Recent view to the results of pulse tests in the IGR reactor with high burn-up fuel
- Russian Research Centre, Moscow (Russian Federation)
Testing of 43 fuel elements (13 fuel elements with high burn-up fuel, 10 fuel elements with preirradiated cladding and fresh fuel, and 20 non-irradiated fuel elements) was carried out in the IGR pulse reactor with a half width of the reactor power pulse of about 0.7 sec. Tests were conducted in capsules with no coolant flow and with standard initial conditions in the capsule of 20{degrees}C and 0.2 MPa. Two types of coolant were used: water and air. One purpose of the test program was to determine the thresholds and mechanisms of fuel rod failure under RIA conditions for VVER fuel rods over their entire exposure range, from zero to high burn-up. These failure thresholds are often used in safety analyses. The tests and analyses were designed to reveal the influence on fuel rod failure of (1) the mechanical properties of the cladding, (2) the pellet-to-cladding gap, (3) fuel burn-up, (4) fuel-to-coolant heat transfer, and other parameters. The resulting data base can also be used for validation of computer codes used for analyzing fuel rod behavior. Three types of test specimens were used in the tests, and diagrams of these specimens are shown in Fig. 1. {open_quotes}Type-C{close_quotes} specimens were re-fabricated from commercial fuel rods of the VVER-1000 type that had been subjected to many power cycles of operation in the Novovoronezh Nuclear Power Plant (NV NPP). {open_quotes}Type-D{close_quotes} specimens were fabricated from the same commercial fuel rods used above, but the high burn-up oxide fuel was removed from the cladding and was replaced with fresh oxide fuel pellets. {open_quotes}Type-D{close_quotes} specimens thus provided a means of separating the effects of the cladding and the oxide fuel pellets and were used to examine cladding effects only.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research; Brookhaven National Lab. (BNL), Upton, NY (United States)
- OSTI ID:
- 269706
- Report Number(s):
- NUREG/CP-0149-Vol.1; CONF-9510156-Vol.1; ON: TI96009154; TRN: 96:016812
- Resource Relation:
- Conference: 23. water reactor safety information meeting, Bethesda, MD (United States), 23-25 Oct 1995; Other Information: PBD: Mar 1996; Related Information: Is Part Of Twenty-third water reactor safety information meeting: Volume 1, plenary session, high burnup fuel behavior, thermal hydraulic research. Proceedings; Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)]; PB: 265 p.
- Country of Publication:
- United States
- Language:
- English
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