Yutaka Udagawa

Effect of Sn and Nb on generalized stacking fault energy surfaces in zirconium and gamma hydride habit planes

We have investigated the effects of Sn and Nb on dislocation properties in a Zr lattice to elucidate the role of these alloying elements in hydride nucleation processes. According to experimental observations, γ-hydride habit planes are close to the prismatic plane in pure Zr and close to the basal plane in Zircaloy. Dislocation loops are observed around hydride precipitates, implying they play a part in hydride formation. Our ab initio generalized stacking-fault energy calculations showed remarkable effects of Sn on unstable-stacking energy and stacking-fault energy: these parameters for basal slip were considerably reduced while those for prismatic slip were increased in the presence of Sn. These results suggest selective stabilization and enhancement of dislocation spreading in the basal plane, promoting possible elementary processes of hydride precipitation with basal habit plane, i.e. screw-dislocation spreading and edge-dislocation emission in the basal plane.

Effect of cooling history on cladding ductility under LOCA conditions

In a Loss-of-Coolant Accident of LWRs, heated and oxidized fuel cladding is slowly cooled and then quenched by re-flooding water. Effects of cooling rate during the slow-cooling process and quench temperature were investigated on post-quench ductility of the oxidized cladding. Unirradiated Zircaloy-4 cladding specimens were oxidized in steam at 1,373 or 1,473 K, cooled at a rate from 2 to 7 K/s, and finally quenched from temperatures ranging 1,073 to 1,373 K. Post-quench ductility was evaluated by ring-compression test, and microscopic properties were examined by metallurgical examination, hardness test, and oxygen analysis. Morphology of oxygen-rich α phase in the metallic prior-β phase layer changed depending on the cooling history. Area fraction of α phase region in the cross section obviously increased and post-quench ductility reduced with decrease in the quenching temperature. Since α phase region with low ductility can be preferential path for crack propagation, increase in area fraction of α phase region possibly decreases resistance for fracture, which results in ductility reduction. On the other hand, the area fraction was nearly constant irrespective of the rate of the slow cooling, and consequently, effect of the cooling rate on the post-quench ductility was negligible.

Stress Intensity Factor at the Tip of Cladding Incipient Crack in RIA-Simulating Experiments for High-Burnup PWR Fuels

RIA-simulating experiments for high-burnup PWR fuels have been performed in the NSRR, and the stress intensity factor K I at the tip of cladding incipient crack has been evaluated in order to investigate its validity as a PCMI failure threshold under RIA conditions. An incipient crack depth was determined by observation of metallographs. The maximum hydride-rim thickness in the cladding of the test fuel rod was regarded as the incipient crack depth in each test case. Hoop stress in the cladding periphery during the pulse power transient was calculated by the RANNS code. K I was calculated based on crack depth and hoop stress. According to the RANNS calculation, PCMI failure cases can be divided into two groups: failure in the elastic phase and failure in the plastic phase. In the former case, elastic deformation was predominant around the incipient crack at failure time. K I is available onlyin this case. In the latter, plastic deformation was predominant around the incipient crack at failure time. Failure in the elastic phase never occurred when K I was less than 17 MPam1/2. For failure in the plastic phase, the plastic hoop strain of the cladding periphery at failure time clearly showed a tendency to decrease with incipient crack depth. The combination of K I, for failure in theelastic phase, and plastic hoop strain at failure, for failure in the plastic phase, can be an effective index of PCMI failure under RIA conditions.

Improved-EDC tests on the Zircaloy-4 cladding tube with an outer surface pre-crack

ABSTRACT In order to investigate the failure behavior of fuel cladding under a reactivity-initiated accident (RIA) condition, biaxial stress tests on unirradiated Zircaloy-4 cladding tube with an outer surface pre-crack were carried out under room temperature conditions by using an improved Expansion-Due-to-Compression (improved-EDC) test method which was developed by Japan Atomic Energy Agency. The specimens with an outer surface pre-crack were prepared by using Rolling-After-Grooving (RAG) method. In each test, a constant longitudinal tensile load of 0, 5.0 or 10.0 kN was applied along the axial direction of specimen, respectively. All specimens failed during the tests, and the morphology at the failure opening of the specimens was similar to that observed in the result of post-irradiation examinations of high burnup fuel which failed during a pulse irradiation experiment. The longitudinal strain (ϵtz) at failure clearly increased with increasing longitudinal tensile loads and the circumferential strain (ϵtϑ) at failure significantly decreased in the case of 5.0 and 10.0 kN tests, compared with the case of 0 kN tests. From these tests, the data of cladding failure were obtained in the range of strain ratio (ϵtz/ϵtϑ) between about −0.6 and 0.7: this range of strain ratio covers the range between about 0.0 and 0.7 which is estimated in the case of RIA-simulated test. It is considered that the data obtained in this study can be used as a fundamental basis for quantifying the failure criteria of fuel cladding under a biaxial stress state.

Simulation of the fracture behavior of Zircaloy-4 cladding under reactivity-initiated accident conditions with a damage mechanics model combined with fuel performance codes FEMAXI-7 and RANNS

A continuum damage mechanics model using FEM calculations was proposed to be applied to an analysis of the fuel failure due to pellet cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions. The model expressed ductile fracture via two processes: damage nucleation related to void nucleation and damage evolution related to void growth and linkage. The boundary conditions for the simulations were input from the fuel performance codes FEMAXI-7 and RANNS. The simulation made reasonable predictions for the cladding hoop strain at failure and reproduced the typical fracture behavior of the fuel cladding under the PCMI loading, characterized by a ductile shear zone in the inner region of the cladding wall. It was shown that occurrence of a through-wall crack is determined at an early stage of crack propagation, and the rest of the through-wall penetration process is achieved with a negligible increment in strain. The effect of a local temperature rise in the cladding inner region on the failure strain was found to be less than 5% for the conditions investigated. Failure strains predicted under a plane strain loading were smaller by 20%–30% than those predicted under equibiaxial tensions between the hoop and the axial directions.

Evaluation of Initial Temperature Effect on Transient Fuel Behavior under Simulated Reactivity-Initiated Accident Conditions

In order to evaluate possible effects of initial temperature on the transient fuel behavior, such as cladding deformation and fission gas release, under reactivity-initiated accident conditions, two comparative pulse-irradiation tests were performed on identical high-burnup PWR fuel rods under different temperature conditions at the Nuclear Safety Research Reactor (NSRR). The test RH-1 was carried out at room temperature of _20_C, while the coolant temperature in the test RH-2 was _280_C corresponding to the hot zero power temperature of PWR. The fuel rods did not fail in both tests against fuel enthalpy increases of 462 and 378 J/g, respectively. The results of the two tests were generally consistent with data previously obtained in a number of tests at room temperature, when the data were plotted as a function of the peak fuel enthalpy, not of the maximum increase in fuel enthalpy. Computer analysis using the RANNS code confirmed that the cladding residual deformation in the test RH-2 was driven only by the pellet thermal expansion and that the gas-induced deformation did not occur because the cladding temperature did not become high enough to enhance creep deformation even in the film boiling on the cladding surface.

Analysis of MOX Fuel Behavior in Halden Reactor by FEMAXI-6 Code

Capabilities of the FEMAXI-6 code to analyze the behavior of high burnup MOX fuels in LWRs have been evaluated. Coolant conditions, detailed power histories and specifications of the MIMAS-MOX fuel rods, rod 10 and rod 11, of IFA-597.4–7 irradiated in the Halden reactor were input, and calculated rod internal pressures and pellet center temperatures were compared with the measured data for the range of 0-31 MWd/kgUO2. Some sensitivity studies were conducted mainly with respect to pellet thermal conductivity and swelling rate to investigate the changes in thermal behavior and their effects on fission gas release. In the irradiation period up to about 23 MWd/kgUO2, the calculated pellet center temperatures sufficiently agreed with the measured data and also the calculated rod internal pressures reproduced the tendency of an increase in the measured rod internal pressures. These results suggest that fission gas release from MOX fuels can be reasonably predicted by a diffusion process that is modeled in UO2 pellet grains. On the other hand, the steep increase in the measured rod internal pressures observed at the power ramp around 23 MWd/kgUO2 cannot be reproduced by FEMAXI-6 and can be regarded as the result of a relatively large amount of gas release, which possibly caused a pellet-cladding-gap closure through pellet gas-bubble swelling.

Deformation behavior of recrystallized and stress-relieved Zircaloy-4 fuel cladding under biaxial stress conditions

ABSTRACT Pellet–cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to the failure of high-burnup fuel rods. Biaxial stress states generated by PCMI in Zircaloy cladding may make the cladding more susceptible to failure. In this study, we investigated the deformation behavior of Zircaloy cladding under biaxial stress conditions based on the concept of contours of equal plastic work. The major axis angles of the initial work contours of recrystallized (RX) and stress-relieved (SR) specimens were investigated and it was found that the shapes of the initial work contours of these kinds of specimens were almost symmetric across the direction where the ratio of axial stress to circumferential stress is 1. The shapes of subsequent work contours tended to change for the RX specimen while being the same as the initial for the SR specimen, as deformation proceeded. It was suggested that the textures and slip systems in the RX and SR specimens affect their initial work contours while the slip system in the RX specimens and the residual strain in the SR specimens influence the subsequent work contours.

Stress biaxiality in high-burnup PWR fuel cladding under reactivity-initiated accident conditions

In order to promote a better understanding of failure mechanisms of high-burnup pressurized water reactor (PWR) fuels under reactivity-initiated accident (RIA) conditions, stress biaxiality in cladding has been estimated for the pellet-cladding (PC) mechanical interaction (PCMI) phase. The estimation was based on an analysis of the transient elongations of a pellet stack and a cladding tube measured in RIA-simulating experiments in the nuclear safety research reactor (NSRR) using the RANNS code. Stress biaxiality in the high-burnup PWR fuel cladding during the PCMI phase has been estimated to be 0.7–0.8, on average, at the mid-wall of the cladding. A comparison with fresh fuel test results and a sensitivity analysis showed that the effects of burnup and pulse width on cladding stress biaxiality are less than 10% for the investigated range. The present analysis also indicated that PC friction is strong, and that the cladding constraint on pellet stack elongation is significant irrespective of burnup. Therefore, it is recommended that strong PC friction be assumed, which is similar to the mechanical bonding condition, and that fuel pellets be treated as deformable materials in models of fuel behavior during the PCMI phase.

Biaxial-EDC Test Attempts With Pre-Cracked Zircaloy-4 Cladding Tubes

When the pellet-cladding mechanical interaction (PCMI) occurs in a reactivity-initiated accident (RIA), the states of stress and strain in the fuel cladding varies in a range depending on the friction and degree of bonding between cladding and pellet. Japan Atomic Energy Agency has developed the improved Expansion-due-to-compression (EDC) test apparatus to investigate the PCMI failure criterion of high-burnup fuel under such conditions. In this study, the failure behavior of cladding tube was investigated by using the improved EDC test apparatus. Cold-worked, stress-relieved and recrystallized Zircaloy-4 tubes with a pre-crack were used as test specimens: this pre-crack simulated the crack which is considered to form in the hydride rim of high-burnup fuel cladding at the beginning of PCMI failure. In the EDC test, a tensile stress in axial direction was applied and displacement-controlled loading was performed to keep the strain ratio of axial/hoop as a constant. The data of cladding deformation had been achieved in the range of strain ratio of 0, 0.25, 0.5 and 0.75 and pre-crack depth of 41–87 micrometers. Failures in hoop direction were observed in all the tested samples, and a general trend that higher strain ratio and deeper crack depth lead to lower failure limit in hoop direction could be seen. Different crack propagation mode was observed between recrystallized and stress relieved and cold worked samples, which might be due to the difference in microstructure caused by the final heat treatment at the fabrication of cladding.Copyright