Pressure Tubes In CANDU Reactors Research Articles

A Model for the Temperature Distribution in a Rolled Joint in a CANDU Reactor Exploiting the Decomposition of the β-Zr Phase

A competing-rates model is presented to account for operational changes in the metastable β-Zr phase of the Zr-2.5Nb alloy used to make CANDU reactor pressure tubes and is used to predict temperature gradients at the outlet rolled joints using the decomposition of the β-Zr phase as a proxy for temperature. High temperatures decompose the β phase by enhancing the formation of small particles of ω and α phases. Fast neutron flux causes the ω and α phases to shrink. This process is assumed to depend on the total volume of the particles, because they are comparable to, or smaller than, the size of the neutron displacement cascades. The barrier energy for thermal growth was determined to be 2.43 eV, when an Arrhenius A factor of 1013/s was assumed. The cross section for (ω+α)-phase shrinkage is 24.5 barns for Zr-2.5Nb irradiated in CANDU reactors. Assuming that the shrinkage is dominated by the migration of self-interstitial point defects, a defect production efficiency of 1.4% was found.

PERFORMANCE OF PRESSURE TUBES IN CANDU REACTORS

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.

Study of the in vitro corrosion behavior and biocompatibility of Zr-2.5Nb and Zr-1.5Nb-1Ta (at%) crystalline alloys

The in vitro corrosion behavior and biocompatibility of two Zr alloys, Zr-2.5Nb, employed for the manufacture of CANDU reactor pressure tubes, and Zr-1.5Nb-1Ta (at%), for use as implant materials have been assessed and compared with those of Grade 2 Ti, which is known to be a highly compatible metallic biomaterial. The in vitro corrosion resistance was investigated by open circuit potential and electrochemical impedance spectroscopy (EIS) measurements, as a function of exposure time to an artificial physiological environment (Ringer's solution). Open circuit potential values indicated that both the Zr alloys and Grade 2 Ti undergo spontaneous passivation due to spontaneously formed oxide film passivating the metallic surface, in the aggressive environment. It also indicated that the tendency for the formation of a spontaneous oxide is greater for the Zr-1.5Nb-1Ta alloy and that this oxide has better corrosion protection characteristics than the ones formed on Grade 2 Ti or on the Zr-2.5Nb alloy. EIS study showed high impedance values for all samples, increasing with exposure time, indicating an improvement in corrosion resistance of the spontaneous oxide film. The fit obtained suggests a single passive film presents on the metals surface, improving their resistance with exposure time, presenting the highest values to the Zr-1.5Nb-1Ta alloy. For the biocompatibility analysis human osteosarcoma cell line (Saos-2) and human primary bone marrow stromal cells (BMSC) were used. Biocompatibility tests showed that Saos-2 cells grow rapidly, independently of the surface, due to reduced dependency from matrix deposition and microenvironment recognition. BMSC instead display a reduced proliferation, possibly caused by a reduced crosstalk with the metal surface microenvironment. However, once the substrate has been colonized, BMSC seem to respond properly to osteoinduction stimuli, thus supporting a substantial equivalence in the biocompatibility among the Zr alloys and Grade 2 titanium. In summary, high in vitro corrosion resistance together with satisfactory biocompatibility make the Zr-2.5Nb and Zr-1.5Nb-1Ta crystalline alloys promising biomaterials for surgical implants.

A Risk-Informed Approach to Leak-Before-Break Assessment of Pressure Tubes in CANDU Reactors

Abstract The leak-before-break (LBB) assessment of pressure tubes is intended to demonstrate that in the event of through-wall cracking of the tube, there will be sufficient time followed by the leak detection, for a controlled shutdown of the reactor prior to the rupture of the pressure tube. CSA Standard N285.8 (2005, “Technical Requirements for In-Service Evaluation of Zirconium Alloy Pressure Tubes in CANDU Reactors,” Canadian Standards Association) has specified deterministic and probabilistic methods for LBB assessment. Although the deterministic method is simple, the associated degree of conservatism is not quantified and it does not provide a risk-informed basis for the fitness for service assessment. On the other hand, full probabilistic methods based on simulations require excessive amount of information and computation time, making them impractical for routine LBB assessment work. This paper presents an innovative, semiprobabilistic method that bridges the gap between a simple deterministic analysis and complex simulations. In the proposed method, a deterministic criterion of CSA Standard N285.8 is calibrated to specified target probabilities of pressure tube rupture based on the concept of partial factors. This paper also highlights the conservatism associated with the current CSA Standard. The main advantage of the proposed approach is that it retains the simplicity of the deterministic method, yet it provides a practical, risk-informed basis for LBB assessment.

In-reactor performance of pressure tubes in CANDU reactors

The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr–2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate V c, and threshold stress intensity factor, K IH) change with irradiation; the former reach a limiting value at a fluence of <1 × 10 25 n m −2, while V c and K IH reach a steady-state condition after a fluence of about 3 × 10 25 n m −2 and 3 × 10 24 n m −2, respectively. At saturation the UTS is raised by about 200 MPa, toughness is reduced to about 40% of its initial value, V c increases by about a factor of ten while K IH is only slightly reduced. The role of microstructure and trace elements in these behaviours is described. Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and 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. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

S100 Neutron Diffraction Study of Intergranular Stress Development in Zr-2.5%Nb

The harsh environment in the core of a nuclear reactor imposes high constraints for the choice of materials for structural components: materials need high corrosion resistance, good mechanical properties at high temperatures and must sustain their properties under high irradiation. Zr-2.5%Nb is a zirconium-based alloy which fulfils these criteria and has a low neutron cross section. Therefore it is used for the manufacture of pressure tubes in CANDU reactors, for example. It is now well understood that Zr crystals, because of their hexagonal close packed structure at room temperature, present a strong plastic anisotropy [1]. In polycrystals, this anisotropy is responsible for the heterogeneity of grain-to-grain behaviour and the generation of high intergranular misfits after deformation. Moreover, the strong texture usually found in processed materials brings the anisotropy to a macroscopic level. It is therefore important to understand the mechanisms of deformation. The use of neutron diffraction during in-situ mechanical deformation has proven to be a real asset for such studies and much work has been performed on various alloy systems [2]. However some gaps remain in our understanding of how to link, model and predict the behaviours at the macro and micro scales. The influence of some parameters such as initial texture, grain size/shape and the initial intergranular stress state is also not very well identified. This paper presents the results of a set of in-situ thermo-mechanical processes on coupons cut from a Zr2.5%Nb pressure tube. The experiment was performed in-situ at ENGIN-X, the engineering neutron diffractometer at ISIS, UK. A first set of samples were compressed at room temperature along the three processing directions and up to 10% deformation. Another two samples were briefly annealed in-situ up to 560!C and 620!C respectively and then compressed at room temperature. During compression in the as received state, the macroscopic as well as the microscopic anisotropy were evidenced. As expected from the texture, the hoop direction was stronger than the other directions. The neutron diffraction results show that the 0002 reflections generally bear most of the load in the alpha hcp phase. In the rolling direction no such reflections are present and during compression along this direction the second phase undergoes very high strains. In the Poisson’ s directions the intergranular strains are particularly high (up to 5000 microstrain). During annealing substantial changes in the diffraction spectra are evidence of the Beta phase decomposition. By comparing the initial and the post heat treatment d-spacings, it was found that some strain relief occurred. [1] E. Tenckhoff, Deformation Mechanisms, Texture and Anisotropy in Zirconium and Zircaloy, ASTM, Philadelphia, 1988, p. 77. [2] J. W. L. Pang, T. M. Holden, P. A. Turner, T. E. Mason, Intergranular stresses in Zircaloy-2 with rod texture, Acta Materialia 47 (1999), pp.373-383.

WITHDRAWN: In-reactor performance of pressure tubes in CANDU reactors

WITHDRAWN: In-reactor performance of pressure tubes in CANDU reactors

Effect of Irradiation on the Fracture Properties of Zr-2.5Nb Pressure Tubes at the End of Design Life

Abstract To determine the fracture properties of Zr-2.5Nb pressure tubes irradiated until the end of design life, cantilever beam, curved compact toughness, and transverse tensile samples were prepared from a typical pressure tube and irradiated in the high flux reactor OSIRIS at CEA, Saclay, France. Experiments were conducted on two batches of samples mounted in two irradiation inserts. Each insert held sixteen samples of each type of specimen. The first insert was irradiated to a fluence corresponding to approximately half of the design life in a CANDU3 reactor. The experimental results were reported in [1]. Samples in the second insert were irradiated for 10.5 years in OSIRIS and received a maximum neutron fluence of 2.61 × 1026 n/m2 (E &amp;gt; 1 MeV), being equivalent to 2.98 × 1026 n/m2 (E &amp;gt; 1 MeV) in a CANDU reactor, i.e., corresponding to ∼30 years operation in CANDU reactors at 80 % capacity factor. The present report describes the results of tensile, fracture toughness, and Delayed Hydride Cracking (DHC) tests and XRD microstructure analysis from the second batch of specimens. A continuous and gradual evolution in tensile, fracture, DHC properties, and dislocation densities is demonstrated without any evidence of a sudden change following the initial transitient at very low fluence. In the whole high fluence range, there is a very slow rate of increase in c-component dislocation density, strength, and DHC velocity and a slow reduction in elongation and Nb concentration in the β-phase. The a-type dislocation density and fracture toughness remain approximately constant. The results from the second insert of specimens confirm that, following the initial transient at very low fluence, there is little further change in the fracture properties of Zr-2.5Nb pressure tube material. Therefore, material properties behave in a stable and predicable manner to the end of a 30 years design life for CANDU reactor pressure tubes.

A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

In modern CANDU nuclear generating stations, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the fuel bundles and the heavy water (D 2O) coolant. The pressure tubes operate at an internal pressure of ∼10 MPa and temperatures ranging from ∼250°C at the inlet to ∼310°C at the outlet. Over the expected 30 year lifetime of these tubes, they would be subjected to a total fluence of ∼3×10 26 n m −2. In addition, these tubes gradually pick up deuterium as a result of a slow corrosion process. When the hydrogen plus deuterium concentration in the tubes exceeds the hydrogen/deuterium solvus, the tubes are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. The service life of the pressure tubes is determined, in part, by changes in the probability for the rupture of a tube. This probability is made up of the probability for crack initiation by DHC multiplied by the sum of the probabilities of break-before-leak and leak-before-break (LBB). A probabilistic model, BLOOM, is described which makes it possible to estimate the cumulative probabilities of break-before-leak and LBB. The probability of break-before-leak depends on the crack length at first leak detection and the critical crack length. The probability of a LBB depends on the shut-down scenario used. The probabilistic approach is described in relation to an example of a possible shut-down scenario. Key physical input parameters into this analysis are pressure tube mechanical properties, such as the crack length at first coolant leakage, the DHC velocity and the critical crack length. Since none of these parameters are known precisely, either because they depend on material properties, which vary within and between pressure tubes, and/or because of measurement errors, they are given in terms of their means and standard deviations at the different temperatures and pressures defined by the shut-down scenario.

Probabilistic techniques for the assessment of pressure tube hydride blistering in CANDU reactor cores

The integrity of pressure tubes in CANDU reactors can be threatened if there is contact with the surrounding calandria tube and hydride blisters form in the pressure tube. In order to assess the state of a reactor core in which such contacts can occur, a probabilistic assessment technique based upon Monte Carlo methods has been developed. This approach was required due to the distributed nature of many of the factors required for such an assessment. The computer code embodies the results of an extensive research program into the growth and fracture characteristics of hydride blisters in pressure tube material. A single assessment generates distributions of the times to first blister formation and first blister cracking in the reactor core as well as the expected number of blistered tubes as a function of time. Multiple assessments can be used to explore the dependence of the assessment on variation in input assumptions. The result is a powerful tool for making key decisions on reactor maintenance requirements.

Development of Zirconium Alloys for Pressure Tubes in Candu Reactors

Pressure tubes were initially made of Zircaloy-2 but for Pickering Unit 3 and all subsequent power reactors, pressure tubes have been made from Zr-2,5 wt %Nb, The current tubes have a design life of 30 yr and are made by the same extrusion and cold drawing process used to make the Zircaloy-2 tubes, Recent developments have shown that the fabrication route can be modified to change the microstructure and produce tubes that have the same strength, but lower axial elongation, Investigation of other zirconium alloy systems for a high strength alloy resulted in the alloy Excel (Zr-3.5wt % Sn, 0.8wt %Mo, 0.8wt %Nb) being developed for tubes with better in-reactor dimensional stability. Results are promising and long term in-reactor tests are continuing. The relative importance of microstructural and alloying effects on in-reactor properties has yet to be determined.

The Effect of Stress on Orientation of Hydrides in Zirconium Alloy Pressure Tube Materials

AbstractThe presence of radial-axial hydride in pressure tubes in CANDU reactors is to be avoided since it can result in significant reductions in fracture toughness. Fabrication procedures produce tubes which have little radial-axial hydride when hydrided without stress. A quantitative technique for evaluating the susceptibility of pressure tube material to the formation of radial-axial hydride when hydrided under stress has been developed and applied to four types of Zr-2.5 wt % Nb material and to Zircaloy-2. The results indicate that the threshold stress for producing such hydride increases with material strength. This probably reflects grain to grain residual stresses built into the material during fabrication.

The Effect of Stress on Orientation of Hydrides in Zirconium Alloy Pressure Tube Materials

The presence of radial-axial hydride in pressure tubes in CANDU reactors is to be avoided since it can result in significant reductions in fracture toughness. Fabrication procedures produce tubes which have little radial-axial hydride when hydrided without stress. A quantitative technique for evaluating the susceptibility of pressure tube material to the formation of radial-axial hydride when hydrided under stress has been developed and applied to four types of Zr-2.5 wt % Nb material and to Zircaloy-2. The results indicate that the threshold stress for producing such hydride increases with material strength. This probably reflects grain to grain residual stresses built into the material during fabrication.

Development of Zirconium Alloys for Pressure Tubes in Candu Reactors

Pressure tubes were initially made of Zircaloy-2 but for Pickering Unit 3 and all subsequent power reactors, pressure tubes have been made from Zr-2,5 wt %Nb, The current tubes have a design life of 30 yr and are made by the same extrusion and cold drawing process used to make the Zircaloy-2 tubes, Recent developments have shown that the fabrication route can be modified to change the microstructure and produce tubes that have the same strength, but lower axial elongation, Investigation of other zirconium alloy systems for a high strength alloy resulted in the alloy Excel (Zr-3.5wt % Sn, 0.8wt %Mo, 0.8wt %Nb) being developed for tubes with better in-reactor dimensional stability. Results are promising and long term in-reactor tests are continuing. The relative importance of microstructural and alloying effects on in-reactor properties has yet to be determined.

A high-temperature longitudinal strain rate equation for Zr-2.5 wt% Nb pressure tubes

The pressure tubes in CANDU reactors are horizontal. Thus, if, during a postulated loss-of-coolant accident, the pressure tube temperature should rise sufficiently, the self-weight of the pressure tubes together with the weight of the fuel could cause the pressure tubes to sag. Since any pressure tube deformation would control how the core heat is transferred to the surrounding moderator, which is a large heat sink, the accurate prediction of this sag is essential. Most CANDU reactors have pressure tubes of cold-worked Zr-2.5 wt% Nb. A longitudinal strain rate equation was developed for this material using four-point bend tests. This strain rate equation was successful in predicting the longitudinal strain, due to bending, in specimens for which the temperature was ramped at 1°C/s and 5°C/s.

A high-temperature creep model for Zr-2.5 wt% Nb pressure tubes

For a postulated loss-of-coolant accident in a CANDU reactor, in which the primary cooling circuit fails to remove the heat generated in the core, the temperature of the pressure tubes could rise very quickly. Since any deformation of the pressure tubes would control how the core heat is transferred to the surrounding moderator, which is a large heat sink, the accurate prediction of this transient deformation is essential. The majority of the pressure tubes in CANDU reactors are cold-worked Zr-2.5 wt% Nb and creep equations for this material have been developed from uniaxial creep tests. These creep equations were successful in predicting the creep strain in constant-stress uniaxial tests in which the temperature was ramped at rates ranging from 1° C/s to 50° C/s. They also successfully predicted the ballooning of internally pressurized sections of pressure tube that were heated at about 5° C/s.

Tool angle optimization in rod and tube drawing

Beginning with the basic concepts which relate total drawing stress to ideal deformation, redundant shear and friction, rods of various copper alloys were cold drawn to determine the effect of die angle on drawability. The experimentally determined drawing forces are compared with values calculated from current theory. Tool angle optimization is then extended to the design of tools for the production of Zr-2.5 wt pct Nb pressure tubes used in CANDU reactors. The predicted tool angles are compared with empirically optimized tools and demonstrate that optimization experiments carried out with small rod or tube samples can be used to predict the tool geometry necessary for the production of full size CANDU reactor pressure tubes. Finally, the study of tool angle optimization is expanded to include its effects on strains and strain rates occurring in a tube during its passage through the drawing tools. The application of this work is illustrated in relation to the production of DHP copper tubing.

In-reactor creep of zirconium-2.5 wt% niobium at 570 K

We have measured the effect of fast neutron flux at 570 K on the creep rate of specimens of zirconium-2.5 wt% niobium alloy taken from tubes in various metallurgical conditions, using both constant load tensile creep machines and bent-beam stress relaxation. Creep rates calculated from stress relaxation fit on the trend line for the constant load creep data. Between 114 MPa and 450 MPa the creep rate is proportional to neutron flux. The creep rate of specimens from the longitudinal direction is about twice that of specimens from the circumferential direction of a tube. This anisotropy in creep strength is attributed partly to crystallographic texture and partly to deformation substructure. Cold-work is detrimental to in-reactor creep strength; as-extruded material has higher creep strength. In cold-worked material at stresses below 100 MPa the stress exponent, n, in ge ∝ σ n is about 1; n gradually increases with stress being about 10 at 525 MPa and about 100 at 660 MPa. In laboratory tests, rupture ductility correlates inversely with n; the lower n the higher the ductility. Inreactor tests support this correlation thus pressure tubes in CANDU reactors, operating at 117 MPa where n ≈ 1, should have good ductility.