Light water reactor fuel response during reactivity initiated accident experiments

Abstract

Experimental results from six recent Power Burst Facility (PBF) reactivity initiated accident (RIA) tests are compared with data from previous Special Power Excursion Reactor Test (SPERT), and Japanese Nuclear Safety Research Reactor (NSRR) tests. The RIA fuel behavior experimental program recently started in the PBF is being conducted with coolant conditions typical of hot-startup conditions in a conmercial boiling water reactor. The SPERT and NSRR test programs investigated the behavior of single or small clusters of light water reactor (LWR) type fuel rods under approximate room temperature and atmospheric pressure conditions in capsules containing stagnant water. As observed in the SPERT-and NSRR tests, energy deposition, and consequent enthalpy increase in the PBF test fuel, appears to be the single most important variable. However, the consequences of failure at boiling water hot-startup system conditions appear to be more severe than previously observed in either the stagnant capsuls SPERT or NSRR tests. Metal 1ographic examination of both previously unirradiated and irradiated PBF fuel rod cross sections revealed extensive variation in cladding wall thicknesses (involving considerable plastic flow) and fuel shattering along grain boundaries in both restructured and unrestructured-tu&\-cesLJons. Oxidation of the cladding resulted in fracture at the location of cladding thinning and disintegration of the rods during quench. In addition, swelling of the gaseous and potentially volatile fission products in previously irradiated fuel resulted in volume increases of up to 180% and blockage of the coolant channels within the flow shrouds surrounding the fuel rods.

Reactivity Initiated Accident Test Series RIA Scoping Test Experiment Predictions

“…Results from these tests have shown that the most important aspect of an RIA or an RIA test, is the magnitude of energy deposited into the fuel [ 7 ]. In RIA experiments of the type carried out in SPERT-CDC and NSRR, the energy deposited near the outside edge of the pellet is the primary heat source for cladding melting, while the energy deposited in the interior of the fuel is not conducted to the cladding surface until after the maximum cladding surface temperature is reached.…”

Thermal Shock Fracture of Silicon Carbide and Its Application to LWR Fuel Cladding Performance During Reflood

“…This is an indicator that major failure modes for SiC cladding would be principally different from failure modes of the Zr cladding stated in the U.S Code of Federal Regulation, Title 10, Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems (ECCS) for Light-Water Nuclear Power Reactors" (10 CFR 50.46) [1,12,13] It is worth noting how pervasive the effects of cladding oxidation are in the establishment of the current U.S NRC LOCA criteria. Indeed, the fundamental mechanism of cladding embrittlement of zircaloy during LOCA is due to micro-structural changes of the cladding with oxidation [14,15]. That is, the oxidized cladding cross section exhibits an oxide layer, an oxygen stabilized alpha-phase layer, and a region of prior beta-phase.…”