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

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.

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

Volume 6B: Materials and Fabrication ◽

10.1115/pvp2017-65968 ◽

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.

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Nondestructive Evaluation of Neutron Irradiation Damage on Austenitic Stainless Steels by Measurement of Magnetic Flux Density

Volume 1: Plant Operations, Maintenance, Engineering, Modifications and Life Cycle; Component Reliability and Materials Issues; Next Generation Systems ◽

10.1115/icone17-75215 ◽

2009 ◽

Author(s):

Shigeru Takaya ◽

Yuji Nagae ◽

Kazumi Aoto ◽

Ichiro Yamagata ◽

Shoichi Ichikawa ◽

...

Keyword(s):

Magnetic Flux ◽

Nondestructive Evaluation ◽

Flux Density ◽

Stainless Steels ◽

Evaluation Method ◽

Austenitic Stainless Steels ◽

Irradiation Damage ◽

Magnetic Flux Density ◽

Materials Testing ◽

Displacement Per Atom

Magnetic flux densities for neutron irradiated specimens of austenitic stainless steels, SUS304 and Fast Breeder Reactor grade type 316 (316FR), were measured by using a flux gate (FG) sensor to investigate the nondestructive evaluation method of irradiation damage parameters, dose and He content. Specimens were irradiated in each one of the experimental fast reactor JOYO, the Japan Materials Testing Reactor, and the Japan Research Reacter-3M (JRR-3M), or in both of JRR-3M and JOYO (coupling irradiation). Irradiation in various reactors and the coupling irradiation provided irradiation conditions which could be hardly obtained by irradiation in a single reactor. The range of dose, He content and irradiation temperature of the neutron irradiated samples studied in this paper were 0.01–30 displacement per atom (dpa), 1.0–17 appm and 470–560 °C, respectively. Magnetic flux density increased with dose although there may be a threshold dose for magnetic property to change between 2 and 5 dpa for 316FR. This result shows the possibility of nondestructive evaluation of dose by measuring magnetic flux density by an FG sensor. On the other hand, magnetic flux density did not depend on He content.

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EPRI P87

Alloy Digest ◽

10.31399/asm.ad.ni0685 ◽

2011 ◽

Vol 60 (1) ◽

Keyword(s):

Fracture Toughness ◽

Tensile Properties ◽

Stainless Steels ◽

Austenitic Stainless Steels

Abstract EPRI P87 is a MMA electrode designed for dissimilation joints between austenitic stainless steels (i.e. 304H) and a creep resisting CrMo alloy (i.e. P91). This datasheet provides information on composition and tensile properties as well as fracture toughness. It also includes information on joining. Filing Code: Ni-685. Producer or source: Metrode Products Ltd.

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CARPENTER STAINLESS 304+B

Alloy Digest ◽

10.31399/asm.ad.ss0121 ◽

1961 ◽

Vol 10 (9) ◽

Keyword(s):

Fracture Toughness ◽

Cross Section ◽

Stainless Steels ◽

Austenitic Stainless Steels ◽

Heat Treating ◽

Neutron Absorption ◽

Absorption Cross ◽

Type 304 ◽

Conventional Type ◽

Neutron Absorption Cross Section

Abstract Carpenter Stainless 304+B is similar to conventional Type 304 with the addition of boron to give it a much higher thermal neutron absorption cross-section than other austenitic stainless steels. This datasheet provides information on composition, physical properties, hardness, elasticity, and tensile properties as well as fracture toughness. It also includes information on corrosion resistance as well as forming, heat treating, machining, joining, and surface treatment. Filing Code: SS-121. Producer or source: Carpenter.

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ALLOY 0Cr25Ni6Mo3CuN

Alloy Digest ◽

10.31399/asm.ad.ss0706 ◽

1998 ◽

Vol 47 (2) ◽

Keyword(s):

Fracture Toughness ◽

Stainless Steel ◽

Corrosion Resistance ◽

High Temperature ◽

Physical Properties ◽

Stainless Steels ◽

Austenitic Stainless Steels ◽

Temperature Performance ◽

Steel Research ◽

Oil Refinement

Abstract ALLOY 0Cr25Ni6Mo3CuN is one of four grades of duplex stainless steel that were developed and have found wide applications in China since 1980. In oil refinement and the petrochemical processing industries, they have substituted for austenitic stainless steels in many types of equipment, valves, and pump parts. This datasheet provides information on composition, physical properties, hardness, elasticity, and tensile properties as well as fracture toughness. It also includes information on low and high temperature performance, and corrosion resistance as well as forming and joining. Filing Code: SS-706. Producer or source: Central Iron & Steel Research Institute.

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Electron irradiation damage in austenitic stainless steels

Journal of Nuclear Materials ◽

10.1016/0022-3115(91)90141-s ◽

1991 ◽

Vol 179-181 ◽

pp. 526-528 ◽

Cited By ~ 5

Author(s):

Jiguang Sun ◽

Jiapu Qian ◽

Zhuoyong Zhao ◽

Jiming Chen ◽

Zengyu Xu

Keyword(s):

Electron Irradiation ◽

Stainless Steels ◽

Austenitic Stainless Steels ◽

Irradiation Damage

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Localized deformation as a key precursor to initiation of intergranular stress corrosion cracking of austenitic stainless steels employed in nuclear power plants

Journal of Nuclear Materials ◽

10.1016/j.jnucmat.2010.01.029 ◽

2010 ◽

Vol 406 (1) ◽

pp. 138-151 ◽

Cited By ~ 47

Author(s):

Wade Karlsen ◽

Gonzalo Diego ◽

Bastian Devrient

Keyword(s):

Stress Corrosion ◽

Stress Corrosion Cracking ◽

Stainless Steels ◽

Nuclear Power ◽

Power Plants ◽

Austenitic Stainless Steels ◽

Corrosion Cracking ◽

Localized Deformation ◽

Intergranular Stress Corrosion ◽

Intergranular Stress Corrosion Cracking

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

Volume 1: Codes and Standards ◽

10.1115/pvp2009-77991 ◽

2009 ◽

Cited By ~ 1

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.

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Fracture Toughness Predictive Models Within the SOTERIA EU Project

Volume 6A: Materials and Fabrication ◽

10.1115/pvp2018-85074 ◽

2018 ◽

Cited By ~ 1

Author(s):

P. M. James ◽

M. Berveiller

Keyword(s):

Cleavage Fracture ◽

Nuclear Power ◽

Power Plants ◽

Radiation Effects ◽

Austenitic Stainless Steels ◽

Irradiation Damage ◽

Tensile Behaviour ◽

Horizon 2020 ◽

Term Operation

SOTERIA is focused on the ‘safe long term operation of light water reactors’. This will be achieved through an improved understanding of radiation effects in nuclear structural materials. This project has received funding from the European Union’s Horizon 2020 research and innovation programme under agreement No 661913. The overall aim of the SOTERIA project is to improve the understanding of the ageing phenomena occurring in ferritic reactor pressure vessel steels and in the austenitic internals in order to provide crucial information to regulators and operators to ensure safe long-term operation (LTO) of existing European nuclear power plants (NPPs). SOTERIA has set up a collaborative research consortium which gathers the main European research centers and industrial partners who will combine advanced modelling tools with the exploitation of experimental data to focus on two major objectives: i) to identify ageing mechanisms when materials face environmental degradation (such as e.g. irradiation and corrosion) and ii) to provide a single platform containing data and tools for reassessment of structural components during NPPs lifetime. This paper provides an overview of the ongoing activities within the SOTERIA Project that are contained within the analytical work-package (WP5.3). These fracture aspects are focused on the estimates of fracture in both ferritic steels and irradiation assisted stress corrosion cracking (IASCC) in austenitic stainless steels, under irradiated conditions. This analytical development is supported by analytical estimates of irradiation damage and the resulting changes in tensile behaviour of the steels elsewhere in SOTERIA, as well as a wider number of experimental programmes. Cleavage fracture estimates are being considered by a range of modelling estimates including the Beremin, Microstructurally Informed Brittle Fracture Model (MIBF), JFJ and Bordet Models with efforts being made to understand the influence of heterogeneity on the predicted toughness’s. Efforts are also being considered to better understand ductile void evolution and the effect of plasticity on the cleavage fracture predictions. IASCC is being modelled through the INITEAC code previously developed within the predecessor project Perform 60 which is being updated to incorporate recent developments from within SOTERIA and elsewhere.

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