet. General structures and requirements of this procedure basically follow those of class 1 and class 2 components in light water reactor plants as specified in Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, and include the evaluation of primary stress, secondary stress and fatigue damage. However, various new aspects have been incorporated considering the features of superconducting magnet structures. They can be summarized as follows: (i) A new procedure to determine allowable stress intensity value was employed to take advantage of the excellent property of newly developed austenitic stainless steels. (ii) Allowable stress system was simplified considering that only austenitic stainless steels and a nickel-based alloy are planned to be used. (iii) A design fatigue curve at 4K was developed for austenitic stainless steels. (iv) In addition to the conventional fatigue assessment based on design fatigue curves, guidelines for fatigue assessment based on crack growth prediction were added as a non-mandatory appendix to provide a tool of assurance for welded joints which are difficult to evaluate nondestructively during the service.

JSME Construction Standard for Superconducting Magnet of Fusion Facility: “General View of the Code”

2009 ◽  
Author(s):  
Yuji Nakasone ◽  
Yukio Takahashi ◽  
Arata Nishimura ◽  
Tetsuya Suzuki ◽  
Hirosada Irie ◽  
...  
Keyword(s):  
Structural Design ◽  
Radiation Effects ◽  
Structural Integrity ◽  
Superconducting Magnets ◽  
Fusion Energy ◽  
Superconducting Magnet ◽  
General View ◽  
Deformation Control ◽  
Roles And Responsibilities ◽  
Set Up

The Japan Society of Mechanical Engineers (JSME) has set up and published the construction standard for superconducting magnet structures to be used in nuclear fusion facilities. The present target of the standard is tokamak-type fusion energy facilities, especially the International Thermonuclear Experimental Reactor called ITER for short. The standard contains rules for structural materials including cryogenic materials, structural design considering magnetic forces, manufacture including welding and installation, nondestructive testing, pressure proof tests and leak tests of toroidal field magnet structures. The standard covers requirements for structural integrity, deformation control, and leak tightness of all the components of the superconducting magnets and their supports except for superconducting strands and electrical insulators. The standard does not cover deterioration which may occur in service as a result of corrosion, radiation effects, or instability of material. The standard consists of seven articles and twelve mandatory and non-mandatory appendices to the articles; i.e., (1) Scope, roles and responsibilities, (2) Materials, (3) Structural design, (4) Fabrication and installation, (5) Non-destructive examination, (6), Pressure and leak testing, and (7) Terms used in general requirement. The present paper describes the general view of the standard. The detailed descriptions of the standard are described by the other papers in this session CS-21, being divided into four categories; i.e., (1) quality assurance, (2) materials, (3) structural design and replacement, and (4) welding, bonding and examination.

New Cryogenic Steels and Design Approach for ITER Superconducting Magnet System

2002 ◽  
Author(s):  
H. Nakajima ◽  
K. Hamada ◽  
K. Okuno ◽  
K. Hada ◽  
E. Tada
Keyword(s):  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Superconducting Magnet ◽  
Design Approach ◽  
Allowable Stress ◽  
Helium Temperature ◽  
New Materials ◽  
Design Code ◽  
Magnet System ◽  
Superconducting Magnet System

A new design code has been developed for construction and operation/maintenance of the International Thermonuclear Experimental Reactor (ITER). A superconducting magnet system is one of the key components of ITER and its design code includes new cryogenic materials and design approach with taking account of unique features of a performance of the superconducting magnet. The new materials are nitrogen strengthened austenitic stainless steels, which have a yield strength (Sy) of over 1000 MPa and fracture toughness (KIc) of over 200 MP√m at liquid helium temperature (4K). The feature of the design approach is use of the allowable stress defined by only 2/3 Sy measured at 4K. A concept and reliability of the new design approach using new cryogenic materials for the ITER superconducting magnet system are discussed in this paper.

EDF/NRC High-Cycle Fatigue Database Proposal

2015 ◽  
Author(s):  
T. P. Métais ◽  
G. Stevens ◽  
G. Blatman ◽  
J. C. Le Roux ◽  
R. L. Tregoning
Keyword(s):  
Stainless Steels ◽  
Data Exchange ◽  
High Cycle Fatigue ◽  
Fatigue Testing ◽  
Austenitic Stainless Steels ◽  
Nuclear Regulatory Commission ◽  
Fatigue Curve ◽  
Research Data ◽  
Cycle Fatigue ◽  
Fatigue Data

Revised fatigue curves for austenitic stainless steels are currently being considered by several organizations in various countries, including Japan, South Korea, and France. The data available from laboratory tests indicate that the mean air curve considering all available austenitic material fatigue data may be overly conservative compared to a mean curve constructed from only those data representative of a particular type of material. In other words, developing separate fatigue curves for each of the different types of austenitic materials may prove useful in terms of removing excess conservatism in the estimation of fatigue lives. In practice, the fatigue curves of interest are documented in the various international design codes. For example, in the 2009 Addenda of Section III of the ASME Boiler and Pressure Vessel (BPV) Code [1], a revised design air fatigue curve for austenitic materials was implemented that was based on NRC research models [2]. More recently, in Japan, various industrial groups have joined their efforts to create the Design Fatigue Curve Sub-committee (DCFS) with the objective to reassess the fatigue curves [3]. In France, EDF/AREVA and CEA are developing a new fatigue curve for austenitic stainless steels [4]. More specifically, in 2014, EDF presented a paper on high-cycle fatigue analysis which demonstrated that the factor on the strain amplitude could be reduced from 2 to 1.4 for the RCC-M austenitic stainless steel grades [5]. Recently, discussions between EDF and the U.S. Nuclear Regulatory Commission (NRC) have led both parties to recognize that there is a need to exchange worldwide research data from fatigue testing to promote a common, vetted database available to all researchers. These discussions have led EDF and NRC to pursue a collaborative agreement and associated fatigue data exchange, with the intent to assemble all available fatigue data for austenitic materials into a standardized format. The longer term objective is to perform common analyses on the consolidated set of data. This paper summarizes the intent and of the preliminary results of this cooperation and also provides insights from both organizations on possible future activities and participation in the global exchange of fatigue research data.

Retardation Effect of Cyclic Overload on Stress Corrosion Crack Growth in Stainless Steel

2017 ◽  
Author(s):  
Toshiyuki Saito ◽  
Takahiro Hayashi ◽  
Chihiro Narazaki ◽  
Mikiro Itow
Keyword(s):  
Stress Corrosion ◽  
Stainless Steels ◽  
Power Plants ◽  
Structural Integrity ◽  
Plastic Region ◽  
Austenitic Stainless Steels ◽  
Growth Behavior ◽  
Welding Residual Stress ◽  
Earthquake Force ◽  
Cyclic Overload

Stress Corrosion Cracking (SCC) has been observed in some components of austenitic stainless steels in the Boiling Water Reactors (BWRs). The structural integrity evaluation for flawed component is performed for continued service for a specified time period based on the Rules on Fitness-for-Service (FFS) for Nuclear Power Plants, such as JSME FFS Code or ASME Section XI. SCC growth evaluation is generally performed only by taking into account steady loads, such as welding residual stress. It is important to examine various factors affecting SCC growth behavior for further understanding and improvement in predicting growth behavior in the BWR environment. Cyclic overloading due to such as earthquake force is one of the important factors to be evaluated. In this study, the effect of cyclic overload on SCC growth in simulated BWR environment has been examined by using CT specimens of cold-rolled stainless steels (Type 316L). The retardation phenomenon was observed in SCC growth behavior immediately after the cyclic overloading was applied. It was considered that SCC propagation was retarded due to the compressive plastic region at the crack tip, introduced by overloads. The method of predicting the SCC growth behavior after cyclic overloading was also discussed.

scholarly journals Development of a Water Environment Fatigue Design Curve for Austenitic Stainless Steels

2003 ◽  
Author(s):  
Thomas R. Leax
Keyword(s):  
Stainless Steels ◽  
Fatigue Behavior ◽  
Water Environment ◽  
Austenitic Stainless Steels ◽  
Fatigue Curve ◽  
Light Water Reactor ◽  
Design Curve ◽  
Fatigue Design ◽  
American Society ◽  
Design Factors

Technical support is provided for a fatigue curve that could potentially be incorporated into Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. This fatigue curve conservatively accounts for the effects of light water reactor environments on the fatigue behavior of austenitic stainless steels. This paper presents the data, statistical methods, and basis for the design factors appropriate for Code applications. A discussion of the assumptions and methods used in design curve development is presented.

Cryogenic Properties of Austenitic Stainless Steels for Superconducting Magnet

1983 ◽  
pp. 117-133 ◽  
Author(s):  
Kiyohiko Nohara ◽  
Tsunehiko Kato ◽  
Terufumi Sasaki ◽  
Shigeharu Suzuki ◽  
Yutaka Ono
Keyword(s):  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Superconducting Magnet

Fracture Toughness Criteria of Irradiated Austenitic Stainless Steels for Structural Integrity Evaluation of BWR Internal Components

2019 ◽  
Author(s):  
Takahiro Hayashi ◽  
Shigeaki Tanaka ◽  
Tomonori Abe ◽  
Seiji Sakuraya ◽  
Suguru Ooki ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Stainless Steels ◽  
Power Plants ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Austenitic Stainless Steels ◽  
Evaluation Methodology ◽  
Irradiation Damage ◽  
Materials Testing ◽  
Testing Data

Abstract Continuous improvement of the structural integrity evaluation methodology in the plant life management (PLM) evaluations is of increasing importance for aged light water reactors. In PLM evaluations, structural integrity evaluations are required for degradation mechanisms considered in the subject equipment and components. Austenitic stainless steels used in reactor internal components are known to show decreases in ductility and fracture toughness due to accumulated neutron irradiation damage. In Japan, “Rules on Fitness-for-Service for Nuclear Power Plants of the Japan Society of Mechanical Engineers Code (JSME FFS Code)” provides fracture evaluation method and criterion, based on the linear elastic fracture mechanics, for irradiated stainless steels of boiling water reactor (BWR) internal components. The fracture toughness criterion, however, was developed with limited materials testing data and knowledge available at that time and it has not been revised since the code originally established. In this study, fracture toughness criteria for structural integrity evaluation were discussed and developed with the latest database on fracture toughness of irradiated austenitic stainless steels, including additional material testing data obtained in this study for the neutron fluence range of interest from 1 to 3 dpa. First, the fracture toughness data of austenitic stainless steels irradiated in BWR conditions were compiled to evaluate the correlation between fracture toughness and neutron fluence. Material characteristics potentially affecting fracture toughness, such as chemical composition and specimen orientation, were also considered and discussed in the development of the fracture toughness criteria. Based on the results, the fracture toughness criteria for irradiated austenitic stainless steels were proposed for fracture evaluation of the BWR internal components.

Plan and Status of Development of Design Fatigue Curves: Phase 1

2013 ◽  
Author(s):  
Seiji Asada ◽  
Akihiko Hirano ◽  
Masao Itatani ◽  
Munehiro Yasuda ◽  
Takehiko Sera ◽  
...  
Keyword(s):  
Stainless Steels ◽  
Phase 1 ◽  
Austenitic Stainless Steels ◽  
Fatigue Curve ◽  
Carbon Steels ◽  
Low Alloy Steels ◽  
Energy Research ◽  
Research Committee ◽  
Engineering Society ◽  
Alloy Steels

In order to develop and propose new design fatigue curves for austenitic stainless steels, carbon steels and low alloy steels that are rational and have clear design basis, Design Fatigue Curve (DFC) subcommittee has been established in the Atomic Energy Research Committee in the Japan Welding Engineering Society and the study on design fatigue curves are going on. This paper introduces the plan and status of the activities of the DFC subcommittee.

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