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High creep strain experiments have been carried out under irradiation using cold worked Zircaloy-2, cold-worked Zr-Nb, and heat treated Zr-Nb 0.6-in.-diameter tubes in DFR, DMTR, and PLUTO reactor. The tests in PLUTO were found to be the most useful since full creep curves were obtained using neutron radiography techniques. Creep strains of up to 9 percent were obtained on cold-worked Zircaloy-2 tubes without failure, thus substantiating the recommended pressure tube creep strain limit of 3 percent. Up to 5 percent creep strain was observed with cold-worked Zr-Nb tubing, but since one tube failed at only 1.8 percent strain it is not yet possible to specify a realistic yet safe creep limit for this material.
The objective of this work was to study the irradiation-thermal creep behavior of cladding tubes made of two Russian zirconium alloys: the widely used E110 (Zr-1Nb) alloy and the advanced E635 (Zr-1Nb-1.2Sn-0.35Fe); and to demonstrate the advantages of E365 claddings for VVER fuel rod application with respect to fuel pellet-cladding gap at high burnups. For this purpose, a creep test concept and an analysis methodology were developed. Creep tests using cladding tubes made from both alloys were carried out. The test results along with published data allowed development of predictive models of irradiation-thermal creep for the cladding tubes of both alloys. The models describe strain processes in non-irradiated claddings and in claddings under and after irradiation. The models benchmark well to dimensional changes in VVER fuel rods. The models show the fuel pellet-cladding gap to disappear at higher burnup values if E635 is used for VVER cladding as opposed to the E110 alloy.
Zirconium alloys are commonly used in pressurized water reactor as fuel rod cladding tubes. After irradiation and cooling in pool, the spent nuclear fuel assemblies are either transported for wet storage to a devoted site or loaded in suitable casks for dry storage on a nuclear power plant site. During dry transportation or at the beginning of dry storage, at temperatures around 400°C, the cladding experiences a creep deformation under the hoop stress induced by the internal pressure of the fuel rod. During creep, a recovery of the radiation damage can occur that can affect the subsequent mechanical properties. The mechanical behavior of the cladding has been investigated in laboratory on two neutron irradiated cladding materials: fully recrystallized Zr-1 % Nb and stress-relieved Zircaloy-4. Creep tests under internal pressure were conducted at 400 and 420°C. After depressurization and cooling, ring tensile tests were carried out at room temperature. In addition, transmission electron microscopy observations have been performed after testing. The post-creep mechanical response exhibited a decrease of the strength compared to the as-irradiated material. This decrease is associated with a significant recovery of the ductility, which becomes close to the ductility of the unirradiated material. The transmission electron microscopy examinations, conducted on recrystallized Zr-1 % Nb ring samples, revealed that the radiation defects have been annealed. It was also observed that as for the unirradiated material, the deformation occurred homogeneously throughout the grains. No dislocation channeling was observed contrary to the as-irradiated material. These observations explain the recovery of the strength and of the ductility after post-irradiation creep that may also occur during dry transportation or at the beginning of dry storage.
Structures developed in zirconium alloys during irradiation creep have been characterized by transmission electron microscopy (TEM). Alloys investigated were annealed Zircaloy-2, cold-worked Zircaloy-2 and cold-worked Zr-2.5Nb pressure tube material. Thin films were taken from material deformed in the NRU, NRX and Pickering-3 reactors at temperatures of 530 to 600 K under stresses of 117 to 552 MPa giving strains in the range 0.14 to 8.8 percent. Stress-induced orientation of dislocation loops makes a negligible contribution to irradiation creep at all stresses. At the lower stresses (and hence strains), the size and distribution of the damage is unaffected by stress, being the same in the head and gage sections of creep specimens. At higher stresses (strains), there is much clearing of the damage by plastic deformation. The deformation however is very uneven, producing structures in different grains of the same specimen that can show no deformation, swaths cleared of irradiation damage, or dislocation tangles or cell formation. The relevance of these TEM observations to irradiation creep mechanisms is discussed.