Effects of Primary Water Environment on the Thermal Aged Cast Austenitic Stainless Steels

2015 ◽  
Author(s):  
Yuichi Fukuta ◽  
Hiroshi Kanasaki ◽  
Takahisa Yamane
Keyword(s):  
Fracture Toughness ◽  
Stainless Steels ◽  
Water Environment ◽  
Prediction Method ◽  
Austenitic Stainless Steels ◽  
Charpy Impact ◽  
Tensile Tests ◽  
Ferrite Content ◽  
Pressurized Water ◽  
Primary Water

This report summarizes the results of a scoping fracture toughness tests at high and low temperature for thermally aged cast austenitic stainless steels (CASSs) in a pressurized water reactor (PWR) environment. CF8M (ferrite content = 10.1%, 18.9%) and CF8 (ferrite content = 10.5%) were thermally aged up to 5,000 hours at 465°C. Tensile tests, Charpy impact tests and fracture toughness tests were conducted in air at 325°C and 50°C. Fracture toughness tests were also performed in simulated PWR primary water. Although the effect of 325°C and 50°C in simulated PWR primary water and dissolved hydrogen on the fracture toughness (JIc and J-Δa relationship) were slightly observed, fracture toughness was greater than that predicted by the thermally aged fracture toughness prediction method (Hyperbolic-Time-Temperature-Toughness (H3T) model).

Study on Effects of Non-Isothermal Condition and Strain Holding on Environmentally Assisted Fatigue in PWR Primary Water Environment (Step III)

2019 ◽  
Author(s):  
Daiki Takagoshi ◽  
Seiji Asada ◽  
Yuichirou Nomura ◽  
Gary Stevens ◽  
Kawaljit Ahluwalia
Keyword(s):  
Water Environment ◽  
Isothermal Condition ◽  
Austenitic Stainless Steels ◽  
Fatigue Tests ◽  
Small Scale ◽  
Pressurized Water ◽  
Transient Conditions ◽  
Isothermal Conditions ◽  
Primary Water ◽  
Isothermal Test

Abstract To further understand the environmentally assisted fatigue (EAF) behavior of Type 316 austenitic stainless steels (SS) in a pressurized water reactor (PWR) primary water environment, the influence of non-isothermal transient conditions was investigated using hollow, small-scale fatigue specimens. In our previous study (Step I: PVP2016-63798), isothermal and non-isothermal EAF tests were performed. The non-isothermal transient conditions for the fatigue tests investigated transients that consisted of both in-phase and out-of-phase temperature and strain variations. The result of this test series showed longer fatigue lives than those predicted using the modified rate approach and the EAF multiplier (Fen) presented in the draft of NUREG/CR-6909, Revision 1. In our follow-on study (Step II: PVP2017-66233), one extra non-isothermal test was performed and application of the Weighted Stress Intensity Factor (K) Rate (WKR) method was discussed. In the latest Step III testing, an extra non-isothermal test with periodic beachmarking has been performed. This paper discusses, based on the completion of all EAF tests performed in Steps I – III of this project, the effects of non-isothermal conditions and recommendations which are made for treatment of non-isothermal conditions in EAF assessments.

Fatigue Crack Growth Threshold of Austenitic Stainless Steels in Simulated PWR Primary Water

2007 ◽  
Author(s):  
Kazuya Tsutsumi ◽  
Kenji Yamamoto ◽  
Yoshikazu Nitta
Keyword(s):  
Fatigue Crack ◽  
High Temperature ◽  
Crack Growth ◽  
Stainless Steels ◽  
Evaluation Method ◽  
Water Environment ◽  
Austenitic Stainless Steels ◽  
High Temperature Water ◽  
Primary Water ◽  
Temperature Water

Fatigue crack propagation behaviors in PWR environment can be evaluated by using crack growth rate (CGR) curves which are given in JSME code on Fitness-for-Service for nuclear power plant. The CGR curves, however, are only defined in crack growth region and crack growth thresholds are not considered. Since there is a case that stresses in low ΔK region is applied to the components in case of fatigue, it is needed to investigate near-threshold fatigue CGR to establish fatigue assessment. In this study, CGR tests for stainless steels were carried out, and CGR in the region and ΔKth were obtained in simulated PWR primary water. It was found that CGR was accelerated in high temperature water compared to that in air and ΔKth existed even in water environment. ΔKth was influenced by temperature, stress ratio and frequency, independently of materials. Oxide-induced-crack-closure has an important role in high temperature water. ΔKth was formulated and ΔKth evaluation method, whose accuracy were ±25% between experimental data and evaluation value, was proposed.

Fracture Toughness and Deformation Behavior of Cast Austenitic Stainless Steels After Thermal Aging

2017 ◽  
Author(s):  
Yiren Chen ◽  
Wei-Ying Chen ◽  
Chi Xu ◽  
Xuan Zhang ◽  
Zhangbo Li ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Stainless Steels ◽  
Thermal Aging ◽  
Cooling System ◽  
Austenitic Stainless Steels ◽  
Ferrite Content ◽  
Delta Ferrite ◽  
Light Water ◽  
Water Reactors ◽  
The Impact

Cast austenitic stainless steels (CASSs) are used in the cooling system of light water reactors (LWRs) for components with complex shapes, such as pump casings, valve bodies, coolant piping, etc. The CF grades of CASS alloys are the cast equivalents of 300-series stainless steels (SSs) and show excellent mechanical properties and corrosion resistance. In contrast to the fully austenitic microstructure of wrought SSs, CASS alloys consist of a dual-phase microstructure of delta ferrite and austenite and are vulnerable to thermal aging embrittlement. The service performance of CASS alloys is of concern after long-term exposure to high-temperature coolant. In this work, we studied the effects of thermal aging and ferrite content on the fracture resistance of CASS alloys. Crack growth rate and fracture toughness J–R curve tests were performed on aged and unaged CASS alloys in simulated light water reactor environments. The impact of thermal aging on the cracking susceptibility was investigated and the effect of ferrite content was evaluated. Significant embrittlement was observed in the CASS alloys after aging at 400°C. To understand the embrittlement mechanism, microstructural characterizations were performed with transmission electron microscope. The thermal aging produced G-phase precipitates and phase separation in the ferrite, but did not affect the microstructure of austenite. Consequently, the ferrite was hardened considerably after thermal aging while the hardness of austenite phase remained unchanged. The difference in hardness created a high incompatible strain at the interface between ferrite and austenite, leading to fracture at phase boundaries.

EPRI P87

Alloy Digest ◽  
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.

CARPENTER STAINLESS 304+B

Alloy Digest ◽  
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.

ALLOY 0Cr25Ni6Mo3CuN

Alloy Digest ◽  
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.

Fatigue Crack Growth Curve for Austenitic Stainless Steels in PWR Environment

2004 ◽  
Author(s):  
Yuichiro Nomura ◽  
Kazuya Tsutsumi ◽  
Hiroshi Kanasaki ◽  
Naoki Chigusa ◽  
Kazuhiro Jotaki ◽  
...  
Keyword(s):  
Fatigue Crack ◽  
Fatigue Crack Growth ◽  
Crack Growth ◽  
Growth Curve ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Stress Intensity Factor Range ◽  
Reference Curve ◽  
Pressurized Water ◽  
Crack Growth Curve

Although reference fatigue crack growth curves for austenitic stainless steels in air environments and boiling water reactor (BWR) environments were prescribed in JSME S NA1-2002, similar curves for pressurized water reactors (PWR) were not prescribed. In order to propose the reference curve in PWR environment, fatigue tests of austenitic stainless steels in simulated PWR primary water environment were carried out. According to the procedure to determine the reference fatigue crack growth curve of BWR, which of PWR is proposed. The reference fatigue crack growth curve in PWR environment have been determines as a function of stress intensity factor range, Temperature, load rising time and stress ratio.

Irradiation Hardening of Austenitic Stainless Steels: From Field Experience to Modeling

2004 ◽  
Author(s):  
C. Pokor ◽  
Y. Thebault ◽  
C. Pujol ◽  
J.-P. Massoud ◽  
D. Loisnard ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Stainless Steels ◽  
Austenitic Stainless Steels ◽  
Pressurized Water Reactors ◽  
Iterative Approach ◽  
Pressurized Water ◽  
Irradiation Hardening ◽  
The Core ◽  
Water Reactors ◽  
Life Predictions

Materials for the core internals of Pressurized Water Reactors (austenitic stainless steels) are submitted to neutron irradiation. To understand the ageing mechanisms associated to irradiation and propose life predictions of the component, a multi step iterative approach consisting in particular in modeling the evolution of the hardening has been undertaken. Combination of characterization and modeling of simplified situations and field expertise is proposed.

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