Characterization of the stress corrosion cracking behavior of thermally sensitized 20Cr-25Ni stainless steel in a simulated cooling pond environment

Alshater, A. F.; Engelberg, Dirk; Lyon, S.B.; Donohoe, C.J.; Walters, S.; Whillock, G.O.H.; Sherry, Andrew H.

doi:10.1080/00223131.2017.1309305

Cited by 20 publications

(13 citation statements)

References 16 publications

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“…Although the stabilising effect of Nb has been studied in the past [16], the effect of these NbC precipitates on corrosion behaviour is not clear. Recent research on this material in open literature has focussed on irradiation effects [17] [18] [19], thermal processing [20] [21] [22], the use of novel corrosion techniques [23], evaluating the alloy in conditions relevant to a pond storage environment [24] [25] [26] [27] [28], dry storage [29] [30], and final geological disposal [31].…”

Section: Introductionmentioning

confidence: 99%

The role of niobium carbides in the localised corrosion initiation of 20Cr-25Ni-Nb advanced gas-cooled reactor fuel cladding

et al. 2020

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Section: Introductionmentioning

confidence: 99%

The role of niobium carbides in the localised corrosion initiation of 20Cr-25Ni-Nb advanced gas-cooled reactor fuel cladding

et al. 2020

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“…In AGR fuel cladding the niobium is present to effectively mop up carbon by preferentially forming niobium carbides thus preventing chromium carbide formation and chromium depletion. Consequently chromium depletion in AGR fuel cladding 95 is attributed to radiation induced segregation (RIS) [7]. RIS is the diffusion of an element, in this case chromium, against the concentration gradient when in the presence of radiation [8,9] and has been found to be most serious between 400 • C and 420 • C 100 [10].…”

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confidence: 99%

Vacuum drying of advanced gas reactor fuel

Goode

Harbottle

Hanson

2018

Progress in Nuclear Energy

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The UK will shortly cease reprocessing spent advanced gas reactor (AGR) fuel in favour of direct disposal, however since a permanent geological disposal facility is not envisaged as being available until 2075 interim storage will be required. The initial intention is to continue wet storage but it is possible that this may not be viable for as long as is hoped. Dry storage is commonly used worldwide for the interim storage of zirconium clad spent nuclear fuel however AGR fuel is stainless steel (SS) clad and as such a new safety case will be required to ensure that fuel can be adequately and safely dried.A rig has been designed to allow a comparison of the two main drying techniques in use; vacuum drying and flowed gas drying. This paper looks primarily at the design and development of the rig. Some of the initial data is presented to indicate how the rig was developed as a result of early results and is followed by some of the later test data to illustrate the improvements made.

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“…Additional information regarding the production route of AGR claddings is reported in Al‐Shater et al . (), Powell et al . (), and Barcellini et al .…”

Section: Methodsmentioning

confidence: 97%

“…The as-received material, whose composition is reported in Table 1, was in the form of a 100 × 50 mm 2 plate with a thickness of 0.5 mm and had been produced to meet the current cladding requirements for UK AGRs. Additional information regarding the production route of AGR claddings is reported in Al-Shater et al (2017), Powell et al (1985), and Barcellini et al (2018). In this work, samples with dimensions of 5×5 mm 2 were cut and heat treated for 1 h at selected temperatures up to 1100°C in inert atmosphere, and subsequently water quenched to room temperature.…”

Section: Methodsmentioning

confidence: 99%

“…The discharged fuel and the cladding are planned to be stored upon removal from the reactor core in caustic-dosed water ponds for at least 25 years (Whillock et al, 2018). A reduction in the local Cr content below 12wt.% would increase the susceptibility of this steel to intergranular corrosion attacks during medium-to-long term storage in aqueous environments ( Al-Shater et al, 2017;Whillock et al, 2018).…”

Section: Introductionmentioning

confidence: 99%

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Recrystallisation behaviour of a fully austenitic Nb‐stabilised stainless steel

Barcellini

Dumbill

Jimenez-Melero

2018

Journal of Microscopy

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Summary We have performed an in‐depth characterisation of the microstructure evolution of 20Cr‐25Ni Nb‐stabilised austenitic stainless steel during 1 h isochronal annealing up to 1100°C using scanning electron microscopy. This steel grade is used as cladding material in Advanced Gas‐cooled fission reactors, due to its resistance to thermal creep and oxidation. The initial deformed microstructure undergoes recrystallisation via a strain‐induced boundary migration mechanism, attaining a fully recrystallised microstructure at 850°C. A number of twins are observed in the vicinity of deformation bands prior to the start of recrystallisation. New Nb(C, N) particles form gradually in the microstructure, and the particle dispersion presents a maximum volume fraction of 2.7% at 930°C. At higher temperatures, the smaller particles become unstable and gradually dissolve in the matrix. Consequently, the Zener pinning pressure exerted on the grain boundaries is progressively released, triggering the growth of the austenite grains up to an average size of ∼47 μm at 1100°C. The observed temperature window for recrystallisation and grain growth can be predicted by a unified model based primarily on the migration of high‐ and low‐angle grain boundaries. Lay Description Austenitic stainless steel containing high percentage of chromium and nickel is currently used as fuel cladding material in the British Advanced Gas‐cooled Reactors (AGR). This material has been chosen because of its high resistance to thermal creep and corrosion, both enhanced by the presence of a fine dispersion of carbonitrides precipitated during the cladding thermomechanical processing. During the time spent in the reactor core, few fuel cladding elements can become susceptible to local chromium depletion at grain boundaries, which is ascribed to the time evolution of the microstructural damage caused by the neutron bombardment in the reactor core. This depletion might increase the susceptibility of this steel to intergranular corrosion attacks during medium‐to‐long term storage of spent fuel elements in water ponds. The severity of the local chromium depletion depends not only on the irradiation conditions, but also on the grain boundary geometry. We have investigated the recovery, recrystallisation and grain growth of AGR stainless steel during 1 h annealing at selected temperatures relevant for the thermomechanical processing of the steel claddings, focusing on the formation and evolution of grain boundaries and second phases. These two features play a key role in the progression of the neutron damage and the subsequent development of local chromium depletion during reactor service operations. A deep understanding of the mechanisms and conditions behind their formation during the thermomechanical processing of the cladding material and their interaction with each other constitutes the foundation to evaluate, and potentially mitigate, the effect of irradiation on the cladding material.

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Characterization of the stress corrosion cracking behavior of thermally sensitized 20Cr-25Ni stainless steel in a simulated cooling pond environment

Cited by 20 publications

References 16 publications

The role of niobium carbides in the localised corrosion initiation of 20Cr-25Ni-Nb advanced gas-cooled reactor fuel cladding

The role of niobium carbides in the localised corrosion initiation of 20Cr-25Ni-Nb advanced gas-cooled reactor fuel cladding

Vacuum drying of advanced gas reactor fuel

Recrystallisation behaviour of a fully austenitic Nb‐stabilised stainless steel

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