CANDU Pressure Tubes Research Articles

Development of CANDU Pressure Tube Integrity Evaluation System;Its Application to Sharp Flaw and Blunt Notch

Development of CANDU Pressure Tube Integrity Evaluation System;Its Application to Sharp Flaw and Blunt Notch

Determination of anisotropic elastic moduli of Zr-2.5Nb CANDU pressure tube materials

In order to determine the elastic stiffness of Zr-2.5Nb CANDU pressure tube materials, the resonance frequencies of a rectangular parallelepiped specimen were measured and compared with the calculated resonance frequencies based on the input data of estimated stiffness of a polycrystalline Zr-2.5Nb specimen, dimensions and density. The estimated elastic stiffness was determined using its orientation distribution function measured by x-ray diffraction and the reported elastic stiffness of a zirconium single crystal. Through comparison of calculated frequencies with that measured by resonance ultrasound spectroscopy, accurate elastic stiffness has been determined by iteration and convergence processes.

Leak-before-break assessment of CANDU pressure tube considering leak detection capability

The leak before break (LBB) concept is employed as defense in depth in CANDU reactors in order to avoid an unstable failure in pressure tubes. The LBB in pressure tubes can be ensured when detection, confirmation and location of the leak are carried out and the reactor is placed in a depressurized condition before the crack exceeds critical crack length. Leak detection and location is provided by the annulus gas system (AGS). Therefore, the evaluation parameters for LBB assessment, such as leak detection and leak location capabilities, should be made available through AGS performance test. Recently, the AGS in-situ tests with a simulated moisture injection were carried out in one of the CANDU reactors in Korea. This paper presents the LBB assessment performed taking into account the leak detection capability and the modification of operating procedure therefrom.

Assessment of aging of Zr2.5Nb pressure tubes in CANDU ™ reactors

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 about 10 MPa and temperatures ranging from about 250°C at the inlet to about 310°C at the outlet. Over the expected 30 year lifetime of these tubes they will be subjected to a total fluence of approximately 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. A fitness-for-service methodology has been developed which assures that this will not happen. A key element in this methodology is the acquisition of data and understanding—from surveillance and accelerated aging testing—to assess and predict changes in the DHC initiation threshold, the DHC velocity and the fracture toughness (critical crack length) as a function of service time. The most recent results of the DHC and fracture toughness properties of CANDU pressure tubes as a function of time in service are presented and used to suggest procedures for mitigation and life extension of the pressure tubes.

Liquid metal embrittlement of Zr-2.5% Nb alloy

Liquid metal embrittlement (LME) of zirconium alloys in mercury is entirely transgranular. Grains not oriented forpseudo-cleavage on the basal plane fail by slip on the prism planes to form the characteristic ‘fluting’. Because of this, fracture in Hg of Zr-2.5 Nb small compact tension specimens (cut from five different CANDU pressure tubes) showed large differences in the macroscopic fracture topography that could be related to differences in the local continuity of the basal planes in the structure, even though the average textures were very similar. These differences could be related to differences in the fracture toughness of the tubes. Although LME in zirconium alloys has been thought to be generally transgranular because of previous experience in Hg, Cs/Cd mixtures, a Hg/In/TI eutectic, and Li, tests in Ga showed no transgranular cracking, but small amounts of intergranular cracking were observed that appeared to be limited by poor wetting of Zr by Ga.

Modelling the softening within the fracture process zone associated with the fissuring mode of crack growth in Zr-2.5Nb CANDU pressure tube material

The fissuring mode of fracture in CANDU pressure tube material, and in particular Stage 1 crack growth (essentially flat JR curve) as observed in some irradiated compact toughness specimens has been investigated. Models are presented of the fracture process zone associated with a crack that tunnels at the specimen mid-section, which extends preliminary work reported earlier. Various types of process zone behaviour are analysed, and based on an appropriate value for Jc, the J value associated with the cumulative mode of crack propagation in irradiated material, together with an estimate of the tensile stress at the leading edge of the process zone, the known failure mechanism (formation, growth and coalescence of voids) of the ligaments between the fissures is shown to be reasonably consistent with the experimental measurements of the fissure spacing and fissure length.

Investigation of texture and interfaces in a Zr-2.5Nb alloy with zirconium hydrides

Neutron diffraction has provided a simultaneous measurement of the crystallographic texture of the α and β zirconium phases, and of the δ zirconium hydride phase, in a Zr-2.5Nb CANDU pressure tube material. The α and β phases exhibit textures typical of drawn hcp and bcc materials, respectively. Surprisingly, the hydrides exhibit a texture similar to fcc material drawing textures, which points to the hydride formation-path in the metal substrate. Analysis of multi-phased interface relationships revealed a new plane matching relationship between the beta and delta phases, which can be approximated by the Kurdjumov—Sachs (K—S) type relationship. The β—δ interfaces were found to have a very low amount (less than 9%) of coincidence site lattice (CSL) boundaries.

SIMS studies of the corrosion of zirconium

The integrity of CANDU pressure tubes is critically dependent on the protectiveness of the oxide film formed on its surface. The actual protective layer may be only a few nanometres in thickness. In an on-going investigation of the properties of the protective oxide film, and the relationship of these properties to the corrosion behaviour of the pressure tubes, SIMS has become an important component of the analytical methods used. In this paper, examples will be given showing the application of static and dynamic SIMS to the study of the interaction of O 2 and H 2 with single-crystal Zr and the diffusion of H and O in Zr, the application of dynamic SIMS to the characterization of D- or H-containing species in pressure tube oxides, and the effects of incorporation of impurities into the oxide. These studies will be related to the current understanding of CANDU pressure tube corrosion.

A comparison of the use of the modified J-integral and the deformation J-integral in procedures for estimating the critical crack length for CANDU pressure tubes

Abstract The paper presents the results of a theoretical investigation whose objective has been to see whether there are advantages to be gained from using the modified J-integral in procedures for estimating the critical crack length for CANDU pressure tubes. For typical operation conditions, and with irradiated tubes having critical crack lengths over a wide range, it is shown that the slope of the modified J-integral JM-Δa crack growth resistance curve for a pressure tube crack is only marginally greater than the slope of the corresponding deformation J-integral JD-Δa curve; the results are expressed in terms of the parameter Z ∗ , which is dJM/da − dJD/da and the parameter Q, which is the fractional difference between dJM/da and dJD/da. In the light of these findings, there would appear to be little advantage to be gained in using JM, rather than JD, as a characterizing parameter for crack growth in a CANDU pressure tube.

Mitigation of degradation mechanisms affecting CANDU pressure tubes

Abstract The Zr-2.5Nb alloy pressure tube provides the primary containment for the coolant in the CANDU reactor. Because these tubes contain the individual fuel strings and pass through the center of the reactor, they are exposed to the full fast neutron flux of up to 4.0 × 1017 neutrons per square meter per second (E > 1 MeV). Any demonstration of the integrity of these tubes relies on a leak-before-break (LBB) assessment, which, in turn, depends on the fracture properties of the tube. These properties become degraded with service due to both the gradual pickup of hydrogen isotopes and the reduction of fracture toughness due to irradiation. This paper gives an overview of the relevant fracture properties used in the LBB scenario and the effect of service conditions on their degradation. The particular properties discussed are the susceptibility to sub-critical crack growth by delayed hydride cracking, and the reduction in fracture toughness which reduces the critical crack length. Methods of protecting the tubes against fracture and improving their properties are discussed.

Analysis of ZrNb fuel channel surfaces for hydrogen and other elements using Secondary Ion Mass Spectrometry (SIMS)

Understanding the mechanisms of hydrogen ingress into CANDU pressure tubes fabricated from zirconium-2.5 wt% niobium alloy requires knowledge of the hydrogen concentration in the surface oxide, the oxide/metal interface and in the alloy phase beneath the interface. Secondary Ion Mass Spectrometry (SIMS) has been shown to have attractive capabilities for detecting hydrogen isotopes in such surface films. The depth concentration profiles of deuterium, lithium, fluorine and iron have been measured on specimens taken from several locations on two pressure tubes from Unit 3 of the Pickering Nuclear Generating Station. These tubes were known to have widely differing deuterium concentrations in their bulk phases. SIMS profiles were obtained from both the outside (OD) and inside (ID) tube surfaces. The OD studies show that for regions where bulk deuterium pickup is small, the deuterium concentration within the surface oxide decreases steadily with depth and is 50–100 times lower near the oxide/alloy interface. For regions of high pickup, the profiles showed only a small concentration gradient of deuterium across the oxide thickness. This infers the importance of the surface oxide in suppressing hydrogen ingress. The ID studies also show that regions with high deuterium pickup have anomalous deuterium depth profile structure near the oxide metal interface. This suggests that the integrity of the ID oxide film may also bear some responsibility for enhanced deuterium ingress at these locations. Variations in iron, lithium or fluorine concentrations within the ID or OD oxide scales do not appear to have a significant effect on the deuterium profiles measured.

The role of leak-before-break in assessments of flaws detected in CANDU pressure tubes

This paper reviews the role of the Leak-Before-Break (LBB) concept in the Fitness for Service Guidelines being developed for cold worked (cw) Zr-2·5Nb pressure tubes in a CANDU reactor. The guidelines complement the rules of Section XI of the ASME Code and the requirements of Canadian Standards Association (CSA) CAN 3-N285.4-M83. The evaluation procedures in the guidelines consist of a flaw growth analysis to determine the maximum size of the flaw at the end of the evaluation period. It must then be demonstrated that the flaw is stable with adequate margins of safety for the various loading conditions. For the delayed hydride cracking failure mode LBB is used as defense in depth against unstable rupture. First the flaw must be shown to be non-susceptible to propagation by delayed hydride cracking during normal operating conditions. In addition, it must then be demonstrated that, if the flaw were to penetrate the tube wall, the leaking coolant would be detected and the reactor shutdown before the postulated crack became unstable. The Guidelines contain criteria for performing both deterministic and probabilistic LBB analyses.

A probabilistic approach to leak-before-break in CANDU pressure tubes

In the candu R reactor, the coolant passes through the core in zirconium alloy pressure tubes. A few of these pressure tubes have leaked at cracks near the rolled joint where the pressure tube is attached to the end fitting. A probabilistic methodology, and associated computer code (called marathon), has been developed to calculate the time from first leakage to unstable fracture in a probabilistic format. The methodology explicitly uses material property distributions, and allows the risk associated with leak-before-break to be estimated. A model of the leak detection system is included to calculate the time between leak detection and unstable fracture. The sensitivity of the risk to changing reactor conditions allows the optimization of reactor management after leak detection. Preliminary material property distributions show the probability of unstable fracture is very low, and that ample time is available to shut down the unit and locate the leaking tube.

The influence of axial end loads on critical crack length behaviour of CANDU-PHW pressure tubing

Slit-burst tests on CANDU pressure tubes are described in which the end loads have been varied. The internal pressure at burst, for tubes containing through-wall axial slits, was unaffected by the end load. These results are compared with prevailing theory.

In-Reactor deformation of internally pressurized Zr-2.5 wt% Nb tubes at 570 K

Small tubes of cold-worked Zr-2.5 wt% Nb, which were internally pressurized to give hoop stresses from −25 to 495 MPa have been irradiated at 571 K for up to 23400 h. Diameters and length have been measured periodically. The resulting circumferential and axial strain rates have been compared to those calculated from an equation derived earlier from CANDU pressure tubes which takes account of texture, grain shape, dislocation density, stress, fast neutron flux and temperature; 80% prism slip and 20% basal slip is assumed for creep. This comparison shows that axial and circumferential irradiation growth is well calculated by the equation but creep of the small tubes is greater than predicted and the anisotropy between circumferential and axial creep strains is not well predicted. Since an unrealistic-ally high proportion of basal slip is required to account for this anisotropy, another factor, e.g. distribution of active dislocations, is influencing anisotropy. The lower creep strength of the small tubes compared to pressure tubes is probably due to the absence of a β-network around the α-grains.

Anisotropy of irradiation creep and growth of zirconium alloy pressure tubes

Irradiation creep and growth anisotropy factors calculated from crystallographic texture and grain shape measurements, together with measured axial and transverse strain rates of zirconium alloy pressure tubes in CANDU ∗ ∗ CANada Deuterium Uranium. power reactors, have been used to calculate proportions of creep and growth. Using the irradiation creep and growth rates derived for cold drawn Zr-2.5%Nb and cold drawn Zircaloy-2 pressure tubes, we have defined ideal textures and grain shapes resulting in minimum axial and transverse strain rates. For the conditions of stress, temperature and fast neutron flux under which CANDU pressure tubes operate, proportions of growth are too small compared to the creep to get the ideal of zero axial and transverse strains. The closest approach to this ideal is when basal poles are in the radial/transverse plane, regardless of grain shape of alloy. The most favourable grain shape is plate shaped with the plates oriented in the axial/radial plane.

In Reactor Deformation of Candu Pressure Tubes

The mechanisms of irradiation growth and irradiation creep are described and their effect on the in-service deformation of pressure tubes in CANDU reactors is discussed. Irradiation growth and creep are shown to be influenced by the metallurgical structure of the pressure tube. To reduce the extent of elongation of the pressure tubes over their service life specific modifications to the metallurgical structure are proposed. The fabrication route required to produce tubes with the required modified metallurgical structure is presented. Résumé On décrit les mécanismes de croissance et de propagation des fissures qu'on retrouve chez les métaux irradiés et on en discute les effets, tels qu'observés chez les tubes sous pression déformés lors de leur utilisation au sein de réacteurs CANDU. Parce que la microstructure métallique des tubes irradiés influe sur les mécanismes de croissance et de propagation des fissures, on propose de la modifier, afin de réduire l'élongation des tubes au cours de leur vie utile. On présente également la technique de fabrication à utiliser pour obtenir la microstructure métallique modifiée.

In Reactor Deformation of Candu Pressure Tubes

Abstract The mechanisms of irradiation growth and irradiation creep are described and their effect on the in-service deformation of pressure tubes in CANDU reactors is discussed. Irradiation growth and creep are shown to be influenced by the metallurgical structure of the pressure tube. To reduce the extent of elongation of the pressure tubes over their service life specific modifications to the metallurgical structure are proposed. The fabrication route required to produce tubes with the required modified metallurgical structure is presented. Resume On decrit les mecanismes de croissance et de propagation des fissures qu'on retrouve chez les metaux irradies et on en discute les effets, tels qu'observes chez les tubes sous pression deformes lors de leur utilisation au sein de reacteurs CANDU. Parce que la microstructure metallique des tubes irradies influe sur les mecanismes de croissance et de propagation des fissures, on propose de la modifier, afin de reduire l'elongation des tubes au cours de leur vie...