Comparison of the in-reactor creep of selected ferritic, solid solution strengthend, and precipitation hardened commercial alloys

Paxton, M.M.; Chin, Bryan A.; Gilbert, E.R.; Nygren, R.E.

doi:10.1016/0022-3115(79)90230-7

Cited by 52 publications

(4 citation statements)

References 6 publications

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“…From standard rate theory calculations, a high sink strength would be predicted to induce lower irradiation creep rates due to suppressed point defect supersaturation levels. Paxton et al reported reduced irradiation creep in austenitic alloys with high precipitate densities [325], whereas other austenitic steel studies have not observed a significant effect of sink strength [206,325]. The reported steady-state irradiation creep compliance for MA957 ODS ferritic steel was comparable or slightly higher than the typical values for standard FM steel [333].…”

Section: Irradiation Creep and Creep-fatigue Interactionmentioning

confidence: 85%

“…Collectively, creep and swelling can induce significant dose-dependent distortion of fusion reactor structures that experience variations in dose rate, temperature or stress state [319]. Irradiation creep is one of the earliest reported radiation degradation phenomena in materials [320], and phenomenologically is proportional to the displacement damage dose and the applied stress with a weak temperature dependence between ∼0.25 and 0.5T m [321][322][323][324][325]. The magnitude of steady-state irradiation creep can be empirically described by the following equation [326]:…”

Section: Irradiation Creep and Creep-fatigue Interactionmentioning

confidence: 99%

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Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

et al. 2022

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Reduced activation ferritic-martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. Performance of these steels will deteriorate in-service due to neutron damage and transmutation-induced gases like helium/hydrogen at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the major irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (i) low temperature hardening-embrittlement (LTHE) - including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile to brittle transition temperature and fracture toughness, (ii) inter-mediate temperature cavity swelling, (iii) the effect of helium on LTHE and cavity swelling, (iv) irradiation creep and (v) tritium management issues are reviewed. The potential causes of LTHE are discussed that highlights need for advanced characterization techniques. Mechanical properties including tensile/Charpy impact of RAFM and ODS steels are compared to show that current generation of ODS steels also suffer from LTHE, and they can show irradiation hardening up to high temperatures, ~400-500 °C. To minimize this, future ODS steel development for FW/B specific application should target materials with lower Cr concentration (to minimize α’) and minimize other elements that could form embrittling phases under irradiation. RAFM steel designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand synergistic effects of helium on thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex including stresses due to magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multi-scale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility

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Section: Irradiation Creep and Creep-fatigue Interactionmentioning

confidence: 85%

Section: Irradiation Creep and Creep-fatigue Interactionmentioning

confidence: 99%

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

et al. 2022

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“…This alloy has shown great promise, especially with respect to its resistance to swelling and excellent compatibility with sodium. With one exception [21] all reported swelling and creep data on this alloy have been derived from very simple, free-standing experimental specimens irradiated under rather well-defined conditions in material test assemblies [17,[22][23][24][25][26][27][28].…”

Section: Introductionmentioning

confidence: 99%

Microstructural stability of an HT-9 fuel assembly duct irradiated in FFTF

Sencer

Kennedy

Cole

et al. 2011

Journal of Nuclear Materials

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“…Further, the presence of the molybdenum (Mo) improves the corrosion resistance [2]. The HT-9 steel shows resistance to swelling [2] and irradiation creep [3,4] in neutron environments, and eventually become an attractive candidate alloy for fusion reactor structures for their apparent resistance to radiation damage. However, irradiation produces an increase in the ductile-brittle transition temperature [5] which could result in failures during off normal events.…”

Section: Introductionmentioning

confidence: 99%

Annealing Behaviour of Ferritic HT-9 Steel

Sahoo¹,

Gaffney

Chatterjee

et al. 2011

MSF

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Annealing behaviour of ferritic HT-9 stainless steel plates was investigated through evolution of microstructural, textural and mechanical properties. Microstructural observations showed two important findings: (a) abnormal grain growth occurs at higher annealing temperatures, (b) drop in in-grain misorientation with increasing annealing temperature. However, textural results showed an increase in texturing till 500oC annealing and a clear drop in texturing beyond 500oC annealing. The strength (both yield and tensile) decreased with increase in testing temperature and the micro-hardness values decreased with increase in annealing temperature. Dynamic strain ageing was observed at certain strain rates and testing temperatures. The onset of the serrated yielding at low temperature limit estimated from the strain rate-temperature plot was found out to be ~ 94kJ/mol identifiable with that for migration of C and N.

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Comparison of the in-reactor creep of selected ferritic, solid solution strengthend, and precipitation hardened commercial alloys

Cited by 52 publications

References 6 publications

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Microstructural stability of an HT-9 fuel assembly duct irradiated in FFTF

Annealing Behaviour of Ferritic HT-9 Steel

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