ICONE17-75868 STUDY ON THE COUPLED NEUTRONIC AND THERMAL-HYDRAULIC CHARACTERISTICS OF THE NEW CONCEPT MOLTEN SALT REACTOR

ABSTRACT The new concept Molten Salt Reactor is the only liquidfuel reactor of the six Generation IV advanced nuclear energy systems. The liquid molten salt serves as the fuel and coolant simultaneously and causes one important feature: the delayed neutrons precursors are drifted by the fuel flow, which leads the spread of delayed neutrons distribution to non-core parts of the primary circuit, and it also can result in a reactivity variation depending on the flow condition of the fuel salt. Therefore, the neutronic and thermal-hydraulic characteristics of the Molten Salt Reactor is quite different from the conventional nuclear reactors using solid fissile materials, and no other reactor design theory and safety analysis methodologies can be used for reference. The neutronic model is derived based on the conservation of particle considering the flow effect of the fuel salt in the Molten Salt Reactor, while the thermal-hydraulic model uses the fundamental conservation laws: the mass, momentum and energy conservation equations. Then the neutronic and thermalhydraulic calculations were coupled and the influences of inflow temperature and flow velocity on the reactor physical properties were obtained. The calculated results show that the flow effect on the distributions of thermal and fast neutron fluxes is very weak, as well as on the effective multiplication factor k eff . While the flow effect on the distribution of delayed neutron precursors is much stronger. The inflow temperature influences the distribution of neutron flux and delayed neutron precursors slightly, and makes significant negative reactivity. Coupled calculation also reveals that the flow velocity of molten salt has little effect on the distribution of neutron fluxes in the steady state, but affects the delayed neutron precursors' distribution significantly

Thermal Hydraulic Analysis of a Passive Residual Heat Removal System for an Integral Pressurized Water Reactor

A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system

STEADY-STATE THERMAL-HYDRAULIC ANALYSIS OF A PLATE TYPE REACTOR

ABSTRACT In the plate type fuel reactor, all the fuel elements are arranged in parallel in the subassembly and set in the core, which create many isolated parallel rectangular channels. The arrangements of the fuel elements may have significant impact on the thermal-hydraulics of the core. In order to fully understand the geometry impact on the steady-state thermalhydraulic characteristics, based on the mass, momentum and energy conservation equations, a multi-channel code was developed to estimate the steady-state thermal-hydraulic characteristics. Two different types of fuel elements arrangements in the subassembly were proposed and studied, one was uniform distribution with the same gap sizes between each fuel elements, and the other was nonuniform distribution with different clearances among boundary fuel elements. The mass flux distribution, the temperature field and the DNBR were obtained by the developed code. It showed that the mass flux distribution was affected by the flow area of the channel and the given power distribution. The flow area of the channel has bigger influence than the second. In addition, the results of the two different types of fuel elements arrangements were compared, and it could be found that the asymmetrical cooling of the fuel elements occurred when fuel elements were nonuniformly arranged, which leaded to the decrease of the maximum temperature in the asymmetrically cooled fuel elements and the increase of the maximum temperature in the core. Furthermore, the DNBR calculated by the Sudo model indicated that the safety margin of the reactor under the present conditions is sufficient. All the obtained results served some valuable information for the design of the new type research reactor

ICONE18-29172 INVESTIGATION OF PRESSURE DROP FOR FLUID FLOW THROUGH POROUS MEDIA: APPLICATION TO A PEBBLE-BED REACTOR

ABSTRACT The present studied Pebble-Bed Reactor is a light-water cooled reactor that consists of millions of Micro-Fuel Elements, and the TRISO-coated fuel particles(MFE) fill the fuel assembly disorderly and form a porous media with internal heat source. Papers on porous media continue to be published at the rate of about 150 per year and the domain of application is wide spread, ranging from chemical particle beds, mass separator units, debris beds, soil investigations, heat pipes and fluidized beds etc. In this paper, investigation is performed on the press drop under conditions of both single-phase and two-phase flow through porous media. Large number of relations are studied and the relational expressions, which generalize the available data of experiments, are suggested for pressure drop calculation in a pebble bed of spheres at random distribution. Finally, the relational expressions are applied to analyze the flow characteristics of the Pebble-Bed Reactor, such as the influence of pressure on two phase friction factor in the core etc

Multiple solvers for implicit temperature calculation of heat conduction with the MPS method

In this study, an implicit algorithm and different solvers are applied to the Moving Particle Semi-Implicit (MPS) method for temperature calculation. The original MPS method uses an explicit method for temperature calculation and is limited by time increment due to diffusion number. In this paper, the heat conduction of plates with Dirichlet boundary condition and Neumann boundary condition is studied. The accuracy of explicit and implicit calculation of plate heat transfer cases are compared. In addition, different solvers are compared. Consequently, it is shown thatthe error of the implicit algorithm is not much different from the original one. And CG is still the better solver,CGS is also a superior solver. The implicit algorithm increases the size of a single time stepby a maximum of 50 times (original diffusion number is 0.2), while the calculation time of a single time step does not increase Substantially, so it has a significant effect on the acceleration of calculation

Development of Safety Analysis Code TACOS and Application to Fuel Qualification Test Loop

In this study, transient analysis code of SCWRs (TACOS), with the ability of simulating transients or accidents under both supercritical water (SCW) conditions and subcritical water conditions, has been developed with FORTRAN 90 language, and simulation has been performed to the European SCWR fuel qualification test (FQT) system. The semi-implicit finite difference technique was adopted for the solution of coolant dynamic behavior in the loop. Furthermore, an illustration of numerical solution for the heat structure model and other models was presented. The code TACOS is then applied to simulate the Edward-O'Brian blow-down experiment to evaluate its capacity in simulating the fast blow-down progress. Therefore, the design basis accidents (DBAs) with the transcritical transient were investigated for the SCWR-FQT system. The results by TACOS indicate that the SCWR-FQT with the existing safety system can be cooled effectively