Fuel Behavior during Large Breaks in the Primary Heat Transport Circuit

Abstract

A large break in the Primary Heat Transport System (PHTS) is considered each break with a size greater than the largest feeder diameter. The break discharges coolant to the containment, causing PHTS depressurization and consequent increase of temperature and pressure inside containment. The PHTS depressurization induces coolant voiding and, due to the positive reactivity void coefficient, power increases until reactor shuts down on a neutronic or a process-conditioned trip parameter. During the power pulse phase, due to degraded fuel cooling, the sheath may fail. The heat transport system flow decreases faster in the core pass downstream the break. Some channels in the broken loop may become steam filled and others can experience stratified two-phase flow, exposing some fuel elements to steam cooling, inducing fuel temperature rises. A rise in fuel temperature increases the internal fuel element gas pressure, whereas a rise in sheath temperature reduces the sheath strength. If the channel coolant pressure falls below the fuel element internal gas pressure, the sheath stress is increased. Increased internal fuel element gas pressure, along with the decreased coolant pressure, increases fuel sheath stresses. If fuel temperature becomes high enough, sheath failure can occur in a large number of fuel bundles, releasing fission products to the coolant. One of the challenges met during the fuel analysis was to set a credible, yet conservative “image” of the in-core fuel power/burnup distribution. Consequently, a statistical analysis was performed to find the best-estimate plus uncertainties map for the power/burnup distribution of all in-core fuel elements. For each power/burnup bin in the map, the fission product inventory and the fuel parameters at the end of the steady-state irradiation stage were computed. Afterwards, for each power/burnup bin in the map, the fuel behavior is simulated during the transient. Based on the fuel failure criteria, the failed fuel elements are identified, providing the total radioactive release to the coolant circuit, base for the final dose assessment. The present paper reviews the methodology and results for a typical Design Basis Safety Analysis – Large LOCA with All Safety System Available. Methodologies used in the analysis and results are presented, focused upon fuel behavior.