Etude de surete du SCWR par prise en compte du couplage neutronique-thermohydraulique.

摘要

The Generation IV Forum is an international initiative that aims at designing improved nuclear reactors. These reactors should come into production in the next 40 years. Generation IV reactors will produce less radioactive waste, will have enhanced safety features, will resist nuclear proliferation and will be less expensive than current nuclear reactors.;Although safety is a key issue in Generation IV reactors, very few SCWR safety studies have been conducted so far. This project aims at addressing this issue by proposing a safety study that takes into account the coupling between neutronic and thermalhydraulic phenomena. Our objective is to analyze how the reactivity of the SCWR is affected by the total power of the reactor and by thermalhydraulic parameters such as the coolant's mass flow rate, the coolant's entry temperature and the coolant's pressure.;This analysis was done in comparison with CANDU-6 reactors. The neutronic-thermalhydraulic coupling was put into place using the ARTHUR code, developed at Ecole Polytechnique de Montreal. ARTHUR allows data exchange between internal thermalhydraulic functions programmed in FORTRAN 90 and external neutronic codes. This tool was designed to simulate CANDU-6 reactors and therefore had to be adapted to take into account the presence of a supercritical coolant. The iterative solution procedure of the thermalhydraulic equations was also simplified to decrease computational time. Several approximations made in the original version of ARTHUR were also abandoned.;The neutronic part of the calculations were performed using a combination of the DRAGON and DONJON codes, both developed at the Ecole Polytechnique de Montreal. The DRAGON code uses transport theory to perform lattice calculations and the DONJON code uses diffusion theory for finite reactor calculations.;The SCWR (SuperCritical Water Reactor) is the main Generation IV concept being studied in Canada. The Canadian version of the SCWR follows the pressure tube reactor concept, much like today's CANDU-6. However, the SCWR uses light water as coolant and slightly enriched uranium as fuel (4.25%). The pressure tubes are contained in heavy water, which acts as moderator. Maintaining the coolant at supercritical pressures will allow the reactor to reach higher thermal efficiencies than the CANDU-6 (45% compared with 30%-35%).;The results of this comparative study showed that the intrinsic safety characteristics of the SCWR are superior to those of the CANDU-6. Much like the CANDU-6, the SCWR is characterized by a negative power coefficient. Every increase in power therefore results in a decrease in reactivity. However, the advantage of the SCWR lies in its positive mass flow rate coefficient. In the SCWR, any decrease of the mass flow rate leads to a reactivity decrease, which in turn prevents the fuel from overheating. As a result, a loss of coolant is much less damageable to the SCWR than it is to the CANDU-6.;Simulations of a unit reactor cell using transport theory showed that SCWR's positive mass flow rate coefficient is mainly caused by the SCWR's small lattice pitch. A smaller lattice pitch decreases the destabilizing effect of a loss of coolant density. The stabilizing effect caused by the increase in fuel temperature becomes dominant.;Finally, this study showed that the behaviour of the SCWR does not change during the reactor's life cycle. This feature is an improvement, as the CANDU-6's power coefficient gradually increases in time before taking positive values.