Fumihisa Nagase

Oxidation kinetics of low-sn Zircaloy-4 at the temperature range from 773 to 1,573 K

Isothermal oxidation tests in flowing steam were performed on low-Sn Zircaloy-4 cladding tubes over the wide temperature range from 773 to 1,573 K in order to obtain oxidation kinetics applicable to various loss-of-coolant accident conditions of LWRs. The oxidation generally obeys a parabolic rate law for the examined time range up to 3,600 s at temperatures from 1,273 to 1,573 K, and for a limited time range up to 900 s from 773 to 1,253 K. A cubic rate law is preferable for evaluating the longer-term oxidation at 1,253 K and below. The parabolic rate law constant and the cubic rate law constant for measured weight gain were evaluated at every examined temperature, and Arrhenius-type equations were determined in order to describe the temperature dependence of the rate constants. It was indicated that the change of the oxidation kinetics from the cubic to the parabolic rate and the discontinuities in the temperature dependence of the rate constants are caused by the monoclinic/tetragonal phase structure change of ZrO2.

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.

Influence of precipitated hydride on the fracture behavior of Zircaloy fuel cladding tube

In order to clarify the influence of precipitated hydride on the fracture behavior of Zircaloy cladding tubes, the stress-strain distribution of the cladding was estimated by finite element method (FEM) analysis. The mechanical properties of α-phase of zirconium and zirconium hydride required for the analysis were examined by means of an ultrasonic pulse-echo method and a tensile test. It was found from the analysis that the non-hydrided cladding has the highest equivalent plastic strain at the inner surface of the cladding, suggesting that the fracture initiated at the inner surface of the cladding. Since the hydride accumulated layer located in the outer surface of the hydrided cladding fails at a lower internal pressure, the crack appears to initiate at the outer surface of the cladding. The fracture behavior estimated from the stress states of the hydrided cladding was in good agreement with the experimental results obtained by pulse irradiation tests of the Nuclear Safety Research Reactor (NSRR) and high-pressurization-rate burst tests in the Japan Atomic Energy Research Institute (JAERI).

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.

Analysis of the fracture behavior of hydrided fuel cladding by fracture mechanics

In order to elucidate the fracture behavior of light water reactor (LWR) fuel rods under reactivity initiated accident (RIA) conditions, an analysis based on fracture mechanics has been performed for several types of hydrided cladding tubes. Fracture mechanics parameters such as stress intensity factor (SIF), J-integral and plastic yield load were estimated by finite element method (FEM) analysis, and the failure assessment diagram (FAD) was constructed using the fracture mechanics parameters to estimate the failure stress of the claddings. It was found from the FAD that the predicted fracture stress of the claddings was qualitatively consistent with the experimental results obtained by pulse irradiation tests in the Nuclear Safety Research Reactor (NSRR) and high-pressurization-rate burst tests at Japan Atomic Energy Research Institute (JAERI).

Behavior of High Burn-up Fuel Cladding under LOCA Conditions

LOCA-simulated experiments were performed with MDA, ZIRLO™, M5®, NDA, and Zircaloy-2 cladding specimens with local burn-ups ranging from 66 to 76 MWd/kg. Short test rods fabricated with the cladding specimens were heated, isothermally oxidized at 1,459 to 1,480K in steam flow, and finally quenched in flooding water. Rod rupture and subsequent double-sided oxidation of the cladding were also simulated in the experiments. Neither split-fracture nor fragmentation occurred during the quench in the cladding specimens which were oxidized to about 18–27% of the metallic thickness. Accordingly, the fracture boundary, a most important safety issue, is not reduced significantly by the high burn-up and use of the new alloys within the examined scope, although it may be somewhat reduced with pre-hydriding during the reactor operation as observed in unirradiated specimens.

Behavior of 60 to 78MWd/kgU PWR Fuels under Reactivity-Initiated Accident Conditions

To provide a data base for the regulatory guide of light water reactors, behavior of reactor fuels during off-normal and postulated accident conditions such as reactivity-initiated accident (RIA) is being studied in the Nuclear Safety Research Reactor (NSRR) program of the Japan Atomic Energy Agency (JAEA). The paper presents recent results obtained from the NSRR power burst experiments with high burnup fuels, and discusses effects of pellet expansion as PCMI (Pellet-Cladding Mechanical Interaction) loading and cladding embrittlement primarily due to hydrogen absorption. Results from the recent four experiments on high burnup (about 60 to 78 MWd/kgU) PWR UO2 rods with advanced cladding alloys showed that the fuel rods with improved corrosion resistance have larger safety margin against the PCMI failure than conventional Zircaloy-4 rods. The tests also suggested that the smaller inventory of inter-granular gas in the pellets with the large grain could reduce the fission gas release during the RIA transient; and high burnup structure in pellet periphery (so-called rim structure) does not have strong effect on reduction of the failure threshold because the PCMI load is produced primarily by solid thermal expansion.

Analysis of the fracture behavior of a hydrided cladding tube at elevated temperatures by fracture mechanics

An analysis, based on fracture mechanics at elevated temperatures, has been carried out for several types of hydrided Zircaloy cladding tubes to elucidate the fracture behavior of high burn up light water reactor fuel cladding during reactivity initiated accidents. Fracture mechanics parameters such as stress intensity factor, J-integral and plastic yield load were estimated by a finite element method analysis, and the material properties of α-phase of zirconium required for the analysis were obtained by tensile tests at elevated temperatures. The failure assessment diagram (FAD) was constructed using the fracture mechanics parameters to estimate the failure stress of the cladding. It was found from the FAD that the predicted failure stress of the cladding qualitatively agreed with the experimental results obtained by burst tests at elevated temperatures for the hydrided Zircaloy cladding tube at Japan Atomic Energy Research Institute.

Chemical interactions between B4C and stainless steel at hightemperatures

Abstract With a view to investigating the chemical interactions between B 4 C and type 304 stainless steel, the reaction couples ofthese two materials were isothermally annealed in the temperature range between 1073 and 1623 K. As a result of the chemical interactions, complicated reaction layers were formed at the interface of the reaction couple. To evaluate the reaction kinetics, the decrease in the thickness of stainless steel and the reaction layer growth were measured as a function of temperature and time. The overall reaction generally obeyed the parabolic rate law. Both a parabolic rate law constant and an apparent activation energy were determined. A discontinuity in the temperature dependence of the parabolic rate constants was found in the temperature range between 1473 and 1498 K. This corresponds to the formation of the liquid phase at the reaction interface.