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Suzuki, Motoe; Saito, Hiroaki*; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.579 - 601, 2005/10
The RANNS code analyzes behaviors of a single fuel rod in reactivity-initiated accident (RIA) conditions. The code has two types of mechanical model; one-dimensional deformation model for each axial segment length of rod, and newly-developed two-dimensional local deformation model for one pellet length. Analyses were performed on the two RIA-simulated experiments in the NSRR, OI-10 and OI-11 with high burnup PWR rods, and results of cladding deformation were compared between calculations by the two models and PIE data. RANNS calculated the deformation profiles of claddings during the power transient of the experiments on the basis of the pre-pulse conditions of rods predicted by FEMAXI-6 code. In the calculations by the two-dimensional model, the plastic strain increase at the cladding ridges was compared with those in between the ridges and with the PIE data, and effect of stress variance induced by local non-uniformity of strain on the crack growth was discussed.
Sugiyama, Tomoyuki; Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.912 - 932, 2005/10
High burnup fuel cladding can fail due to mechanical interaction with expanding fuel pellet under reactivity initiated accident (RIA) conditions. In order to evaluate the cladding failure limit, investigations to modify ring tensile test have been performed to measure mechanical properties of Zircaloy cladding properly. JAERI developed the test method and geometry minimizing undesirable effects of friction and bending moment in the specimen. Using the modified test method, mechanical properties of unirradiated Zircaloy-4 cladding were evaluated as functions of hydrogen concentration and temperature. For hydrogen concentrations above 700 ppm, obvious increase of ductility is observed with the temperature increase from 300 to 473 K. For hydrogen concentrations below 500 ppm, on the other hand, temperature dependence of ductility is relatively small in the present temperature range from 300 to 573 K.
Fuketa, Toyoshi; Sugiyama, Tomoyuki; Sasajima, Hideo; Nagase, Fumihisa
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.633 - 645, 2005/10
LWR fuel behaviors during a reactivity initiated accident (RIA) are being studied in the NSRR program. Results from recent NSRR experiments, no failures in Tests OI-10 and -12 and the higher failure enthalpy in Test OI-11, reflect the better performance of the new cladding materials in terms of corrosion during PWR operations. Accordingly, these rods with improved corrosion resistance have larger safety margin than conventional Zircaloy-4 rods. In addition, the smaller inventory of inter-granular gas in the large grain pellet could reduce the fission gas release in RIA as observed in the OI-10. Test VA-1 was conducted with an MDA sheathed 78 MWd/kgU PWR fuel rod. Despite of the higher burnup and thicker oxide layer of 
81
m, the enthalpy at failure remained in a same level as those for rods with of 40m-oxide at 50 - 60 MWd/kgU. This result suggests high burnup structure (rim structure) in pellet periphery does not have strong effect on the failure enthalpy reduction because the PCMI load is produced primarily by solid thermal expansion of the pellet.
Kudo, Tamotsu; Hidaka, Akihide*; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.883 - 889, 2005/10
The VEGA program have been performed at Japan Atomic Energy Research Institute (JAERI). The program was comprised of series of experiments on radionuclides release from fuel under severe accident conditions and post-test evaluation with numerical calculations. Effects on the release of ambient pressure, fuel temperature, inert or steam environment and MOX-effect were studied in the program. These effects had been hardly investigated in previous studies due to difficulties in experiments with high temperature and pressure conditions. Release of cesium was mitigated at elevated pressure in comparison with atmospheric pressure. Cesium release was enhanced in the temperature region where fuel foaming occurred below the melting point of UO
. Release of cesium and ruthenium under steam condition was greater than that under the inert helium condition. Released mass of plutonium above 2800 K was higher by nearly three orders of magnitude than that in lower temperature than 2800 K.
Nagase, Fumihisa; Fuketa, Toyoshi
Proceedings of 2005 Water Reactor Fuel Performance Meeting (CD-ROM), p.668 - 677, 2005/10
A systematic research program on high burnup fuel behavior under LOCA conditions is being conducted at JAERI. As a part of the program, integral thermal shock tests simulating the whole LOCA sequence were conducted with Zircaloy-4 fuel claddings, irradiated to 39 and 44 GWd/t at a PWR, to investigate behavior and condition of cladding fracture during quenching for safety evaluation. Differences were not clearly observed between irradiated and unirradiated claddings at similar hydrogen concentrations in terms of threshold of fracture during quenching, though the threshold is reduced as initial hydrogen concentration increases. Ductility of pre-hydrided, oxidized and quenched claddings was also evaluated by using ring-tensile and ring-compression tests. Embrittlement criteria (zero-ductility limits) from both the tests were lower than the fracture conditions in the integral thermal shock tests. This indicates that loading conditions should be well simulated to evaluate cladding performance under LOCA conditions.