CANDU Pressure Tube Material Research Articles

Analysis of Lattice Change in Aged Zr-2.5%Nb CANDU Pressure Tube Material Using a Neutron Diffraction

Analysis of Lattice Change in Aged Zr-2.5%Nb CANDU Pressure Tube Material Using a Neutron Diffraction

Analysis of Dimensional Change in Zr-2.5%Nb CANDU Pressure Tube Material aged at 300-400°C using Neutron Diffraction

The primary boundary of the CANDU reactor consists of Zr-2.5%Nb alloy pressure tubes and SA106 alloy feeder pipes. The pressure tubes are exposed to fast neutron (E>1Mev) irradiation during reactor operation, and experience dimensional changes. The actual dimensional change in the pressure tube during in-service operation is greater than the design estimate. It is not fully understood what is responsible for the dimensional changes in the pressure tubes. In this study, to accelerate the aging effect on the pressure tube material, aging was performed at 300-400°C for up to 20,000 hours and neutron diffraction was used to track variation in lattice spacing, thereby identifying the reason for the dimensional change. The analysis result showed that α-Zr, the main phase of Zr-2.5%Nb, had little dimensional change at 350°C or lower. On the other hand, the aging treatment at 400°C resulted in anisotropic expansion in the α-Zr of the HCP crystal, of 0.02% and 0.08% in the (1010) and (0002) directions, respectively. It was confirmed that above 350oC the β-phase in the Zr-2.5%Nb alloy was decomposed to precipitate β-Nb . The driving force of the lattice expansion may be due to the diffusing out tendency of the super saturated Nb in α-Zr and due to the residual cold working effect applied during the manufacturing process. The enhancing effects of fast neutron irradiation on lattice diffusion are discussed in detail.

Effect of Zr+ irradiation damage and crystal orientation on the uniaxial deformation of Zr–2.5%Nb micro-pillars: part 1, deformation of single-phase α Zr micro-pillars

ABSTRACT The uniaxial flow stress of 1 μm diameter micro-pillars made from single α Zr grains cut from Zr–2.5%Nb CANDU pressure tube material was assessed in the non-implanted condition and after 6.0 dpa Zr+ implantation performed at 25°C and 300°C to simulate neutron irradiation. The normal flow stress, at 10% strain, of the non-implanted micro-pillars was about 70% higher than that for larger diameter polycrystalline pillars indicating length-scale dependence. The flow stress anisotropy displays a stronger length-scale dependence for micro-pillars that were loaded along the <0001> basal pole direction than along other directions. The normalised shear stress of α Zr micro-pillars aligned for single-slip deformation displayed a dependence upon crystal orientation. These findings provide new information on the mechanisms by which small volume α Zr phase ligaments, located in crack-tip regions, deform plastically and thus contribute to the ductile fracture toughness of neutron-irradiated Zr–2.5%Nb CANDU pressure tubes.

Deducing the stress–strain response of anisotropic Zr–2.5%Nb pressure tubing by spherical indentation testing

In this paper we developed a method of analysis by which the average stress–plastic strain flow curve of mechanically anisotropic materials can be deduced from spherical indentation test data. Our analysis is based upon spherical indentation tests performed on the extruded and cold-drawn Zr–2.5%Nb CANDU pressure tube material over the range of temperature from 25°C to 300°C. The indentation force and depth data were analyzed and σavg and ɛavg were calculated using previously reported equations developed for spherical indentation of isotropic material which were then modified, by incorporating the appropriate Hill’s anisotropy coefficients, to characterize the anisotropic yield stress of the indented material. The resulting flow curves were dependent on indentation direction and correspond closely with flow curves obtained from previously reported conventional uniaxial stress tests performed on the Zr–2.5%Nb material. Indentation tests performed with large, 200μm, and small, 40μm, diameter spheres indicate that for small diameter indentations, when the indentation depth is less than several micrometers, the calculated σavg is heavily influenced by the depth dependence of the yield strength of the indented material.

Assessment of the anisotropic flow stress and plastic strain of Zr–2.5%Nb pressure tubes at temperature from 25 °C to 300 °C

In this paper uniaxial compression testing is used to assess the stress versus plastic strain response in the axial, radial and transverse directions of the mechanically anisotropic Zr–2.5%Nb CANDU pressure tube material over the temperature range from 25 to 300°C. The data from these tests are used to determine the Hill’s anisotropy coefficients F=0.38±0.01, G=0.202±0.001, and H=0.62±0.01 of the pressure tube material. These coefficients were found to be independent of temperature and plastic strain. The results of this study represent the only experimentation-based determination of the Hill’s anisotropy coefficients F, G, and H of extruded and cold-drawn Zr–2.5%Nb CANDU pressure tubes over a wide temperature range extending up to the 300°C in-service temperature of these pressure tubes.

Assessment of the Effect of Irradiation Temperature on the Mechanical Anisotropy of the Zr+ Ion Irradiated Zr-2.5%Nb

ABSTRACTWe present here new information on the effect of irradiation temperature on the strength and mechanical anisotropy of Zr-2.5%Nb CANDU pressure tube material. Polished samples aligned normal to the transverse (TN), axial (AN) and radial (RN) directions of the pressure tube were irradiated at 300°C with 8.5 MeV Zr+ ions to assess the effect of concurrent thermal annealing of the irradiation damage. Constant-load micro-indentation creep tests were performed at 25°C at indentation depths from 0.1 to 2.0 μm on the ion irradiated samples.The increase in the initial indentation stress with increasing levels of Zr+ ion irradiation at 300°C was lower than that reported earlier for similar samples exposed to Zr+ irradiation at 25°C. While the anisotropy of the indentation stress decreased significantly with Zr+ ion irradiation, the level of the decrease was reduced when the irradiation was performed at 300oC compared to 25oC. The apparent activation energy ΔG0 of the obstacles that limit the rate of dislocation glide during indentation creep did not change with indentation direction but did increase with increasing levels of Zr+ ion damage. The values of ΔG0 were, again, lower for samples that were irradiated at 300°C than for those irradiated at 25oC.The observed differences in the magnitude of, and the anisotropy of, the initial indentation stress and also the decrease in the apparent activation energy of the indentation creep process of Zr-2.5%Nb samples irradiated with Zr+ ions at 300oC compared to those irradiated at 25oC indicate the effect that concurrent thermal annealing has on the accumulation of irradiation damage. The effect of irradiation temperature on reducing the degree of, and the strength of, irradiation induced crystallographic damage must therefore be considered when predicting the strength and thermal creep behaviour of irradiated nuclear materials.

Effect of ion irradiation and indentation depth on the kinetics of deformation during micro-indentation of Zr–2.5%Nb pressure tube material at 25 °C

Micro-indentation creep tests were performed at 25 °C on radial-normal samples cut from Zr–2.5Nb CANDU pressure tube material in both the as-fabricated condition and after irradiation with 8.5 MeV Zr + ions. The average indentation stress, and hence the yield stress, was found to increase with decreasing indentation depth and with increasing levels of ion irradiation. The activation energy of the indentation creep rate and hence the, activation energy of the obstacles that limit the rate of dislocation glide, was independent of indentation depth but increased from Δ G 0 = 0.185 to 0.215 μb 3 with increasing ion irradiation damage. The magnitude of the activation energy indicates that ion irradiation introduces a new type of obstacle into the microstructure which reduces the low temperature indentation creep rate of Zr–2.5Nb pressure tubes. This is supported by TEM images showing that Zr + ion irradiation produces small, nanometer size, dislocation loops which act as obstacles to dislocation glide and thus influence both the yield stress and the activation energy of the low-temperature thermal creep of Zr–2.5Nb pressure tube material. These findings suggest that neutron irradiation will have similar effect upon yield stress and low-temperature thermal creep as the Zr + ion irradiation since both create similar crystallographic defects in Zr–2.5Nb pressure tubes.

Irradiation and Operating Temperature Effects on the Tensile Characteristics of CANDU Pressure Tube Material

To investigate the degradation of mechanical properties induced mainly by neutron irradiation and operating temperatures, tensile tests were conducted from room temperature to 300°C using irradiated and unirradiated Zr-2.5Nb pressure tube materials. The irradiated longitudinal and transverse specimens collected from the coolant inlet, middle and outlet parts of the tube which had been operated in the CANDU reactor and showed different operating temperatures and irradiation fluence. The different tensile behavior was characterized by the tensile loading direction in the unirradiated tube. The transverse specimen showed higher strength and lower elongation than those of the longitudinal one. The increased strength hardening and the decreased elongation embrittlement of the irradiated material were compared to those of the unirradiated one. While the tensile strength of the inlet was higher than that of the outlet, the elongation of the inlet was lower than that of outlet. Considering the operation condition, it was proposed that the operating temperature could be a more effective parameter than the irradiation fluence for long-time life. Through TEM observation, it was also found that while the a-type dislocation density was increased, the c-type dislocation was not changed in the irradiated material. The fact that the higher dislocation density was sequentially distributed over the inlet to the outlet parts was consistent with the distribution of the tensile strength.

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.

Modelling the fissuring flat-fracture mode of crack growth in Zr-2.5Nb Candu pressure-tube material

Modelling of 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, is reported. A preliminary attempt has been made at modelling the characteristics of the fracture process zone associated with a crack that tunnels at the specimen mid-section. Based on the experimentally determined Jr curve, both the fracture-process zone size and the crack opening displacement at the trailing edge of the zone have been predicted, and their values are seemingly not unreasonable. The initial considerations enable specific issues to be highlighted which need to be addressed before a complete picture is obtained of the fissuring mode of crack growth. A theoretical analysis has also shown that the non-tunnelling material at the specimen side surfaces exerts very little restraint on the cumulative (tunnelling) mode of crack propagation, a prediction that is consistent with the experimental finding that Stage 1 crack growth in irradiated material is associated with an essentially flat Jr curve.