Engineering and Physics Optimization of Breed and Burn Fast Reactor Systems: Annual and Final Report

Weaver, Kevan D. ; Marshall, Theron ; Parry, James

Idaho National Laboratory

关键词: Heat Transfer;    Fuel Cycle;    Reactor Components;    Fast Reactors;    Thermodynamic Properties;   

DOI  :  10.2172/911594
RP-ID  :  INL/EXT-05-00886
RP-ID  :  DE-AC07-99ID-13727
RP-ID  :  911594

美国|英语

来源: UNT Digital Library

【 摘 要 】

The Idaho National Laboratory (INL) contribution to the Nuclear Energy Research Initiative (NERI) project number 2002-005 was divided into reactor physics, and thermal-hydraulics and plant design. The research targeted credible physics and thermal-hydraulics models for a gas-cooled fast reactor, analyzing various fuel and in-core fuel cycle options to achieve a true breed and burn core, and performing a design basis Loss of Coolant Accident (LOCA) analysis on that design. For the physics analysis, a 1/8 core model was created using different enrichments and simulated equilibrium fuel loadings. The model was used to locate the hot spot of the reactor, and the peak to average energy deposition at that location. The model was also used to create contour plots of the flux and energy deposition over the volume of the reactor. The eigenvalue over time was evaluated using three different fuel configurations with the same core geometry. The breeding capabilities of this configuration were excellent for a 7% U-235 model and good in both a plutonium model and a 14% U-235 model. Changing the fuel composition from the Pu fuel which provided about 78% U-238 for breeding to the 14% U-235 fuel with about 86% U-238 slowed the rate of decrease in the eigenvalue a noticeable amount. Switching to the 7% U-235 fuel with about 93% U-238 showed an increase in the eigenvalue over time. For the thermal-hydraulic analysis, the reactor design used was the one forwarded by the MIT team. This reactor design uses helium coolant, a Brayton cycle, and has a thermal power of 600 MW. The core design parameters were supplied by MIT; however, the other key reactor components that were necessary for a plausible simulation of a LOCA were not defined. The thermal-hydraulic and plant design research concentrated on determining reasonable values for those undefined components. The LOCA simulation was intended to provide insights on the influence of the Reactor Cavity Cooling System (RCCS), the containment building, and a Decay Heat Removal System (DHRS) on the natural circulation heat transfer of the core's decay heat. A baseline case for natural circulation had to be established in order to truly understand the impact of the added safety systems. This baseline case did not include a DHRS, although the current MIT design does have a DHRS that features the highly efficient Printed Circuit Heat Exchangers (PCHEs). The initial LOCA analysis revealed that the RCCS was insufficient to maintain the reactor core below the fuel matrix decomposition temperature. A guard containment was added to the model in order to maintain a prescribed backpressure during the LOCA to enhance the natural circulation. The backpressure approach did provide satisfactory natural convection during the LOCA. The necessary backpressure was 1.8 MPa, which was not especially different from the values reported by other gas fast reactor researchers. However, as the model evolved to be more physically representative of a nuclear reactor, i.e., it included radial peaking factors, inlet plenum orificing, and the degradation of SiC thermal properties as a result of irradiation, the LOCA-induced fuel temperatures were not consistently below the decomposition limit.