Radiation Effects in Reactor Structural Alloys

Mansur, L.K.; Bloom, E.E.

doi:10.1007/bf03338114

Cited by 15 publications

(13 citation statements)

References 15 publications

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“…The evolution of local chemistry and structure at these scales is responsible for changes in physical and mechanical properties [10][11][12]. Spatial and temporal correlations associated with the displacement cascades continue to play an important role over much larger scales, as do processes including defect recombination, clustering, migration, and gas and solute diffusion and trapping.…”

Section: Introductionmentioning

confidence: 99%

Multiscale modeling of radiation damage in Fe-based alloys in the fusion environment

Wirth

Odette

Marian

et al. 2004

Journal of Nuclear Materials

108

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Section: Introductionmentioning

confidence: 99%

Multiscale modeling of radiation damage in Fe-based alloys in the fusion environment

Wirth

Odette

Marian

et al. 2004

Journal of Nuclear Materials

108

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“…It is well established that neutron radiation damage to structural and cladding materials in current nuclear environments (Generation I and II reactor systems) produces significant mechanical property degradation. [1][2][3][4][5][6][7][8][9][10][11] However, despite the extensive effort which has produced a significant body of knowledge regarding the macroscopic degradation, we lack the physically-based models that are capable of predicting this degradation and materials failure. This is particular apparent as we attempt to predict the material response in the elastic-plastic regime.…”

Section: Background and Research Approachmentioning

confidence: 99%

Dislocation-Radiation Obstacle Interactions: Developing Improved Mechanical Property Constitutive Models

Wirth¹,

Robertson²

2007

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“…Figure 4.35 shows microstructures for unirradiated material and for irradiated stainless steels at three different temperatures (Mansur and Bloom 1982). For a given dose and irradiation temperature, specific irradiation-induced defects are responsible for such changes in properties.…”

mentioning

confidence: 99%

“…a) microstructure of material before irradiation, b) dislocation loops formed during irradiation at 300ºC (≈0.35 T m ), c) dislocation loops, voids, dislocation networks and precipitates formed during irradiation at about 500ºC (≈0.47 T m ), d) helium bubbles on a grain boundary formed during irradiation at about 700 ºC (≈0.59 T m ). After Mansur and Bloom (1982).…”

mentioning

confidence: 99%

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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Radiation Effects in Reactor Structural Alloys

Abstract: SUMMARY

Cited by 15 publications

References 15 publications

Multiscale modeling of radiation damage in Fe-based alloys in the fusion environment

Multiscale modeling of radiation damage in Fe-based alloys in the fusion environment

Dislocation-Radiation Obstacle Interactions: Developing Improved Mechanical Property Constitutive Models

Code qualification of structural materials for AFCI advanced recycling reactors.

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