Pressure-tube Supercritical Water-cooled Reactor Research Articles

Determination of the optimal few-energy group structure for the Canadian Super Critical Water-cooled Reactor

Most deterministic neutron transport codes implement the multi-group approximation theory to solve the linear Boltzmann equation by which the neutron flux and the interaction cross-sections are averaged over discretized energy groups. In order to further reduce the computational cost for large scale and/or time dependent problems, this multi-group information undergoes additional energy group condensation into the so-called few energy group structure which may be used in a broad range of diffusion based full-core analyses. The accuracy and efficiency of few-group based computations is dependent on the number and the boundaries of the discrete energy group structure. Since the flux spectrum used in the homogenization process may not be known a priori and indeed may evolve in space, with burnup, and during transients, the optimal energy group structure depends on reactor type, design, operating conditions, fuel type, and composition. The Canadian Pressure Tube Super-Critical Water-cooled Reactor (PT-SCWR) is a Generation IV advanced reactor system that uses light water above its thermodynamic critical point as coolant and a plutonium-driven thorium fuel mixture. Considering that the anticipated flux spectrum for such a design deviates significantly from the thermal-neutron dominated CANDU designs, there may be a need for improvements in the number and boundaries of the few-group nuclear data. This paper presents a systematic methodology for the delineation of few-group energy structure for the Canadian PT-SCWR. The methodology used the SCALE (Standardized Computer Analysis for Licensing Evaluation) code package to examine the effect of energy group homogenization and determine the structure which minimizes the sensitivity of the results to energy group partitioning. As such, it utilizes normal and off-normal operating conditions to determine the effect of energy homogenization and determines the minimal group of energy boundaries which can accurately capture the lattice physics phenomena within the lattice cell over a wide range of operating and accident conditions.

Investigation of Fuel Burnup Impacts on Nuclear Reactor Safety Parameters in the Canadian Pressure Tube Supercritical Water-Cool Reactor

The Canadian pressure-tube super critical water-cooled reactor (PT-SCWR) is an advanced generation IV reactor concept which is considered as an evolution of the conventional Canada Deuterium Uranium (CANDU) reactor that includes both pressure tubes and a low temperature and pressure heavy water moderator. The Canadian PT-SCWR fuel assembly utilizes a plutonium and thorium fuel mixture with supercritical light water coolant flowing through the high-efficiency re-entrance channel (HERC). In this work, the impact of fuel depletion on the evolution of lattice physics phenomena was investigated starting from fresh fuel to burnup conditions (25 MW d kg−1 [HM]) through sensitivity and uncertainty analyses using the lattice physics modules in standardized computer analysis for licensing evaluation (SCALE). Given the evolution of key phenomena such as void reactivity in traditional CANDU reactors with burnup, this study focuses on the impact of fission products, 233U breeding, and minor actinides on fuel performance. The work shows that the most significant change in fuel properties with burnup is the depletion of fission isotopes of Pu and the buildup of high-neutron cross section fission products, resulting in a decrease in cell k∞ with burnup as expected. Other impacts such as the presence of protactinium and uranium-233 are also discussed. When the feedback coefficients are assessed in terms of reactivity, there is considerable variation as a function of fuel depletion; however, when assessed as Δk (without normalization to the reference reactivity which changes with burnup), the net changes are almost invariant with depletion.

Practical environment-corrected discontinuity factors and homogenized parameters for improved PT-SCWR neutron diffusion solutions

A novel application of practical discontinuity factors in coarse-mesh finite-difference solutions of heterogeneous multi cells with Pressure Tube Supercritical Water-cooled Reactor (PT-SCWR) type fuel and moderator cells has been investigated. In addition to discontinuity factors, homogenization schemes have also been studied and applied to reflector-adjacent-fuel cells and reflector cells. A 49 node inner-core multi cell model containing only fuel and a 40 node outer-core multi cell model containing reflector and fuel cells were simulated. In addition to nominal conditions, various discontinuity factors and homogenization techniques, along with conventional techniques, were applied to static beyond-nominal condition scenarios to evaluate the error associated with neutron power changes from nominal conditions, relative to reference transport solutions. The use of novel mean reference discontinuity factors and environment homogenized reflector-adjacent-fuel cells, over conventional methods, reduces core-wide reactivity errors, RMS neutron power errors, and maximum channel specific power errors by up to 2.6mk, 2.9%, and 6.7%, respectively.

Axial Power and Coolant-Temperature Profiles for a Non-Re-Entrant Pressure-Tube Supercritical Water-Cooled Reactor Fuel Channel

The axial power and coolant-temperature distributions in a fuel channel of the Generation IV pressure-tube super-critical water-cooled reactor (PT-SCWR) are found using coupled neutronics-thermal-hydraulics calculations. The simulations are performed for a channel loaded with a fresh, 78-element Th-Pu fuel assembly. Neutronics calculations are performed using the DONJON diffusion code using two-group homogenized cross sections produced using the lattice code DRAGON. The axial coolant temperature profile corresponding to a certain axial linear heat generation rate is found using a code developed in-house at University of Ontario Institute of Technology (UOIT). The effect of coolant density, coolant temperature, and fuel temperature variation along the channel is accounted for by generating macroscopic cross sections at several axial positions. Fixed-point iterations are performed between neutronics and thermal-hydraulics calculations. Neutronics calculations include the generation of two-group macroscopic cross sections at several axial positions, taking into account local parameters such as coolant temperature and density and average fuel temperature. The coolant flow rate is adjusted so that the outlet temperature of the coolant corresponds to the SCWR technical specifications. The converged axial power distribution is found to be asymmetric, resembling a cosine shape skewed toward the inlet (reactor top).

Temperature Distribution Inside Fresh-Fuel Pins of Pressure-Tube Supercritical-Water-Cooled Reactor

The radial temperature profile for ThO2-PuO2 fresh-fuel pins of the pressure-tube supercritical-water-cooled reactor (PT-SCWR) is simulated for two different radial distributions of the ThO2 and PuO2, one whereby the ThO2 and PuO2 are mixed together in a single, homogeneous region, and another one whereby ThO2 and PuO2 occupy separate radial regions inside the pin with the ThO2 occupying the inner region. The radial power density in each fuel pin is calculated from a lattice-level transport calculation using the lattice code DRAGON and is used together with the temperature-dependent fuel thermal conductivity to analytically solve the heat conduction equation and thus calculate the temperature profile inside the fuel pin. Results indicate that, for fresh fuel, the centerline temperature is much lower for a two-region pin than for a single-region pin, and thus, the risk of fuel melting is greatly reduced for the two-region fuel compared to the single-region fuel.

Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO2) and fissile (PuO2) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650K for fresh fuel compared to the case of a homogeneous (Th–Pu)O2 mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is determined from the analytical solution of the steady-state heat conduction equation in the fuel pin, ignoring azimuthal dependence and axial heat flow. The temperature dependence of the thermal conductivity is accounted for, as is the radial dependence of the volumetric power density, which is determined from detailed lattice-level transport calculations performed using the lattice code DRAGON. The reduction in the centreline fuel temperature is shown to be caused partly by a reduction in the heat flux in the case of the two-region fuel and partly by the higher thermal conductivity of both pure ThO2 and pure PuO2 compared to that of a ThO2–PuO2 mixture. The centreline temperature of the two-region fuel is shown to be higher for irradiated fuel than for fresh fuel, a fact explained by the depletion of the fissile material in the peripheral region and the buildup of fissile material in the central region.