Volume conservation during irradiation growth of Zr–2.5Nb

Holt, R.A.; Causey, A.R.

doi:10.1016/j.jnucmat.2004.07.042

Cited by 18 publications

(10 citation statements)

References 4 publications

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“…Measurements of the growth of compact toughness specimens in three dimensions at about 520 K, has confirmed that there is negligible volume change, so that the strain rates in all three directions evolve with fast fluence, Fig. 5.1.3 [59], and the strain in the thickness direction can be inferred to be the negative sum of the strains in the axial and transverse directions.…”

Section: Effect Of Fast Neutron Fluencementioning

confidence: 87%

“…The implication of a linear fast flux dependence is that growth Growth strains in three dimensions as a function of fast neutron fluence for Zr-2.5Nb pressure tube material irradiated at 520 K in Osiris at $2 · 10 18 n m À2 s À1 , E > 1 MeV, from Ref. [59]. Note several points in the range 9-13 · 10 25 n m À2 are thought to be erroneous.…”

Section: Effect Of Fast Neutron Fluxmentioning

confidence: 99%

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In-reactor deformation of cold-worked Zr–2.5Nb pressure tubes

Holt

2008

Journal of Nuclear Materials

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203

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Section: Effect Of Fast Neutron Fluencementioning

confidence: 87%

Section: Effect Of Fast Neutron Fluxmentioning

confidence: 99%

In-reactor deformation of cold-worked Zr–2.5Nb pressure tubes

Holt

2008

Journal of Nuclear Materials

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203

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“…Holt and Causey [24] conducted an irradiation growth test on cold-worked Zr-2.5Nb CANDU pressure tube materials irradiated to a fast fluence of 17.4 × 10 25 n·m −2 , E > 1 MeV, at a nominal temperature of 523 K. The dimensional changes in 3 directions were measured, and a small volume increase (0.05-0.1%) was observed initially, then gradually decayed to <0.05%. Within the accuracy of the measurement, the long-term rate of volume change is negligible.…”

Section: Effect Of Texturementioning

confidence: 99%

Irradiation Effects on ZR-2.5NB in Power Reactors

Song

2016

CNL Nuclear Review

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Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, “High Power Channel-type Reactor”) reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr-2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge.

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“…A = 1.2 × 10 −21 N/cm 2 , n = 0.794. Due to the volume conserving of the irradiation growth of the cladding [26], the growth in the other two directions was assumed to be consistent; the radial strain ε r is given by…”

Section: Thermomechanical Models Of Zr Claddingmentioning

confidence: 99%

Coupled Thermomechanical Responses of Zirconium Alloy System Claddings under Neutron Irradiation

et al. 2021

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Zirconium (Zr) alloy is a promising fuel cladding material used widely in nuclear reactors. Usually, it is in service for a long time under the effects of neutron radiation with high temperature and high pressure, which results in thermomechanical coupling behavior during the service process. Focusing on the UO2/Zr fuel elements, the macroscopic thermomechanical coupling responses of pure Zr, Zr-Sn, and Zr-Nb binary system alloys, as well as Zr-Sn-Nb ternary system alloy as cladding materials, were studied under neutron irradiation. As a heat source, the thermal conductivity and thermal expansion coefficient models of the UO2 core were established, and an irradiation growth model of a pure Zr and Zr alloy multisystem was built. Based on the user material subroutine (UMAT) with ABAQUS, the current theoretical model was implemented into the finite element framework, and the consequent thermomechanical coupling behavior under irradiation was calculated. The distribution of temperature, the stress field of the fuel cladding, and their evolution over time were analyzed. It was found that the stress and displacement of the cladding were sensitive to alloying elements due to irradiated growth.

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Volume conservation during irradiation growth of Zr–2.5Nb

Cited by 18 publications

References 4 publications

In-reactor deformation of cold-worked Zr–2.5Nb pressure tubes

In-reactor deformation of cold-worked Zr–2.5Nb pressure tubes

Irradiation Effects on ZR-2.5NB in Power Reactors

Coupled Thermomechanical Responses of Zirconium Alloy System Claddings under Neutron Irradiation

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