Effect of hydrogen and neutron irradiation on the failure of flawed Zircaloy-2 pressure tubes

Experiments were made with Zircaloy-2 tubes containing axial slits. Failure stress at 20 and 300 °C is increased by neutron irradiation: hydride precipitates reduce failure stress at 20 °C but not at 300 °C. There is evidence that stress applied during irradiation, and neutron dose, affect failure stress.

Crack opening displacement measurements on notched bend specimens predicted within 20% the critical crack lengths measured on full size tubes.     

Liquid metal embrittlement of Zircaloy-2 tubes by cesium and/or cadmium

The susceptibility of stress relieved Zircaloy-2 tubes to liquid metal embrittlement by cesium (Cs) and/or cadmium (Cd) at 340°C using the argon pressurization technique was investigated. The results of liquid Cs experiments showed considerable scattering while reproducible results were obtained in liquid Cs-Cd experiments. Considerable embrittlement was observed by Cd, particularly at 100 mg and liquid Cs-Cd mixture. Brittle fracture with X-mark and burst was observed. The embrittlement effect was attributed to Cd due to the formation of a Zr-Cd intermetallic layer which was observed by SEM and metallography. The increase in wettability by liquid Cs led to considerable embrittlement of the Cs-Cd mixture even when small amounts of Cd were used. It also led to considerable and reproducible results.     

Creep ductility of cold-worked Zr-2.5 wt% Nb and Zircaloy-2 tubes in-reactor

Deformation of internally pressurized 15 mm diameter tubes of cold-worked Zr-2.5 wt% Nb and Zircaloy-2 has been measured in axial and hoop directions for irradiations up to 24 000 h at 571 K. Specimens have deformed up to 18.1% in the hoop direction, with hoop stresses up to 495 MPa, without rupture. Since the stress sensitivity of creep rate is high (>5) under the test conditions, and is low (≈1) under operating conditions for pressure tubes, creep ductility of pressure tubes should be very high.     

Stress-corrosion-cracking of Zircaloy-4 cladding tubes: Part 1. Threshold in the presence of iodine

The burst pressure has been measured as a function of the pressurizing rate, with and without iodine present and at the two temperatures 343 and 400° C, for two batches (designated A and B) of Zircaloy-4 cladding tubes made by two different manufacturers to the same specification. Diameter increases were recorded during pressurization. The burst pressure in the absence of iodine decreases with decreasing pressurizing rate. In the presence of iodine the burst pressure, for fast pressure ramps, essentially equals that measured without iodine present. Thereafter, for decreasing ramp speeds the burst pressure in the presence of iodine first rapidly decreases and then tends asymptotically to approach a constant value — the stress-corrosion cracking (SCC) threshold stress. Analysis of these data in terms of a simple model yields threshold stresses and crack velocities in good agreement with literature values. The higher threshold for tube type A correlates to a lower stress relieving temperature.     

Formulation and verification of design criteria for Zircaloy-2 pressure tubing

Results from experimental and surveillance programs for the Zircaloy-2 pressure tubes in the Plutonium Recycle Test Reactor are reviewed. The review of these results shows that the original PRTR design basis was conservative, that a higher working stress and longer tube life can be used in present reactor applications, and that there are no indications of an increased likelihood of catastrophic failure due to normal reactor operations.     

Circumferential creep properties of stress-relieved Zircaloy-4 and Zr-Nb-Sn-Fe cladding tubes

In this study, the circumferential creep behaviors of stress-relieved Zircaloy-4 and Zr-Nb-Sn-Fe alloys were investigated. The out-of reactor creep resistance of Zircaloy-4 was found to be better than that of the Zr-Nb-Sn-Fe alloy in the temperature and stress range of this study, suggesting that the solution strengthening effect of Sn is more effective in restraining dislocation motion than the precipitate strengthening effect of Nb. Zircaloy-4 and Zr-Nb-Sn-Fe have stress exponents between 6.5 and 7.5 in the lower stress region. The stress exponent decreased to about 3 for Zircaloy-4 and to about 4.2 for Zr-Nb-Sn-Fe in the high stress region. The activation energy increased from 240 to 270 kJ/mol for Zircaloy-4 and Zr-Nb-Sn-Fe as the applied stress increased from 80 to 150 MPa. The creep activation energy observed in this study is close to those of creep for Zircaloy-2 and Zr. The lower stress exponent in the higher stress region for Zircaloy-4 with a relatively higher Sn content compared to Zr-Nb-Sn-Fe alloy may be supported by the fact that the drag force of Sn atoms acting on dislocations increases with increase of Sn content. Larson-Miller Parameter (LMP) decreased with the increase of applied stress, and Zircaloy-4 has a higher LMP than the Zr-Nb-Sn-Fe alloy, compatible with the lower creep rate in Zircaloy-4.     

Stress-induced reorientation of hydrides and mechanical properties of Zircaloy-4 cladding tubes

Stress-induced reorientation of hydrides and its effect on the stress–strain response of Zircaloy-4 cladding tubes were investigated. The reorientation of hydrides along the radial direction was most pronounced if the tube was cooled from 300 to 200 °C under circumferential loading. Reorientation occurred much less frequently at either higher (cooled from 400 to 300 °C) or lower (cooled from 200 to 100 °C) temperature range. The population of radial hydrides in R43H7 (which was cooled from 400 to 300 °C and maintained at 300 °C for 7 h) increased drastically during annealing at 300 °C, suggesting time dependent stress-aided dissolution of circumferential hydrides and reprecipitation of radial hydrides. The drastic decrease of the strength and the complete loss of the ductility were observed in R32AC and R43H7.     

Elastic-plastic analysis of small cracks in tubes under internal pressure and bending

Elastic-plastic finite element analyses were conducted to generate new solutions of J-integral and crack-opening displacement (COD) for short through-wall cracks in pipes subjected to combined bending and tension loads. The results are presented in terms of the well-known GE/EPRI influence functions to allow comparisons with some limited results in the literature. Two different pipe pressures with values of 7.24 MPa (1050 psi) and 15.51 MPa (2250 psi) simulating BWR and PWR operating conditions, respectively, were used to evaluate the effects of pressure on J and COD. Pipes with various radius-to-thickness ratios, crack sizes, and material parameters were analyzed. Limited analyses were also performed to evaluate the effects of hoop stresses in pipes under pure pressure loads. The results suggest that the fracture response parameters can be significantly increased by pressure-induced axial tension for larger crack size, material hardening constant, and radius-to-thickness ratio of the pipe. The presence of pressure-induced hoop stresses also increases the fracture response, but in low-hardening materials their effects are insignificant due to small plastic-zone size that was expected for the intensity of pipe pressure and crack size considered in this study. However, for high-hardening materials when the plastic-zone size is not negligible, the hoop stresses can moderately increase J and COD.     

Irradiation-Induced Dissolution of Precipitates in Zircaloy-2

Calibration curves of extremely low concentrations of the alloying elements Sn, Fe, Cr and Ni in Zircaloy were obtained, using standard samples, by energy dispersive X-ray spectroscopy to measure concentration distributions of alloying elements dissolved in the Zircaloy matrix. Their detectable limits were 0.21 at% for Sn, 0.06 at% for Fe. 0.04 at% for Cr and 0.03 at% for Ni. Then concentration distributions of alloying elements in unirradiated and neutron irradiated Zircaloy-2 were measured using these calibration curves. It was confirmed that neutron irradiation increased the dissolved concentrations of Fe. Cr and Ni. Furthermore, Cr diffused slower than Fe and Ni. It was suggested that the rate limiting process of irradiation-induced dissolution from Fe, Cr-type precipitates into the matrix was the diffusion of alloying atoms in the precipitates and that the dissolution process proceeded due to displacement of alloying atoms from the precipitates into the matrix and diffusion in the matrix.     

Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 300°C

Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300°C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300°C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300°C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor.     

Low-cycle fatigue studies on nuclear reactor Zircaloy-2 fuel tubes at room temperature, 300 and 350°C

Constant amplitude strain controlled fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature, 300 and 350°C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The relationship between the plastic strain range, Δ?_p and the number of cyles to failure, N_f was found to be of a simple power law of the form N^β_f · Δ?_p = constant, at all the test temperatures. Strain-hardening coefficients were determined from monotonic tension tests at these temperatures and it was concluded that the material is more or less in a cyclically stable condition.     

The Effect of Oxide Microstructure on Kinetic Transition in Out-of-pile Steam Corrosion Test for Zircaloy-2 and Nb-added Zircaloy-2

In order to study the mechanism of kinetic transition of corrosion rate for zirconium alloys, oxide films formed on Zircaloy-2 (Zry-2) and Nb-added Zircaloy-2 (0.5Nb/Zry-2) in steam at 673 K and 10.3 MPa were examined with TEM and SIMS.

Kinetic transition occurred at almost the same oxide thicknesses for both Zry-2 and 0.5Nb/Zry-2, but the corrosion rate after the transitions were quite different for the two alloys. Zircaloy-2 showed cyclical oxidation, while the weight gain of 0.5Nb/Zry-2 increased linearly.

The morphology and crystal structure were similar for the oxides of the two alloys and both the oxide films still mainly consisted of columnar grains even after the transition. Interface layers which mainly consisted of a-Zr crystallites were observed for both alloys and the oxygen content in the interface layers increased after the transition.

The solute concentrations of Fe, Cr and Ni became higher, accompanying the increase of oxygen concentrations at columnar grain boundaries in the oxide films after the transition for 0.5Nb/Zry-2. It was thought that the properties of grain boundaries of the 0.5Nb/Zry-2 oxide films changed after the transition, and the increase in oxygen diffusivity at grain boundaries caused the linear increase in weight gain.     

Fracture of Zircaloy-2 Cladding in Iodine Environment

The sensitivity coefficients of neutronic performance parameters in high-conversion LWR cells have been calculated by means of the SAINT code. In order to show the specific features of the sensitivity coefficients in the HCLWR cells, the differences between sensitivities were investigated for cells with different moderator to fuel volume ratios and different Pu enrichments. The burnup dependence of the sensitivities was also discussed with an emphasis on the effect of fission products on the cell parameters.

We have performed the sensitivity analysis for the PROTEUS cores. Group constants of main heavy nuclides were compared for the different cell calculational methods; SRAC and VIM, and the different cross section libraries; JENDL-2 and ENDF/B-IV. The effect of the differences in group constants was estimated for the cell parameters k _∞, reaction rate ratio and coolant void worth. The differences in the ²³⁸U capture and ²³⁹Pu fission group constants in the resolved and unresolved resonance range produced 0.3～1.0% change in k _∞ and 1～6% change in the coolant void worth. These effect was largely dependent on the coolant void fraction.     

Stress-corrosion cracking of Zircaloy-2 in methanol-iodine solutions

Annealed Zircaloy-2 cracks intergranularly when uniaxially stressed in boiling methanol-iodine solutions containing 0.002 to 20.0 g $\text{I}{}_2\text{1}$. The time to failure decreases with increasing iodine concentration. Stressing to 75% of the 0.2% proof-stress increases the dissolution rate by about 20 times. Both the corrosion rate and the time to failure under stress are independent of impressed electrochemical potential. The activation energy for dissolution, 8350 cals/mole for the unstressed condition, is reduced to 3360 cals/ mole under applied stress. This activation energy probably corresponds to the adsorption of I₂ on the Zircaloy surface. The mechanism of the stress-corrosion cracking seems to be one of stress—assisted dissolution controlled by the adsorption of I₂.     

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 (D₂O) 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 3×10²⁶ 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.     

Influence of iodine on the burst strain of Zircaloy-4 cladding tubes under simulated reactor accident conditions

The Stress Corrosion Cracking (SCC) behavior of short tubular Zircaloy-4 (Zry) specimens by the action of iodine was investigated out-of-pile between 600 and 1100°C under inert gas conditions. The Zry cladding tubes were used as-received, with an inner preliminary oxidation and pre-damaged, respectively. Moreover, the influence of the UO₂ oxygen potential on the SCC behavior was studied. The burst and creep-rupture tests (time-to-rupture ?15 min) clearly show that the deformation behavior of Zry cladding tubes below 850°C is influenced by the presence of iodine. A low ductility failure of Zry tubes takes place that is characterized by little deformation as compared to reference specimens without iodine. Moreover, in the isothermal, isobaric experiments, the time-to-failure of the iodine containing specimens is markedly shorter as compared to the iodine-free reference specimens. Internal preoxidation of the cladding tube or UO₂ in the specimens exert an additional influence on the mechanical properties of Zry. SEM examinations of the rupture surface of the cladding tubes show that in the presence of iodine and at burst temperatures below 850°C the cracks in the cladding material are mainly intergranular followed by ductile residual rupture.     

Stress Reorientation of Hydrides in Recrystallized Zircaloy-2 Sheet

A series of measurements was performed to examine the effects of texture, hydrogen concentration, thermal cycling and morphology of hydrides on stress reorientation of hydrides using unirradiated recrystallized Zircaloy-2 sheet materials. It was found that the threshold stress for reorientation of hydrides is approximately 80 MPa, and the threshold stress is not affected by hydrogen concentration, thermal cycling or hydride morphology before the hydride reorientation test. Over the threshold stress, texture, thermal cycling and morphology of hydrides affect the stress reorientation. The modified Ells' equation was proposed for expressing the effect of the texture of Zircaloy-2.     

The selection of the optimum parameters for an atomic electric generating station

A system of equations has been derived, giving a set of optimum parameters for an atomic electric generating station, which insure minimum cost of the electrical energy produced. It is assumed that the layout, materials, and type and elements of equipment have been previously selected, and that the problem is one of finding the optimum numerical values of the constructional and operational parameters of the elements of the station. Quite general approximate relationships are used to express the costs of the elements of the installation as a function of the parameters being sought. Attention is paid to the mutual relationships which exist among the parameters, through the equations which describe the processes going on in the station, as well as to the limitations which are imposed by the need for reliable operation.The system of equations derived may be used to check the optimum parameters of various projected two-loop installations with nonbolling reactors, employing the maximum allowable fuel element temperatures. Examples are given to show the importance of the independent problem of expressing some of the optimum parameters in terms of other parameters obtained from individual equations of the system.     

Mechanical integrity analysis of TMI-1 OTSG unplugged tubes

Small I.D. circumferential defects have been identified in many steam generator tubes. The origin of the cracks is known to be chemical, not mechanical. A fracture mechanics evaluation has been conducted to ascertain the stability of tube cracks under steady-state and anticipated transient conditions. A spectrum of hypothetical crack sizes was interacted with tube stresses derived from the load evaluation using the methods of linear elastic fracture mechanics (LEFM). Stress intensities were calculated for part-through wall cracks in cylinders combining components due to membrane stress, bending stress, and stresses due to internal pressure acting on the parting crack faces as the loads are cycled.The LEFM computational code, “BIGIF”, developed for EPRI, was used to integrate over a range of stress intensities following the model to describe crack growth in INCO 600 at operating temperature using the equation (ΔK)^3.5.The code was modified by applying ΔK_Th, the threshold stress intensity range. Below ΔK_Th small cracks will not propagate at all. Appropriate R ratio values were employed when calculating crack propagation due to high cycle or low cycle loading.Cracks that may have escaped detection by ECT will not jeopardize tube integrity during normal cooldown unless these cracks are greater than 180° in extent. Large non-through-wall cracks that would jeopardize tube integrity are not expected to evolve because in axi-symmetric tensile stress fields, cracks propagate preferentially through the tube wall rather than around the circumference. Tube integrity can be demonstrated for mid-span tube regions and for the transition region as well.The as-repaired transition geometry is a design no less adequate than the original. The as-repaired condition represents an improvement in the state of stress due to mechanical and thermal loads as compared to the original.