Measurement of the Deformability of Austenitic Stainless Steels and Nickel-Base Alloys in Light Water Reactor Cores

AbstractA model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 ∼ 325 °C, up to ∼ 10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRs. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels.

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Simulations of advanced reactor cores in research light water reactor LR-0

Nuclear Engineering and Design ◽

10.1016/j.nucengdes.2018.12.007 ◽

2019 ◽

Vol 342 ◽

pp. 205-209 ◽

Cited By ~ 1

Author(s):

E. Losa ◽

M. Košťál ◽

B. Jánský ◽

E. Novák ◽

V. Rypar ◽

...

Keyword(s):

Light Water Reactor ◽

Light Water ◽

Water Reactor ◽

Reactor Cores

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Protective structures on the surface of zirconium components of light water reactor cores: Formation, testing, and prototype equipment

Physics of Atomic Nuclei ◽

10.1134/s1063778815140033 ◽

2015 ◽

Vol 78 (14) ◽

pp. 1646-1648

Author(s):

L. B. Begrambekov ◽

A. A. Gordeev ◽

A. E. Evsin ◽

S. V. Ivanova ◽

A. S. Kaplevsky ◽

...

Keyword(s):

Light Water Reactor ◽

Light Water ◽

Water Reactor ◽

Reactor Cores ◽

Protective Structures

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Fracture Resistance of Cast Austenitic Stainless Steels

Volume 1: Operations and Maintenance, Aging Management and Plant Upgrades; Nuclear Fuel, Fuel Cycle, Reactor Physics and Transport Theory; Plant Systems, Structures, Components and Materials; I&C, Digital Controls, and Influence of Human Factors ◽

10.1115/icone24-60736 ◽

2016 ◽

Author(s):

Y. Chen ◽

W-Y. Chen ◽

A. S. Rao ◽

Z. Li ◽

Y. Yang ◽

...

Keyword(s):

Fracture Toughness ◽

Growth Rate ◽

Corrosion Resistance ◽

Neutron Irradiation ◽

Stainless Steels ◽

Thermal Aging ◽

Austenitic Stainless Steels ◽

Light Water Reactor ◽

Delta Ferrite ◽

Light Water

Cast austenitic stainless steels (CASS) possess excellent corrosion resistance and mechanical properties and are used alongside with wrought stainless steels (SS) in light water reactors for primary pressure boundaries and reactor core internal components. In contrast to the fully austenitic microstructure of wrought SS, CASS alloys consist of a dual-phase microstructure of delta ferrite and austenite. The delta ferrite is critical for the service performance since it improves the strength, weldability, corrosion resistance, and soundness of CASS alloys. On the other hand, the delta ferrite is also vulnerable to embrittlement when exposed to reactor service temperatures and fast neutron irradiations. In this study, the combined effect of thermal aging and neutron irradiation on the degradation of CASS alloys was investigated. Neutron-irradiated CASS specimens with and without prior thermal aging were tested in simulated light water reactor environments for crack growth rate and fracture toughness. Miniature compact-tension specimens of CF-3 and CF-8 alloys were tested to evaluate the extent of embrittlement resulting from thermal aging and neutron irradiation. The materials used are static casts containing more than 23% delta ferrite. Some specimens were thermally aged at 400 °C for 10,000 hours prior to the neutron irradiation to simulate thermal aging embrittlement. Both the unaged and aged specimens were irradiated at ∼320°C to a low displacement damage dose of 0.08 dpa. Crack growth rate and fracture toughness J-integral resistance curve tests were carried out on the irradiated and unirradiated control samples in simulated light water reactor environments with low corrosion potentials. While no elevated crack propagation rates were detected in the test environments, significant reductions in fracture toughness were observed after either thermal aging or neutron irradiation. The loss of fracture toughness due to neutron irradiation seemed more evident in the samples without prior thermal aging. Transmission electron microscope (TEM) examination was carried out on the thermally aged and neutron irradiated specimens. The result showed that both neutron irradiation and thermal aging can induce significant changes in the delta ferrite. A high density of G-phase precipitates was observed with TEM in the thermally aged specimens, consistent with previous results. Similar precipitate microstructures were also observed in the neutron-irradiated specimens with or without prior thermal aging. A more extensive precipitate microstructure can be seen in the samples subjected to both thermal aging and neutron irradiation. The similar precipitate microstructures resulting from thermal aging and neutron irradiation are consistent with the fracture toughness results, suggesting a common microstructural origin of the observed embrittlement after thermal aging and neutron irradiation.

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Technical Basis Part 2 for Code Case N-889: Reference Stress Corrosion Crack Growth Rate Curves for Irradiated Austenitic Stainless Steels in Light Water Reactor Environments

Journal of Pressure Vessel Technology ◽

10.1115/1.4047878 ◽

2020 ◽

Vol 143 (2) ◽

Author(s):

Ernest D. Eason ◽

Raj Pathania ◽

Anders Jenssen ◽

Dennis P. Weakland

Keyword(s):

Stress Corrosion ◽

Stainless Steels ◽

Austenitic Stainless Steels ◽

Light Water Reactor ◽

Calibration Data ◽

Water Reactor ◽

Data Points ◽

Reference Curves ◽

Basis Part ◽

Technical Basis

Abstract A multiyear international data collection, data review, modeling, and implementation project was recently completed, producing stress corrosion crack growth rate (CGR) reference curves for irradiated austenitic stainless steels in light water reactor (LWR) environments that were adopted as ASME B&PV Section XI Code Case N-889. As described in a technical basis Part 1 paper, over 800 CGR data points were collected from six laboratories worldwide, an international expert panel reviewed and ranked the data, and the better-ranked data were used to calibrate empirical models for irradiation-assisted stress corrosion cracking (IASCC) CGR in boiling water reactor (BWR) normal water chemistry (NWC), BWR hydrogen water chemistry (HWC) and pressurized water reactor (PWR) primary water environments. Part 1 also describes the custom fitting process, quality of fit, and comparisons with related literature and data not used for fitting. This technical basis Part 2 paper describes shifting the mean models to the 75th percentile of the calibration data, simplifying to produce the N-889 curves, and comparing with previous reference curves and over 500 data points not used for developing the N-889 curves, including weld, cast, and heat-affected-zone (HAZ) materials, additional wrought laboratory data, and field data from repeated inspection of BWR core shrouds. Part 2 also describes the irradiated yield stress model in Case N-889, compares that model with its calibration data and other data not used for calibration, and presents example calculations using both yield stress and CGR equations.

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Evaluation of Fatigue Damage in LWR Water With and Without Threshold and Moderation Factor

Pressure Vessel and Piping Codes and Standards ◽

10.1115/pvp2003-1774 ◽

2003 ◽

Author(s):

Makoto Higuchi ◽

Kazuya Tsutsumi ◽

Katsumi Sakaguchi

Keyword(s):

Stainless Steel ◽

Fatigue Life ◽

Stainless Steels ◽

Power Plants ◽

Light Water Reactor ◽

Low Alloy Steels ◽

Light Water ◽

Water Reactor ◽

Fatigue Data ◽

Fatigue Life Reduction

During the past twenty years, the fatigue initiation life of LWR structural materials, carbon, low alloy and stainless steels has been shown to decrease remarkably in the simulated LWR (light water reactor) coolant environments. Several models for evaluating the effects of environment on fatigue life reduction have been developed based on published environmental fatigue data. Initially, based on Japanese fatigue data, Higuchi and Iida proposed a model for evaluating such effects quantitatively for carbon and low alloy steels in 1991. Thereafter, Chopra et al. proposed other models for carbon, low alloy and stainless steels by adding American fatigue data in 1993. Mehta developed a new model which features the threshold concept and moderation factor in Chopra’s model in 1995. All these models have undergone various revisions. In Japan, the MITI (Ministry of International Trade and Industry) guideline on environmental fatigue life reduction for carbon, low alloy and stainless steels was issued in September 2000, for evaluating of aged light water reactor power plants. The MITI guideline provide equations for calculations applicable only to stainless steel in PWR water and consequently Higuchi et al. proposed in 2002 a revised model for stainless steel which incorporates new equations for evaluation of environmental fatigue reduction in BWR water. The paper compares the latest versions of these models and discusses the conservativeness of the models by comparison of the models with available test data.

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