The microstructural origins of yield strength changes in aisi 316 during fission or fusion irradiation

Garner, F.А.; Hamilton, M.L.; Panayotou, N.F.; Johnson, George

doi:10.1016/0022-3115(82)90698-5

Cited by 126 publications

(26 citation statements)

References 13 publications

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“…For annealed Type 316 stainless steel, yield strength increased from 165 MPa in the unirradiated condition to 544 and 644 MPa after exposures to 10 and 16 dpa, respectively. The yield strength value after 16 dpa is somewhat close to those reported by Garner et al (1981) in Figure 4.41, but September 15, 2008 51 the increase in yield strength with dpa exposure was much slower. The tensile results reported by Mills (1986) for Type 304 base metal and type 308 weld metal were similar to those of the type 316, even though the unirradiated yield strength of the 308 weld metal was 312 MPa.…”

Section: Code Qualification Of Structural Materials For Afci Advancedsupporting

confidence: 73%

“…It can be seen that the irradiated values are in the range 2-4 times higher than the unirradiated values and that hardening in the irradiated materials shows higher values near 300ºC. Figure 4.41 shows data from Garner et al (1981) for Type 316 stainless steel in both the annealed and 20% cold-worked conditions. At an irradiation temperature of 538°C, the yield strengths of the material in both conditions appears to stabilize at about 400 MPa and at about 2x10 22 n/cm 2 , with the cold worked material exhibiting softening and the annealed material hardening.…”

Section: Code Qualification Of Structural Materials For Afci Advancedmentioning

confidence: 99%

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Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

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This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package "Advanced Materials Code Qualification".The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors.Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue in...

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Section: Code Qualification Of Structural Materials For Afci Advancedsupporting

confidence: 73%

Section: Code Qualification Of Structural Materials For Afci Advancedmentioning

confidence: 99%

Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

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show abstract

“…The unirradiated yield strength for 20% cold-worked Type 316 stainless steel tested at 370°C used in calculating the data in Figure 9 is 580MPa [1]. The estimated dislocation density for 20% cold-worked Type 316 stainless steel is 3x10 15 cm -2 [16]. Figure 9 indicates that the changes in measured yield strength generally track the changes in yield strength calculated from the microstructure using the inputs from Table 23.…”

Section: Discussionmentioning

confidence: 96%

“…The values of M and b used to calculate the yield strength increment are taken from Reference [16] and are also listed in Table 5. The unirradiated yield strength for 20% cold-worked Type 316 stainless steel tested at 370°C used in calculating the data in Figure 9 is 580MPa [1].…”

Section: Discussionmentioning

confidence: 99%

Properties of 20 % Cold-Worked 316 Stainless Steel Irradiated at Low Dose Rate

Allen

Tsai

Cole

et al. 2004

Effects of Radiation on Materials: 21st International Symposium

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Abstract:To assess the effects of long-term, low-dose-rate neutron exposure, tensile, hardness, and fracture properties were measured and microstructural characterization performed on irradiated 20% cold-worked Type 316 stainless steel. Samples were prepared from reactor core components retrieved from the EBR-II reactor following final shutdown. Sample locations were chosen to cover a dose range of 1-56 dpa at temperatures from 371-390°C and dose rates from 0.8-3.3 x10 -7 dpa/s. Irradiation caused hardening, with the ultimate tensile strength (UTS) reaching about 800 MPa near 20 dpa and appearing to saturate at higher doses. The yield strength (YS) follows approximately the same trend as the ultimate tensile strength. At higher dose, the difference between the UTS and YS decreases, suggesting the work-hardening capability of the material is decreasing with increasing dose. The hardness and yield strength increases occur roughly over the same range of dose. While the material retained respectable ductility at 20 dpa, the uniform and total elongation decreased to <1 and <3%, respectively, at 47 dpa. Fracture in the 30 dpa specimen is mainly ductile but with local regions of mixed-mode failure, consisting mainly of dimples and microvoids. The fracture surface of the higher-exposure 47 dpa specimen displays more brittle features. Changes in yield strength predicted from the microstructural components are roughly consistent with the measured changes in yield strength.

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“…The irradiation swelling resistance of silicon may be due to the fast diffusing mechanism [9,10] . When materials are suffered from irradiation, the swelling could be divided into three stages, that is, incubation, steady-state and saturation regime.…”

Section: Fig 3 Microstructures Of Fesi After Deuterium Implatation Amentioning

confidence: 99%

Effect of Silicon on microstructures of iron after high-temperature deuterium implantation

Jiang¹

2017

Proceedings of the 2017 6th International Conference on Measurement, Instrumentation and Automation (ICMIA 2017)

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Abstract:To study the effect of silicon on microstructures in CLAM (China Low Activation Martensitic) steel after irradiation, the microstructures of Fe and FeSi model alloys were investigated after implanted deuterium ions at 773K. TEM (Transmission Electron Microscope) observation and EDX (Energy Dispersive X-ray Spectrum) analysis have been carried out. The results showed that tiny voids were observed in Fe after implantation, while no void observed in FeSi model alloys. Instead, there were some carbides which indicated silicon played an important role in the improved irradiation resistance. Fast diffusing mechanism was adopted to analyze the effect of silicon during irradiation.

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The microstructural origins of yield strength changes in aisi 316 during fission or fusion irradiation

Cited by 126 publications

References 13 publications

Code qualification of structural materials for AFCI advanced recycling reactors.

Code qualification of structural materials for AFCI advanced recycling reactors.

Properties of 20 % Cold-Worked 316 Stainless Steel Irradiated at Low Dose Rate

Effect of Silicon on microstructures of iron after high-temperature deuterium implantation

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