LOCA simulation tests were performed in the QUENCH facility of KIT on fuel rod bundle scale. The first two tests using Zircaloy-4 claddings, out of a series of six tests with different cladding alloys, were already performed. The test conditions and results are described. The secondary hydrogenation of the Zircaloy-4 cladding tubes was investigated by means of neutron imaging. In the cladding of the inner rods, hydrogen enriched bended bands were found. They are non-symmetrical to the tube axis. The bands are located at the position where significant inner oxidation ends. X-ray diffractometry (XRD) measurements show that the hydrogen remains at least partially in the zirconium lattice. The formation of the hydrogen enriched bands results in an embrittlement of the cladding tubes and in changes of the fracture mode in tensile tests. The micro-hardness of the cladding increases at the band position. The reasons of the formation of these hydrogen enriched bands and the dependence of its appearance on the temperature scenario are discussed.

Previous papers pointed out the influence of long-term service exposures on the thermal-mechanical behavior of Zr alloys in LOCA conditions and, especially, the impact of in-service hydrogen pick-up on post-quench mechanical properties. Moreover, the oxide layer grown under in-service conditions was occasionally expected to have a protective effect against high temperature oxidation. Finally, the oxygen and hydrogen distributions within the prior-? layer appear as a key parameter with regard to the residual ductility of the alloy, especially as a function of the cooling scenario. The objective of the study presented here was to further investigate the influence of these parameters on the post-quench mechanical properties. Unirradiated Zircaloy-4 and M5® cladding tubes were consequently hydrided up to different concentration levels, then oxidized at high temperature (1000-1200°C) up to at least 10 % measured equivalent cladding reacted (ECR) and directly quenched to room temperature (RT). Ring compression tests (RCT), 3-point bending tests (3PBT) at RT and 135°C, as well as impact tests at RT were then performed to determine the evolution of the post-quench mechanical properties of Zircaloy-4 and M5® alloys with H content. Similarly, specimens preoxidized out-of-pile were also submitted to high temperature oxidation and direct quench, as well as to post-quench ring compression tests. Along with calculations of oxygen diffusion in the metal, results from those tests allowed us to estimate the assumed protective effect of the pretransient oxide layer. Finally, using specimens in the as-received condition or hydrided to typical end-of-life H contents, the effect of temperature history after oxidation at 1200°C was studied, i.e., at the end of the high temperature isothermal oxidation, samples were either submitted to direct quenching to RT or to slow cooling to different final quenching temperatures. It was thus demonstrated that the cooling scenario has a significant impact on the post-quench mechanical properties. All test samples were investigated by means of fractographic examinations to assess the type of failure mode. Moreover, a deep metallurgical analysis has been performed: SEM and image analysis were used for accurate phase thickness measurements, nuclear and electron microprobes for quantitative mapping of hydrogen and oxygen. It proved that the oxygen and hydrogen contents and their distribution in the prior-? layer have a first-order influence on the residual ductility. From all the results obtained on as-received and hydrided samples directly quenched from the oxidation temperature, it was then possible to derive a relationship between structural parameters, i.e., oxygen and hydrogen contents and thickness of the prior-? layer, and the post-quench impact properties at RT.

The effect of fuel burnup on the embrittlement of various cladding alloys was examined with laboratory tests conducted under conditions relevant to loss-of-coolant accidents (LOCAs). The cladding materials tested were Zircaloy-4, Zircaloy-2, ZIRLO, M5, and E110. Tests were performed with specimens sectioned from as-fabricated cladding, from prehydrided (surrogate for high-burnup) cladding, and from high-burnup fuel rods which had been irradiated in commercial reactors. The tests were designed to determine for each cladding material the ductile-to-brittle transition as a function of steam oxidation temperature, weight gain due to oxidation, hydrogen content, pre-transient cladding thickness, and pre-transient corrosion-layer thickness. For short, defueled cladding specimens oxidized at 1000-1200 C, ring compression tests were performed to determine post-quench ductility at (less-than or equal to) 135 C. The effect of breakaway oxidation on embrittlement was also examined for short specimens oxidized at 800-1000 C. Among other findings, embrittlement was found to be sensitive to fabrication processes--especially surface finish--but insensitive to alloy constituents for these dilute zirconium alloys used as cladding materials. It was also demonstrated that burnup effects on embrittlement are largely due to hydrogen that is absorbed in the cladding during normal operation. Some tests were also performed with longer, fueled-and-pressurized cladding segments subjected to LOCA-relevant heating and cooling rates. Recommendations are given for types of tests that would identify LOCA conditions under which embrittlement would occur.

The creep behavior of Zircaloy cladding materials depends on materials texture, degree of recrystallization, and chemical composition. This study is devoted mainly to the analysis of the effect of the final annealing (i.e., the degree of recrystallization) on the creep characteristics. For this purpose, data from a series of thermal creep tests are presented and evaluated. In addition, the in-reactor creep data presented by Franklin et al. are used to evaluate the effect of irradiation on cladding creep performance. The out-of-reactor tests are performed under internal pressurization, and the test matrix covers seven conditions with temperatures from 330 to 400°C and hoop stresses between 80 and 160 MPa. Three lots of Zircaloy-2 claddings and one lot of Zircaloy-4 are considered. The difference between the three Zircaloy-2 lots is in their final annealing conditions. The claddings are either stress relief annealed (SRA), recrystallization annealed (RXA), or partially recrystallization annealed (PRXA). The materials used when fabricating the Zircaloy-2 claddings are from the same ingot, and the chemical compositions of the three types of claddings are almost identical. The Zircaloy-4 cladding included in the test is SRA, and the tin content in this material is similar to that in the Zircaloy-2 materials. The creep data are analyzed by separating the primary (transient) and the secondary (steady-state) creep. In this analysis, the Matsuo creep model, which accounts for both primary and secondary creep, is modified, calibrated, and verified using the new thermal creep data. Based on in-reactor data, the thermal creep model is extended to cover also the creep behavior under irradiation. The claddings considered in the in-reactor test were of both SRA and RXA types, and the experiments were made under external pressure. It is observed that for moderate hoop stresses (

To estimate the degree of conservatism of the present embrittlement criteria, the following examinations have been conducted. Zircaloy-4 tubes were oxidized at temperatures ranging from 1173 to 1573 K, and then compressed at 373 K to determine the embrittlement due to oxygen absorption. The simulated Zircaloy-clad fuel rods were burst and oxidized in steam at temperatures within the range 1200 to 1500 K. The segment sectioned from each burst cladding was also compressed to determine the effect of hydrogen absorption on cladding embrittlement. The results obtained from these experiments have revealed that the additional embrittlement due to hydrogen absorption is found to be significant in burst cladding.

To establish the mechanical response of Zircaloy cladding under thermal shock conditions typical of hypothetical loss-of-coolant accident (LOCA) situations in light-water reactors (LWRs), cladding specimens were ruptured in steam during transient heating (10 K/s), oxidized at maximum temperatures between 1140 and 1770 K for various times, and cooled from the isothermal oxidation temperature to ~1100 K at a rate of 5 K/s, and rapidly quenched by bottom flooding with water at a rate of ~0.05 m/s. Failure "maps" for fracture of the cladding by thermal shock were developed relative to the maximum oxidation temperature and various time-dependent oxidation parameters. In situ pendulum-load impact tests were conducted at room temperature on tubes that survived the thermal quench. Information on the total absorbed energy from these tests was correlated with more extensive results from instrumented drop-weight impact tests. The thermal shock results indicate that the present Zircaloy embrittlement criterion (that is, a total oxidation limit of 17 percent of the wall thickness and a maximum cladding temperature of 1477 K) is conservative and that a more quantitative criterion, based upon the mechanical behavior of the oxidized material, can be formulated with a specified degree of conservatism consistent with the mechanical loads imposed on the cladding during reflood and the maximum amount of oxidation set by the margin of performance of emergency core-cooling systems in LWRs.