In-reactor performance of pressure tubes in CANDU reactors

Rodgers, D.K.; Coleman, C.E.; Griffiths, M.; Bickel, G. A.; Theaker, JR; Muir, I.J.; Bahurmuz, A.A.; Lawrence, S. St.; Levi, M. Resta

doi:10.1016/j.jnucmat.2008.08.037

Cited by 41 publications

(15 citation statements)

References 5 publications

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“…The , E > 1 MeV. 3 OSIRIS is a research reactor at CEA (the French Alternative Energies and Atomic Energy Commission), Saclay, France. 4 ATR (Advanced Test Reactor) is a research reactor at the Idaho National Laboratory, Idaho, United States.…”

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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Section: Effect Of Texturementioning

confidence: 99%

Irradiation Effects on ZR-2.5NB in Power Reactors

Song

2016

CNL Nuclear Review

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“…Figure 5 [2] shows the effects of irradiation and fluence on the radial crack growth rate at 240°C [8,9]. The tests were performed on specimens containing nominally 100 ppm hydrogen at an applied K I of 17 MPa p m. The data show that irradiation increases the DHC growth rate by a factor of about 10 and the effect saturates after a fluence between 5 and 10 × 10 25 n·m −2 .…”

Section: Fracture Toughnessmentioning

confidence: 99%

“…Figure 4 [2] shows the effect of irradiation and fluence on K IH from cantilever beam specimens from CANDU reactor outlet rolled-joint region irradiated at about 290°C and tested at temperatures between 140°C and 240°C. K IH is weakly dependent on test temperature [14].…”

Section: Fracture Toughnessmentioning

confidence: 99%

“…In material from the outlet end, where the irradiation temperature was high, the beta-phase was highly FIGURE 6. The upper bound, mean, and lower bound axial DHC growth rates for irradiated Zr-2.5Nb pressure tubes [2]. decomposed.…”

Section: Dhc Mechanismmentioning

confidence: 99%

“…This paper is directed at reviewing the most recent results of this monitoring program for periods of operation up to the design life of the reactor and also describes how manufacturing has been modified to improve some aspects of pressure-tube performance. Performance targets described previously [1,2], are shown in Table 1. The pressure tube characteristics important for the performance of the reactor are: (i) fracture toughness, (ii) delayed hydridecracking (DHC), (iii) deformation, (iv) corrosion and hydrogen ingress, (v) manufacturing variability, and (vi) microstructure evolution during service.…”

Section: Introductionmentioning

confidence: 99%

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Performance of Pressure Tubes in Candu Reactors

et al. 2016

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The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815 °C, cold worked nominally 27%, and stress relieved at 400 °C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 °C and 310 °C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 × 1017 n·m−2·s−1 (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (KIH)) change with irradiation, but all reach a nearly limiting value at a fluence of less than 1025 n·m−2 (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases by about a factor of 6, while KIH is only slightly reduced. The role of microstructure and trace elements in these behaviours is described. Pressure tubes exhibit elongation, diametral expansion, and sag. The deformation behaviour is a function of operating conditions and the material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. The effect of material variability on corrosion behaviour is less well defined compared with other properties but there are instances where tube orientation and ingot source can be identified as factors that have an effect on hydrogen pick-up.

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Delayed Hydride Cracking: Theory and Experiment

Püls¹

2012

The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components

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As noted in Chaps. 1 and 9, zirconium alloy components can fail by a time-dependent mechanism of cracking if hydrides can preferentially form and subsequently crack at locations of elevated tensile stress. This process of time-dependent hydride cracking, called delayed hydride cracking (DHC), is based on the mechanism of diffusion of hydrogen to a region of elevated tensile stress followed by nucleation, growth, and fracture of the hydrided region. By repeating these steps, a crack can propagate in a component at a rate that, above a threshold stress intensity factor, K IH ; is mainly dependent on temperature. In this chapter we present the status of the present state of understanding of DHC in zirconium alloys, emphasizing the connections between the models developed and the experimental data. The intent of this chapter is not to provide an exhaustive review of the literature but to focus on results that are deemed to have collectively advanced our understanding of this phenomenon. This chapter is divided into two main parts, the first part dealing with DHC propagation, the other with DHC initiation. Although it may appear backwards to start with DHC propagation rather than initiation, this approach is consistent with the historical development of the field and thus also helps somewhat in simplifying the exposition. Readers wishing to obtain a succinct recent overview of DHC and other aspects of the effects of hydrogen and hydrides on the integrity of zirconium alloys are referred to a recent review by Coleman [15]. In addition, an earlier review by Northwood and Kosasih [47] provides a comprehensive summary account of work published in this field up to the date of that publication.M. P. Puls, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components, Engineering Materials,

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In-reactor performance of pressure tubes in CANDU reactors

Cited by 41 publications

References 5 publications

Irradiation Effects on ZR-2.5NB in Power Reactors

Irradiation Effects on ZR-2.5NB in Power Reactors

Performance of Pressure Tubes in Candu Reactors

Delayed Hydride Cracking: Theory and Experiment

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