Effect of fast-neutron irradiation on mechanical properties of stainless steels: AISI types 304, 316 and 347

Higgy, H.R.; Hammad, F.H.

doi:10.1016/0022-3115(75)90151-8

Cited by 98 publications

(29 citation statements)

References 11 publications

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“…19. The curves are very similar to those reported for 304, 316, and 347 stainless steel irradiated elsewhere at ~50 o C to similar doses with larger specimens [15]. In all these stainless steels, there is no yield point drop in the unirradiated materials and in those irradiated to the lower fluences.…”

Section: (D)supporting

confidence: 83%

MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

Farrell¹

2003

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Section: (D)supporting

confidence: 83%

MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

Farrell¹

2003

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“…Further, the data from Higgy and Hammad [65] suggest that mechanical property measurements on 304 and 316 SS after neutron irradiation at <100°C suggest that defect cluster saturation occurs before 0.1 dpa, consistent with the results of this study. The model developed by Busby has also been used to simulate the annealing of samples irradiated at low temperature.…”

Section: Isolating the Dislocation Microstructuresupporting

confidence: 89%

Mechanism of Irradiation Assisted Cracking of Core Components in Light Water Reactors

Was¹,

Atzmon²

2003

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Executive SummaryThe overall goal of the project is to determine the mechanism of irradiation assisted stress corrosion cracking (IASCC). IASCC has been linked to hardening, microstructural and microchemical changes during irradiation. Unfortunately, all of these changes occur simultaneously and at similar rates during irradiation, making attribution of IASCC to any one of these features nearly impossible to determine. The strategy set forth in this project is to develop means to separate microstructural from microchemical changes to evaluate each separately for their effect on IASCC. In the first part, post irradiation annealing (PIA) treatments are used to anneal the irradiated microstructure, leaving only radiation induced segregation (RIS) for evaluation for its contribution to IASCC. The second part of the strategy is to use low temperature irradiation to produce a radiation damage dislocation loop microstructure without radiation induced segregation in order to evaluate the effect of the dislocation microstructure alone.A radiation annealing model was developed based on the elimination of dislocation loops by vacancy absorption. Results showed that there were indeed, time-temperature annealing combinations that leave the radiation induced segregation profile largely unaltered while the dislocation microstructure is significantly reduced. Proton irradiation of 304 stainless steel irradiated with 3.2 MeV protons to 1.0 or 2.5 dpa resulted in grain boundary depletion of chromium and enrichment of nickel and a radiation damaged microstructure. Post irradiation annealing at temperatures of 500 -600°C for times of up to 45 min. removed the dislocation microstructure to a greater degree with increasing temperatures, or times at temperature, while leaving the radiation induced segregation profile relatively unaltered. Constant extension rate tensile (CERT) experiments in 288°C water containing 2 ppm O 2 and with a conductivity of 0.2 mS/cm and at a strain rate of 3 x 10 -7 s -1 showed that the IASCC susceptibility, as measured by the crack length per unit strain, decreased with very short anneals and was almost completely removed by an anneal at 500°C for 45 min. This annealing treatment removed about 15% of the dislocation microstructure and the irradiation hardening, but did not affect the grain boundary chromium depletion or nickel segregation, nor did it affect the grain boundary content of other minor impurities. These results indicate that RIS is not the sole controlling feature of IASCC in irradiated stainless steels in normal water chemistry. DE-FG07-99ID13768Final Report 3The isolation of the irradiated microstructure was approached using low temperature irradiation or combinations of low and high temperature irradiations to achieve a stable, irradiated microstructure without RIS. Experiments were successful in achieving a high degree of irradiation hardening without any evidence of RIS of either major or minor elements. The low temperature irradiations to doses up to 0.3 dpa at T<75°C were also very succ...

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“…Austenitic stainless steels irradiated at low temperatures (< 300°C) show "saturation" of radiation hardening or hardening which is regarded as insensitive to exposure [9]. Prior to saturation, the change in yield strength is linear with (fast fluence) ≈1/2 as shown by Higgy and Hammad for Type 304, 316, and 347 stainless steels in a low temperature (< 100°C) irradiation performed in the High Flux Reactor at Petten [10]. Saturation of hardening at a fast fluence of 4 x 10 19 n/cm 2 (E n > 1 MeV) was identified [10].…”

Section: Effect Of Microstructure On Mechanical Responsementioning

confidence: 99%

Development and Application of Materials Properties for Flaw Stability Analysis in Extreme Environment Service

Sindelar

Lam

Duncan

et al. 2007

Volume 6: Materials and Fabrication

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Discovery of aging phenomena in the materials of a structure may arise after its design and construction that impact its structural integrity. This condition can be addressed through a demonstration of integrity with the material-specific degraded conditions. Two case studies of development of fracture and crack growth property data, and their application in development of in-service inspection programs for nuclear structures in the defense complex are presented. The first case study covers the development of fracture toughness properties in the form of J-R curves for rolled plate Type 304 stainless steel with Type 308 stainless steel filler in the application to demonstrate the integrity of the reactor tanks of the heavy water production reactors at the Savannah River Site. The fracture properties for the base, weld, and heat-affected zone of the weldments irradiated at low temperatures (110°-150°C) up to 6.4 dpa NRT and 275 appm helium were developed. An expert group provided consensus for application of the irradiated properties for material input to acceptance criteria for ultrasonic examination of the reactor tanks. Dr. Spencer H. Bush played a lead advisory role in this work. The second case study covers the development of fracture toughness for A285 carbon steel in high level radioactive waste tanks. The approach in this case study incorporated a statistical experimental design for material testing to address metallurgical factors important to fracture toughness. Tolerance intervals were constructed to identify the lower bound fracture toughness for material input to flaw disposition through acceptance by analysis.

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Effect of fast-neutron irradiation on mechanical properties of stainless steels: AISI types 304, 316 and 347

Cited by 98 publications

References 11 publications

MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

MAPPING FLOW LOCALIZATION PROCESSES IN DEFORMATION OF IRRADIATED REACTOR STRUCTURAL ALLOYS - FINAL REPORT. Nuclear Energy Research Initiative Program No. MSF99-0072. Period: August 1999 through September 2002. (ORNL/TM-2003/63)

Mechanism of Irradiation Assisted Cracking of Core Components in Light Water Reactors

Development and Application of Materials Properties for Flaw Stability Analysis in Extreme Environment Service

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