A review of irradiation assisted stress corrosion cracking

Scott, Peter

doi:10.1016/0022-3115(94)90360-3

Cited by 247 publications

(75 citation statements)

References 16 publications

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“…3) [8][9][10][11]. These cracks usually attributed to IASCC (Irradiation Assisted Stress Corrosion Cracking), can be seen as a consequence of the evolution of plasticity in these materials loaded in a corrosive medium, together with possible evolutions of grain boundary chemistry (due to RIS-Radiation Induced Segregation) [12]. Since evolutions of the macroscopic properties are related to the microstructure ones, understanding microstructure evolution under irradiation is essential to predict lifetime of internals.…”

Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

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Abstract:Even if temperature, pressure and chemistry of the cooling water are not very high and aggressive, materials used in PWRs (Pressurized Water Reactors) are exposed to different degradation mechanisms. One of the main goals of the research programs in this field is to develop physical model of the mechanisms down to the atomic scale. Such approach needs a clear description and understanding of the degradation mechanisms at the same small scale. This paper illustrates the benefits of different microscopies and of their last improvements up to the promising possibilities of monochromated and aberrations corrected TEM/STEM. A specific focus is placed on four different degradation mechanisms observed in austenitic stainless steel: irradiation ageing, corrosion fatigue, stress corrosion cracking and corrosion.

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Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

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“…Over the past decade, there have been significant advances in the understanding of degradation mechanisms in austenitic stainless steel reactor internals as summarized in several review papers [1,2]. Based substantially on this advance in knowledge, efforts to develop austenitic stainless steel structural materials with enhanced radiation resistance have been undertaken.…”

Section: Introductionmentioning

confidence: 99%

The Influence of Pre-Irradiation Heat Treatments on Thermal Non-Equilibrium and Radiation-Induced Segregation Behavior in Model Austenitic Stainless Steel Alloys

Cole

Allen

Was

et al. 2004

Effects of Radiation on Materials: 21st International Symposium

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Abstract:The effect of pre-irradiation heat treatments on thermal non-equilibrium grain boundary segregation (TNES) and subsequent radiation-induced grain boundary segregation (RIS) is studied in a series of model austenitic stainless steels. The alloys used for this study are based on AISI 316 stainless steel and have the following nominal compositions: Fe-16Cr-13Ni-1.25Mn (base 316), Fe-16Cr-13Ni-1.25Mn-2.0Mo (316 + Mo) and Fe-16Cr-13Ni-1.25Mn-2.0Mo-0.07P (316 + Mo + P). Samples were heat treated at temperatures ranging from 1100 to 1300°C and cooled at 4 different rates (salt brine quench, water quench, air cool and furnace cool) to evaluate the effect of annealing temperature and quench rate on TNES. The alloys were than processed with the treatment (temperature and cooling rate) that resulted in the maximum Cr enrichment. Alloys with and without the heat treatment to enrich the grain boundaries with Cr were characterized following irradiation to 1 dpa at 400°C with high-energy protons in order to understand the influence of alloying additions and pre-irradiation grain boundary chemistry on irradiation-induced elemental enrichment and depletion profiles. Various mechanistic models will be examined to explain the observed behavior.

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“…[8][9][10][11] Under the bombardment of energetic neutrons, a large number of lattice atoms are unlocked from their original sites, generating vacancies and self-interstitials in crystalline materials. While most of the vacancies and interstitials quickly recombine after an initial collision phase, the remaining point defects evolve subsequently into metastable irradiation defects, giving rise to various irradiation effects, such as hardening, embrittlement, enhanced diffusion and segregation, void swelling, creep, etc.…”

Section: Bwrmentioning

confidence: 99%

“…Chromium depletion at grain boundaries resulting from radiation-induced segregation is a possible mechanism to explain the increased susceptibility to intergranular SCC in SSs. 7,9,10 A highly localized plastic flow caused by small irradiation defects is also considered a critical factor of reduced cracking resistance after irradiation. 11,12 Without doubt, irradiation damage plays a key role in elevating the cracking susceptibility of non-sensitized SSs that would otherwise show excellent SCC performance in high-temperature water.…”

Section: Bwrmentioning

confidence: 99%

Technical Letter Report on the Cracking of Irradiated Stainless Steels in Low-Corrosion-Potential Environments

Chen¹,

Alexandreanu²,

Natesan³

2013

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Irradiation-assisted stress corrosion cracking (IASCC) is a major degradation mechanism for reactor internal components of austenitic stainless steels (SSs) in light water reactors. Crack growth rate (CGR) and fracture toughness J-R curve tests were performed on four irradiated SS specimens in high-purity water with low dissolved oxygen and in a simulated pressurized water reactor (PWR) environment. The samples had been previously irradiated from ~5 to 8 dpa in the BOR-60 reactor. The materials were cold-worked (CW) 316 and solution-annealed (SA) 304L SSs, which are commonly used in PWR core internals. Cyclic and constant-load CGRs were obtained from these samples to assess their cracking behaviors in low-corrosion-potential environments. A Ti-stabilized specimen, CW 316-Ti SS, was found to be highly vulnerable to IASCC, and its cracking susceptibility increased with neutron dose between ~5 and 8 dpa. The CW 316 and SA 304L SSs showed moderate cracking susceptibilities in the low-corrosionpotential environments up to ~7 dpa. Periodic partial unloading (PPU) had a significant impact on the crack growth behavior of all irradiated specimens, and the measured CGRs were more than one order of magnitude higher in the test periods with than without PPU. A "stair-step" crack growth behavior was also observed for all irradiated specimens during the test periods with PPU, suggesting a strong effect of dynamic loading on IASCC. The SA 304L SS showed the best fracture resistance, and the CW 316 SS was the most brittle among the tested specimens. The fracture toughness J values of the two CW 316-Ti SS specimens were nearly identical at ~5 and ~8 dpa.

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A review of irradiation assisted stress corrosion cracking

Cited by 247 publications

References 16 publications

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

The Influence of Pre-Irradiation Heat Treatments on Thermal Non-Equilibrium and Radiation-Induced Segregation Behavior in Model Austenitic Stainless Steel Alloys

Technical Letter Report on the Cracking of Irradiated Stainless Steels in Low-Corrosion-Potential Environments

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