30. Analysis of irradiation-induced creep of stainless steel in fast spectrum reactors

Foster, Jann P; Wolfer, W.G.; Biancheria, A.; Boltax, A.

doi:10.1680/ieacifcacc.48748.0034

Cited by 17 publications

(4 citation statements)

References 8 publications

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“…At low stresses (<120 MPa in PWRs), stainless steel components exhibit irradiation creep that is enhanced by irradiation and is linear with stress. In fast reactors at high temperatures, above 400 °C, when the swelling is significant, the steady-state irradiation creep can be described by a simple empirical law [ 79 ], based on work by Foster et al [ 80 ],

where

is the steady-state irradiation creep rate compliance in the absence of swelling in units of dpa −1 MPa −1 , D is the irradiation creep-swelling compliance in units of MPa −1 and

is the linear swelling rate in units of dpa −1 . The compliance (

relates the equivalent strain

to the equivalent stress (

[ 79 ] as defined by,

where

and

are the stress and strain rates in each of three orthogonal directions (d) referred to the principal axes of stress.…”

Section: Irradiation Creepmentioning

confidence: 99%

Effect of Neutron Irradiation on the Mechanical Properties, Swelling and Creep of Austenitic Stainless Steels

Griffiths

2021

Materials

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Austenitic stainless steels are used for core internal structures in sodium-cooled fast reactors (SFRs) and light-water reactors (LWRs) because of their high strength and retained toughness after irradiation (up to 80 dpa in LWRs), unlike ferritic steels that are embrittled at low doses (<1 dpa). For fast reactors, operating temperatures vary from 400 to 550 °C for the internal structures and up to 650 °C for the fuel cladding. The internal structures of the LWRs operate at temperatures between approximately 270 and 320 °C although some parts can be hotter (more than 400 °C) because of localised nuclear heating. The ongoing operability relies on being able to understand and predict how the mechanical properties and dimensional stability change over extended periods of operation. Test reactor irradiations and power reactor operating experience over more than 50 years has resulted in the accumulation of a large amount of data from which one can assess the effects of irradiation on the properties of austenitic stainless steels. The effect of irradiation on the intrinsic mechanical properties (strength, ductility, toughness, etc.) and dimensional stability derived from in- and out-reactor (post-irradiation) measurements and tests will be described and discussed. The main observations will be assessed using radiation damage and gas production models. Rate theory models will be used to show how the microstructural changes during irradiation affect mechanical properties and dimensional stability.

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where

is the steady-state irradiation creep rate compliance in the absence of swelling in units of dpa −1 MPa −1 , D is the irradiation creep-swelling compliance in units of MPa −1 and

is the linear swelling rate in units of dpa −1 . The compliance (

relates the equivalent strain

to the equivalent stress (

[ 79 ] as defined by,

where

and

are the stress and strain rates in each of three orthogonal directions (d) referred to the principal axes of stress.…”

Section: Irradiation Creepmentioning

confidence: 99%

Effect of Neutron Irradiation on the Mechanical Properties, Swelling and Creep of Austenitic Stainless Steels

Griffiths

2021

Materials

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“…It has been consistently shown that the irradiationcreep rate is the sum of two terms: one is independent of the swelling and the other is strongly coupled with the swelling rate. The major components of the instantaneous creep rate can be expressed by the following empirical equation [8,9]: whereε is an instantaneous irradiation-creep rate, σ is an effective stress, B 0 is the creep compliance, D is the irradiation creep -a swelling coupling coefficient -anḋ S is the instantaneous volumetric swelling rate per dpa. Irradiation creep has been studied by the irradiation of pressurized tubes.…”

Section: Resultsmentioning

confidence: 99%

Change in shape and size of fuel assembly duct in BN-350 reactor

Sarsekeyev

Chumakov

Mendebayev

2013

Journal of Nuclear Science and Technology

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The paper presents analytical data obtained by directly measuring samples taken at different levels of a hexagonal fuel assembly duct used in the BN-350 fast reactor. The difference between the length of the deflection arc and the width of an intact duct face displays the accumulated irradiation-creep deformation during the operation period. The deflection of the duct faces induced by the irradiation creep was analyzed, and irradiation-creep rate versus distance from the core center was analyzed for the assembly duct in the BN-350 fast reactor.

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“…At low stresses (<150 MPa), austenitic stainless steel and Ni-alloy components exhibit irradiation creep that is enhanced by irradiation and is linear with stress [ 5 , 6 ]. In fast reactors at high temperatures (above 400 °C), when the swelling is significant, the steady-state irradiation creep can be described by a simple empirical law relating the effective strain rate,

, to the effective stress,

[ 20 ], based on work by Foster et al [ 21 ],

where

is the steady state irradiation creep rate compliance in the absence of swelling in units of dpa −1 .MPa −1 , D is the irradiation creep-swelling compliance in units of MPa −1 and

is the linear swelling rate, which is independent of stress, in units of dpa −1 . There are no separate thermal and irradiation creep components to

, and the term

is dependent on diffusional mass transport.…”

Section: Irradiation Creep In Nuclear Reactor Coresmentioning

confidence: 99%

Microstructural Effects on Irradiation Creep of Reactor Core Materials

Griffiths

2023

Materials

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The processes that control irradiation creep are dependent on the temperature and the rate of production of freely migrating point defects, affecting both the microstructure and the mechanisms of mass transport. Because of the experimental difficulties in studying irradiation creep, many different hypothetical models have been developed that either favour a dislocation slip or a mass transport mechanism. Irradiation creep mechanisms and models that are dependent on the microstructure, which are either fully or partially mechanistic in nature, are described and discussed in terms of their ability to account for the in-reactor creep behaviour of various nuclear reactor core materials. A rate theory model for creep of Zr-2.5Nb pressure tubing in CANDU reactors incorporating the as-fabricated microstructure has been developed that gives good agreement with measurements for tubes manufactured by different fabrication routes having very different microstructures. One can therefore conclude that for Zr-alloys at temperatures < 300 °C and stresses < 150 MPa, diffusional mass transport is the dominant creep mechanism. The most important microstructural parameter controlling irradiation creep for these conditions is the grain structure. Austenitic alloys follow similar microstructural dependencies as Zr-alloys, but up to higher temperature and stress ranges. The exception is that dislocation slip is dominant in austenitic alloys at temperatures < 100 °C because there are few barriers to dislocation slip at these low temperatures, which is linked to the enhanced recombination of irradiation-induced point defects.

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30. Analysis of irradiation-induced creep of stainless steel in fast spectrum reactors

Cited by 17 publications

References 8 publications

Effect of Neutron Irradiation on the Mechanical Properties, Swelling and Creep of Austenitic Stainless Steels

Effect of Neutron Irradiation on the Mechanical Properties, Swelling and Creep of Austenitic Stainless Steels

Change in shape and size of fuel assembly duct in BN-350 reactor

Microstructural Effects on Irradiation Creep of Reactor Core Materials

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