Toyoshi Fuketa

Behavior of High-Burnup PWR Fuels with Low-Tin Zircaloy-4 Cladding Under Reactivity-Initiated-Accident Conditions

Experimental programs on fuel behavior during simulated reactivity-initiated-accident (RIA) conditions at the Nuclear Safety Research Reactor (NSRR) in Japan and the CABRI test reactor in France appear to indicate that cladding failures may occur at enthalpy values lower than would be expected. Results from two experiments designated as HBO-1 in NSRR and REP Na-1 in CABRI indicate that the occurrence of fuel failure is strongly influenced by corrosion of cladding in the tested fuels. However, data had been limited to fuel rods with conventional (1.5% Sn) Zircaloy-4 cladding. Results are described from newly conducted NSRR experiments, TK test series, for 38 to 50 MWd/kg U pressurized water reactor fuels with low-tin (1.3% Sn) Zircaloy-4 cladding, and anticipated processes of fuel behavior during the transient are discussed.

Fuel failure and fission gas release in high burnup PWR fuels under RIA conditions

Abstract To study the fuel behavior and to evaluate the fuel enthalpy threshold of fuel rod failure under reactivity initiated accident (RIA) conditions, a series of experiments using pulse irradiation capability of the Nuclear Safety Research Reactor (NSRR) has been performed. During the experiments with 50 MWd/kg U PWR fuel rods (HBO test series; an acronym for high burnup fuels irradiated in Ohi unit 1 reactor), significant cladding failure occurred. The energy deposition level at the instant of the fuel failure in the test is 60 cal/g fuel, and is considerably lower than those expected and pre-evaluated. The result suggests that mechanical interaction between the fuel pellets and the cladding tube with decreased integrity due to hydrogen embrittlement causes fuel failure at the low energy deposition level. After the pulse irradiation, the fuel pellets were found as fragmented debris in the coolant water, and most of these were finely fragmented. This paper describes several key observations in the NSRR experiments, which include cladding failure at the lower enthalpy level, possible post-failure events and large fission gas release.

Investigation of Hydride Rim Effect on Failure of Zircaloy-4 Cladding with Tube Burst Test

To promote a better understanding of failure behavior of high burnup PWR fuel rods during reactivity initiated accidents (RIAs), tube burst tests have been performed with artificially hydrided Zircaloy-4 specimens at room temperature and at 620 K. Pressurization rate was increased to a maximum of 3.4 GPa/s in order to simulate rapid pellet/cladding mechanical interaction (PCMI) that occurs in high burnup fuel rods during a pulse-irradiation in the Nuclear Safety Research Reactor (NSRR). Hydrogen content in the specimens ranged from 150 to 1,050 ppm. Hydrides were accumulated in the cladding periphery and formed ‘hydride rim’ (radially-localized hydride layer) as observed in high burnup PWR fuel claddings. The hydrided cladding tubes failed with an axial crack at the room temperature tests. Brittle fracture appeared in the hydride rim, and failure morphology was similar to that observed in the NSRR experiments. The hydrides rim obviously reduced burst pressure and residual hoop strain at the tests. The residual hoop strain was very small even at 620K when thickness of the hydride rim exceeded 18% of cladding thickness. The present result accordingly indicates an important role of the hydrides layer in high burnup fuel rod failure under RIA conditions.

Behavior of pre-hydrided Zircaloy-4 cladding under simulated LOCA conditions

To promote a better understanding of high burn-up fuel rod behavior in a loss-of-coolant accident, laboratory-scale experiments were performed varying sample and test conditions with non-irradiated Zircaloy-4 claddings. Short test rods, fabricated with claddings having a wide range of hydrogen concentrations (about 100 to 1,450 ppm), were heated, isothermally oxidized at 1,220 to 1,500 K in steam flow, and quenched in flooding water. Axial shrinkage of the rods during the quench was restrained, controlling the maximum restraint load at four different levels. Test rods ruptured during the heat-up, and slight hydrogen concentration effects were seen on rupture temperature and strain. Depending primarily on the oxidized fraction of the cladding thickness, a part of claddings sustained circumferential cracking and fractured into two pieces during the quench. The fracture/no-fracture threshold of the oxidized fraction decreases as both the initial hydrogen concentration and axial restraint load increase. Consequently, when the restraint load is below 535 N, the fracture threshold is higher than 20% cladding oxidation, irrespective of the hydrogen concentration. This is sufficiently higher than the limit in the Japanese ECCS acceptance criteria.

Effect of pre-hydriding on thermal shock resistance of Zircaloy-4 cladding under simulated loss-of-coolant accident conditions

Experiments simulating loss-of-coolant accident (LOCA) conditions were performed to evaluate the effect of pre-hydriding on the thermal-shock resistance of oxidized Zircaloy-4 cladding. Test rods fabricated with 580-mm long claddings were isothermally oxidized at temperatures ranging from 1,220 to 1,530 K in steam and then were quenched with flooding water. Both artificially hydrided (400 to 600ppm) and non-hydrided claddings were subjected to these tests. Since cladding fracture on quenching primarily depends on the amount of oxidation, the fracture threshold was evaluated in terms of “equivalent cladding reacted (ECR).” Under an axially unrestrained condition, the fracture threshold is about 56% ECR, and the influence of pre-hydriding is not observed. The fracture threshold is decreased by restraining the test rods on quenching, and it is more remarkable in pre-hydrided claddings than in non-hydrided claddings. Consequently, the fracture threshold was about 20% ECR and 10% ECR for non-hydrided and pre-hydrided claddings, respectively, under the fully restrained condition. These results indicate a possible decrease of fracture threshold of high burn-up fuel claddings under LOCA conditions.

Failure Thresholds of High Burnup BWR Fuel Rods under RIA Conditions

Transient deformation of high burnup boiling water reactor (BWR) fuel rods was measured and failure limit was examined under simulated reactivity-initiated accident (RIA) conditions. Brittle cladding failure occurred at a small hoop strain of about 0.4% during an early phase of the pulse irradiation tests at the Nuclear Safety Research Reactor (NSRR). Strain rates were in an order of tens %/s at the time of the failure. Comparison of the results with thermal expansion of pellets suggested that the cladding deformation was caused by thermal expansion of the pellets. In other words, the influence of fission gases in the pellets was small in the early phase deformation. Separate effect tests were conducted to examine influence of the cladding temperature on the cladding failure behavior. Influence of the pulse width on the failure threshold was discussed in terms of the strain rate, magnitude of the deformation and temperature of the cladding for high burnup BWR fuel rods under the RIA conditions.

Influence of Hydride Re-orientation on BWR Cladding Rupture under Accidental Conditions

Hydride precipitation along the radial-axial plane increases in high burn-up boiling water reactor (BWR) fuel claddings. The radially-oriented hydrides may have an important role during fuel behavior in a reactivity-initiated accident and may reduce ductility of the cladding under pellet-cladding mechanical interaction (PCMI) conditions. In order to promote a better understanding of the influence of the radial hydrides on cladding failure behavior under the PCMI conditions, tube burst tests were conducted for unirradiated BWR claddings charged with 200 to 650 ppm of hydrogen. About 20 to 30% of hydrides were re-oriented and precipitated along the radial-axial plane. The claddings exhibited large rupture openings with an axial crack at room temperature and 373 K. The crack penetrated through cladding wall preferentially along the radial hydrides, and radial cross section showed cladding failure in a brittle manner. However, reduction in residual hoop strain by precipitation of the radial hydrides was very small. It is accordingly expected that ductility of high burn-up BWR cladding is significantly reduced not only by precipitation of radial hydrides as far as hydrogen concentration and radial hydride fraction range in the present study.

Effect of Cladding Surface Pre-oxidation on Rod Coolability under Reactivity Initiated Accident Conditions

The effect of cladding surface pre-oxidation on the rod coolability under reactivity initiated accidents (RIAs) was investigated. NSRR experiments on irradiated fuel rods have shown higher rod coolability than that of fresh rods, which arose from suppressed DNB and early quench at the surface. To identify the dominating factor, possible factors such as pellet cracking, porosity increase and so on, were assessed. The most probable factor, i.e., the effect of cladding surface pre-oxidation, was examined by pulse irradiation experiments on fresh rods with three cladding surface conditions, no oxide layer, 1 μm and 10(o.m-thick oxide layers. Temperature measurements showed the DNB thresholds in terms of cladding temperature and fuel enthalpy increase at the pre-oxidized surface. The cladding temperature at quench also rises at the pre-oxidized cladding, leading to a reduced film boiling duration. These shifts of the critical and minimum heat flux points could be caused by the surface wettability increase. The test results indicate the dominating factor is the wettability change probably due to the surface chemical potential change by pre-oxidation rather than the thermal conductivity change in the oxide layer, because the results do not depend on the oxide layer thickness, but on the presence of the surface oxide.

Proposal of Simplified Model of Radionuclide Release from Fuel under Severe Accident Conditions Considering Pressure Effect

The VEGA tests on radionuclides release from fuel under severe accident conditions showed that the Cs release rate at 1.0 MPa decreased by about 30% compared with that at 0.1 MPa. To explain this pressure effect, a numerical release model that considers the lattice diffusion in grains followed by the gaseous diffusion in open pores was developed. However, this model is not practical for the PSA analyses due to much computation time and therefore a simplified model called CORSOR-M with the release rate coefficient multiplied by (P≧1 atm) was derived from the numerical model. The multiplier comes from the pressure dependency of gaseous diffusion flux in pores at the pellet surface. The effect of pressure on source term was also estimated for a transient sequence at BWR with JAERIs THALES-2 code in which the simplified model was incorporated. Since the adequacy and applicability of CORSOR-M model were confirmed for the pressures up to 16 MPa through comparison with the VEGA tests and mechanistic models, it is proposed that the model be used for the source term analyses.

Fission gas release in irradiated UO2 fuel at burnup of 45 GWd/t during simulated reactivity initiated accident (RIA) condition

Pulse irradiation simulating reactivity-initiated-accident (RIA) condition was conducted for test rod prepared from fuel irradiated in a commercial reactor and fission gas release behavior from fuel pellet during pulse irradiation was investigated. Optical microscopy (OM) and scanning electron microscopy (SEM) observations and electron probe micro analysis (EPMA) were carried out for the test rod as a part of destructive tests after the pulse irradiation. Fission gas release during pulse irradiation was evaluated by EPMA and puncture test. Xenon depression was observed in the fuel pellet after pulse irradiation at periphery and center region. It is considered that the xenon depression at pellet periphery was due to rim structure formation and fission gas was mainly released from the pellet center region during pulse irradiation. Comparing the fractional fission gas release during pulse irradiation with other results of out-of-pile annealing tests, it was concluded that most fission gas, which was accumulated at grain boundary during base irradiation, was released from the center region of test fuel pellet during pulse irradiation.