Plastic Deformation and Fracture Behavior of Zircaloy-2 Fuel Cladding Tubes under Biaxial Stress

Various combinations of biaxial stress were applied on five batches of recrystallized zircaloy-2 fuel cladding tubes with different textures; elongation in both axial and circumferential directions of the specimen was measured continuously up to 5% plastic deformation.

The anisotropic theory of plasticity proposed by Hill was applied to the resulting data, and anisotropy constants were obtained through the two media of plastic strain loci and plastic strain ratios. Comparison of the results obtained with the two methods proved that the plastic strain loci provide data that are more effective in predicting quantitatively the plastic deformation behavior of the zircaloy-2 tubes. The anisotropy constants change their value with progress of plastic deformation, and judicious application of the effective stress and effective strain obtained on anisotropic materials will permit the relationship between stress and strain under various biaxialities of stresses to be approximated by the work hardening law.

The test specimens used in the plastic deformation experiments were then stressed to fracture under the same combination of biaxial stress as in the proceeding experiments, and the deformation in the fractured part was measured. The result proved that the tilt angle of the c-axis which serves as the index of texture is related to fracture ductility under biaxial stress. Based on this relationship, it was concluded that material with a tilt angle ranging from 10° to 15° is the most suitable for fuel cladding tubes, from the viewpoint of fracture ductility, at least in the case of unirradiated material.     

Mechanical Properties of Neutron Irradiated Fuel Cladding Tubes

There was few post irradiation examination data on the mechanical properties of domestic fuel cladding tubes used for light water reactors, then those data obtained abroad have been often used in the fuel design or fuel performance codes. Although, many reports discussed the deformation mechanism of the tube, almost all the data were not obtained from irradiated specimens but unirradiated ones. In recent years, systematic post irradiation examinations on domestic fuel elements used in Japanese light water reactors and the related studies were performed.

This report first summarizes briefly the crystallographic texture which characterizes the properties of Zircaloy fuel cladding tubes, followed by an explanation of basic properties such as elasticity, plasticity, creep and fatigue. Finally, the up-to-date results are introduced.     

Development of the Ultra-microhardness Technique for Post Irradiation Examination of Fuel Cladding Tubes

Hardness measurements are potentially valuable for a quantitative discussion of embrittlement in the inner portions of fuel cladding tubes. The size of the indentation, however, is not negligible compared to the measuring region, even when a micro Vickers hardness tester is employed. This limits the measuring technique, and very little has been studied about degradation phenomena in the inner portion of the tubes.

A hardness measurement system, equipped with a depth-sensing indentation instrument, and the necessary post irradiation examination technique for specimens with high radioactivity were successfully developed and the following observations were obtained from the system's application example. The diffusion coefficient of oxygen obtained from the hardness of an unirradiated zirconium lined cladding with simulated oxidation in the fuel rod showed good agreement with literature data. The calculated diffusion coefficient from hardness in the inner portion of irradiated Zircaloy-2 fuel rods was almost the same value as that of unirradiated zirconium, which implied that neither neutron irradiation nor fission fragment bombardment enhanced the oxygen diffusion in the inner portion of cladding tube.     

Effect of Iodine on Fatigue Properties of Zircaloy-2 Fuel Cladding under Cyclic Tensile Stress

Effects of iodine on the cyclic tensile properties of Zircaloy-2 have been investigated at 350°C. Notched specimens, which were grooved circumferentially, and unnotched specimens were machined from the stress-relieved fuel cladding. Cyclic tensile stress was loaded on the specimens in order to get the fatigue life and failure ductility in air and in an Ar atmosphere containing iodine.

The fatigue life and failure ductility of the unnotched specimens, which were pulled parallel to the longitudinal direction of the cladding, showed little decrease due to effects of the iodine environment, though many pittings due to iodine corrosion were observed on the inner surfaces.

On the other hand, the fatigue properties of the notched specimens indicated the effects of iodine. This difference implies that the corrosion fatigue property of Zircaloy-2 fuel cladding under the iodine environment depends on the texture and stress state of the specimens.     

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.     

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.     

Effect of Hydrogen Water Chemistry on Zircaloy-2 Fuel Cladding and Structure Material Performance

The effects of hydrogen addition to the feedwater on the corrosion and hydrogen uptake performance of Zircaloy-2 fuel cladding tubes, a water rod tube and spacer materials irradiated for four cycles in a BWR were evaluated. The uniform oxide behaviors of the cladding tubes, water rod and spacer materials were not affected by hydrogen water chemistry (HWC) condition. The hydrogen uptake and pickup fractions of the water rod and spacer materials were similar to those of water rods and spacer materials under normal water chemistry (NWC) conditions. As for the fuel rods, in spite of comparably heavy crud deposition, their hydrogen uptake and pickup fractions were clearly lower than the values under NWC conditions. Overall, the results indicated that HWC had no adverse effects on fuel performance.     

Effects of Neutron-Irradiation on Tensile Properties of 3.5%Ni Steel

In an aerated soil layer under natural condition, water that infiltrates into it flows discontinuously because of repeated cycles of rainfall and drying. Therefore, column tests, in which 2,000 mi of deionized water was fed into sandy soil layer dried for 3~90 days after it is contaminated by ⁶⁰Co, ⁸⁵Sr and ¹³⁷Cs aqueous solution, were carried out to examine influence of the drying period on migration behavior of those nuclides.

All the radionuclides showed both a steeply decreasing part corresponding to their cationic form and a gradually decreasing part corresponding to their particulate form in the concentration distribution curve along the soil column depth. In the vicinity of the top surface of soil layer, no influence of drying on ⁶⁰Co and ¹³⁷Cs concentrations was observed, but ^S5Sr moved a little into a deeper part from the top surface with longer drying period. Such moving tendency of ^8ESr was found to be connected with the Ca²⁺ concentration in the interstitial water. In the deeper soil layer, no influence of drying on ⁸⁵Sr concentration was observed, but the ⁶⁰Co and ¹³⁷Cs concentrations were increased with the drying period. This is considered to be caused from that each product of the movable ⁶⁰Co(OH)₂ and the ¹³¹Cs adsorbed on the fine particle increased with the rise of pH and the fine particle concentration, respectively, during the drying period. Thus, this study has suggested possible causes affecting on the radionuclide migration under discontinuous flow condition.     

Effect of Neutron Irradiation on Iodine Stress Corrosion Cracking Susceptibility of Zircaloy-2 Tubing

The effect of neutron irradiation on the iodine stress corrosion cracking (SCC) of Zircaloy-2 tubing of 8×8 type design was studied by means of ring tension test, using specimens unirradiated and irradiated to 3.2×l0¹⁹ and 3.0×10²⁰ n/cm² (E>lMeV). The SCC threshold stresses were determined from constant load tests and the SCC initiation stresses were defined from constant cross-head rate tests. Both stresses increased with increasing neutron fluence, reaching a maximum at a neutron fluence between 10¹⁹ and 10²⁰ n/cm² and then decreased. The tendency is qualitatively in good agreement with the hypothetical conclusion derived by Lunde & Videm, for SCC failure stresses from internal gas pressurization tests on various Zircaloy cladding tubes irradiated at different reactor conditions. The cause of the increase in the SCC susceptibility at neutron fluences above 10²⁰ n/cm² would be ascribed to radiation anneal hardening phenomenon and resultant inhomogeneous incipient deformation characterized by dislocation channels.     

Stress Corrosion Cracking of Zircaloy-2 Cladding in Iodine Vapor

Iodine stress corrosion cracking of Zircaloy-2 cladding under static tensile stress has been studied over the temperature range of 250~450°C. The lowest iodine concentration required to cause cracking was 0.15 mg iodine per cm³ free volume in test ampoule at 300°C, and 1.1 mg/cm' at 350°C. The minimum circumferential plastic strain to cause failure was about 0.4% for specimen tubes possessing high cracking susceptibility. Softer tubes tolerate more circumferential deformation before they failed. This is attributed to greater margin retained for work hardening, rather than to smaller residual tensile hoop stress. Cracking was also inhibited by the presence of oxygen in the iodine atmosphere, the threshold partial pressures being 28 torr at 300°C and 9 torr at 350°C. Scanning electron micrographs of the fractured surfaces revealed brittle transgranular fracture carrying river patterns on part of the cleavage facets.     

Thermal Diffusion of Hydrogen in Zircaloy-2 Containing Hydrogen beyond Terminal Solid Solubility

The thermal diffusion of hydrogen is one of causes of uneven hydride precipitation in zircaloy fuel cladding tubes that are used in water reactors. In the diffusion model of hydrogen in zircaloy, the effects of the hydride on the diffusibility of hydrogen has been regarded as negligibly small in comparison with that of hydrogen dissolved in the matrix. Contrary to the indications given by this model, phenomena are often encountered that cannot be explained unless hydride platelets have considerable ostensible diffusibility in zircaloy.

In order to determine quantitatively the diffusion characteristics of hydrogen in zircaloy, a thermal diffusion experiment was performed with zircaloy-2 fuel cladding tubes containing hydrogen beyond the terminal solid solubility. In this experiment, a temperature difference of 20°–30°C was applied between the inside and outside surfaces of the specimen in a thermal simulator.

To explain the experimental results, a modified diffusion model is presented, in which the effects of stress are introduced into Markowitz's model with the diffusion of hydrogen in the hydride taken into account. The diffusion equation derived from this model can be written in a form that ostensibly represents direct diffusion of hydride in zircaloy. The apparent diffusion characteristics of the hydride at around 300°C are D_p = 2.3×10⁵exp(?32,000/RT), (where R: gas constant, T: temperature) and the apparent heat of transport Q _p *=?60,000 cal/mol. The modified diffusion model well explains the experimental results in such respects as reaches a steady state after several hours.     

Effect of Metallographic Factor on Hydrogen Pick-up Properties of Zircaloy-2

Zirconium alloy sheets were prepared with varying Fe, Cr and Ni systematically. The corrosion and hydrogen pickup property were estimated in steam at 673 K, in water at 633 K and in super critical water at 673 K. The effect of the SPP and the oxide film on the hydrogen pick-up was studied from the hydrogen pick-up route using D₂O and the microstructure of the oxide film and secondary phase particle (SPP) in the oxide film. The hydrogen pick-up ratio decreased with increase of Fe and decrease of Ni. It was affected by Fe/Ni ratio of the matrix. The hydrogen pickup was not related to SPP but was related to the oxide film when the oxide film was relatively thick. The tetragonal ZrO₂ is considered to act as a barrier for hydrogen pick-up.     

Effect of Neutron Irradiation on Mechanical Properties of Niobium and Nb-Y Alloy

In a high-temperature gas-cooled reactor core, which consists of prismatic graphite fuel elements, leakage flows of coolant gas occur through gaps between blocks. Since the effects of these leakage flows on the total flow distribution are significant, their flow features must be clarified. In this paper, the leakage flows (crossflow through the interface gap between contacting fuel elements and the permeation flow through the fuel elements) in the normally stacked fuel elements were studied. In the basic experiments, leakage flow rates were measured using small-scale graphite blocks to determine the equivalent interface gap width and the permeability. The experiments using the full-scale fuel element were also carried out and the results agreed well with those of the basic experiments. Furthermore, a simple flow model was devised to predict the leakage flow in the fuel element.     

Neutron Irradiation Effect on Magnetic Properties of Cr-, Mo- and Si-Doped FeNi-Alloys

Sample of FeNi, FeNi: Cr, FeNi: Mo, and FeNi: Si alloys were irradiated by neutron beams in the vicinity of a 2 MW reactor core and their magnetic hysteresis curves, magnetic after effects as well as variations of magnetic permeability with temperature were determined. From these curves Curie temperature shift, spectrum widthk, and activation energy for self-diffusion of Fe and Ni atoms were obtained.

The presence of Si impurities in the FeNi alloy produces a considerable attenuation in the vacancy supersaturation and the analysis of the spectrum width in this sample indicates that the defect structure due to neutron irradiations is more complex than that due to thermal effect.     

Optimized Ring Tensile Test Method and Hydrogen Effect on Mechanical Properties of Zircaloy Cladding in Hoop Direction

Data pertaining to the mechanical properties of the fuel cladding in the hoop direction are required for the analysis of high burn-up fuel behavior under reactivity-initiated accident (RIA) conditions. In the present study, by minimizing undesirable effects of friction and bending, the ring tensile test method was optimized to obtain precise data pertaining to the mechanical properties of the fuel cladding in the hoop direction. The optimized specimen has a single gauge section and is stretched using the tooling consisting of two half-mandrels. The gauge section is set on top of a half-mandrel; this arrangement is unique in comparison with other methods. Using the optimized test method, the mechanical properties of the prehydrided Zircaloy-4 cladding in the hoop direction were evaluated as functions of hydrogen concentration and test temperature. When the hydrogen concentration is below 500 ppm, the decrease in ductility due to hydriding is relatively small at all test temperatures. When the hydrogen concentration is above 600 ppm, the ductility in the tests at 300K remarkably decreases due to hydriding, while the hydrogen effect decreases in magnitude in the tests above 473 K.     

Tensile Properties of llCr-0.5Mo-2W,V, Nb Stainless Steel in LMFBR Environment

The tensile strength of ferritic-martensitic HCr-0.5Mo-2W, Nb, V stainless steel (PNC-FMS), which had been developed for core component applications in LMFBR by Japan Nuclear Cycle Development Institute, was evaluated for the effects of thermal aging, sodium exposure, and neutron irradiation. The tensile strength of thermal aged specimens (~1,023K, ~12,000h) decreased at aging conditions above the initial tempering parameter, and the aging effect was considerably enhanced for the wrapper tubes tempered at lower temperatures. The tensile strength of sodium exposed specimens (~973K, ~10,000h) decreased more than aged specimens due to decarburization, and the effect of decarburization was greater in thin wall cladding tubes. Evaluation of the contribution of both thermal aging and decarburization effects on the tensile strength of cladding tubes irradiated in JOYO (~1,013K, ~6,030h, ~29dpa) suggested that the radiation showed smaller effect on tensile properties than thermal aging and decarburization. By using the derived correlations for thermal aging and decarburization effects, the tensile strength decrease for PNC-FMS after long period (30,000 h) in LMFBR environment was quantitatively calculated.     

Reduction of Plastic Anisotropy of Zircaloy Cladding by Neutron Irradiation, (II)

Initial plastic deformation behavior of zirconium alloy fuel cladding was described quantitatively by the deformation system of single crystal of α-zirconium, and a model was proposed to simulate the yield behavior of polycrystalline material. Based on the model, effects of crystallographic texture and stress state on the plastic deformation of the cladding were evaluated. Conclusions obtained from this investigation are:

(1) The proposed model shows good agreement with the von Mises' yield criteria for a material with isotropic properties.

(2) Plastic anisotropy of the cladding decreases when neutron irradiation affects prism slip more strongly than the other deformation systems.

(3) Dominant deformation systems for axial tension or internal pressurization of the cladding are predicted to be prism slip or tensile twin, respectively, when the stress state of the cladding reaches the yield condition.     

Effects of Temperature,Strain Rate,and Specimen Orientation on Localized Plastic Deformation of Irradiated Zircaloy-2

Both transverse and longitudinal Zircaloy-2 specimens irradiated up to 1.2 × 10²⁰ n/cm² (E> 1 MeV) were tested in tension with strain rates ranging 1.1 × 10^-4~1.1 × 10^-2 s^-1 in the temperature range 200~400°C. Detailed observations of the specimen wall surface and microstructure were also made on samples deformed to various amounts of plastic strain, with a projector and an optical microscope.

It was found that localized plastic deformation bands occurred in the temperature range approximately 280~330°C during straining to the ultimate tensile stress. Results also showed that the strain rate dependence of tensile properties, particularly the strain to the ultimate tensile stress, was associated with changes in the number and width of the localized deformation band with strain rates at a temperature of 300°C at which localized bands occurred. From a break of the straight line tracing the true stress-true plastic strain relationship, it was established that the onset stress and strain of the localized deformation band could be estimated.

The effect of specimen orientation on localized deformation band was also discussed on the basis of differences in the onset stress and strain between the transverse and longitudinal specimens.