PULSAR+: A program for calculating the behavior of fuel elements in a nuclear fuel assembly in accidents with coolant losses and reactivity spikes

Abstract

Loss-of-coolant accidents are characterized by an abrupt deterioration of heat transfer from fuel elements. As a result of residual energy release rapid heating ensues and the yield stress of the fuel-can material decreases. When the loop develops a leak the pressure in the reactor core drops while the pressure in the fuel can rises as the fuel element heats up. The difference of the gas pressure in the fuel can and that of the coolant sets up tensile stresses in the fuel can and when a certain temperature is reached the can bulges rapidly as a result of plastic and viscous strains.The opposite situation may arise if the reactor reactivity and, hence, the load on the fuel elements increase abruptly. A sudden rise of fuel temperature is not always accompanied by an instantaneous release of gas from the fuel. If the gas pressure is lower than the coolant pressure, the fuel cans are crumpled. Experimental investigations have shown that fuel-can rupture is usually accompanied by local strains at the site of the rupture, both axisymmetric and asymmetric. The shape of the bulge substantially affects the local hydraulic resistance. Furthermore, when the fuel can bursts depends on the degree of oxidation of the cans by steam.Simulation of the behavior of one fuel element in an accident situation has been the subject of many studies (e.g., [1]). The attempts to consider the behavior of a fuel assembly as a whole have been fairly arbitrary. The assumptions about the similarity of the deformation pattern led to much more conservative results [2].The purpose of the computational code PULSAR + is to simulate the behavior of the fuel can of one fuel element or fuel assembly under the conditions of a loss-of-coolant accident (LOCA) or reactivity increase accident (RIA). The objects simulated may be individual fuel elements or fuel assemblies of the VVI~R--440, VVI~R-1000, RBMK-1000, and RBMK-1500 reactors. Fuel assemblies of other reactors could also be considered. For this purpose a library of fuel-can properties (in this case if they differ from the properties of the alloy Zr-1% Nb) must be filled and the configuration and geometric characteristics of the fuel assembly must be specified.Not all processes that occur in a fuel element during an accident are elucidated in this article. Dispersion, melting, and fragmentation of fuel pellets, and fuel-can embrittlement and failure when the fuel element is cooled down again are the subject of separate research.The main simulated characteristics include the temperature or the time of failure of the fuel element (fuel elements), changes in fuel-can geometry, the mass (thickness of the layer) of steam-oxidized zirconium in each fuel element, and the change in the flow-passage section of the fuel assembly.Simulation of the behavior of fuel cans considers the following processes: unsteady-state heat transfer to and from the fuel element; thermoelastic, plastic, and viscous deformation of the fuel can (high-temperature creep); mechanical interaction of the can with the fuel when crumpled...