Modeling LOCA performance for the generation IV gas-cooled fast reactor design

Full text of publication follows: Generation IV nuclear energy systems are next-generation technologies that will offer significant advances in sustainability, safety and reliability, economics, and proliferation resistance. Expected to be available for worldwide deployment by 2030, these energy systems would provide electrical power for the subsequent decades. The Gas-Cooled Fast Reactor (GFR) is a Generation IV concept that features a fast-neutron spectrum, direct Brayton cycle gas turbine, and a closed fuel cycle. Through the combination of a fast neutron spectrum and the full recycle of actinides, the GFR minimizes the production of long-lived radioactive waste and makes it possible to use existing fissile and fertile materials (including depleted uranium) more efficiently than existing thermal spectrum gas reactors. The prominent GFR design features a 'pancake' style core (H/D ∼ 1.7/2.9 m) that produces 600 MW of thermal power with an average power density of 55 MW/m3. The core is comprised of SiC-coated UPuC spheres that are collected in channels to form a prismatic, hexagonal fuel assembly or coagulated to form fuel pebbles. The 11 m3 core is enveloped by TiN reflectors and stainless steel shields in both the radial and axial directions. The initial GFR design used He gas at a pressure of 7 MPa and an outlet temperature of 850 deg. C, however the design has been expanded to consider supercritical CO2 (S-CO) gas at a pressure of 19 MPa and an outlet temperature of 550 - 650 deg. C. The higher density S-CO has advantageous characteristics during off-normal low flow and pressure conditions. One of the strengths of the Generation IV reactor concepts is their inherent safety and extensive use of passive safety systems. This paper discusses an analysis performed to study the GFR's response during a severe off-normal scenario. The loss of coolant accident was chosen because it will be one of the more severe challenges to the reactors decay heat removal system, which must maintain the peak fuel temperature below 1600 deg. C. The thermal-hydraulics computer code RELAP is used for the accident analyses, which includes both the He and S-CO designs. The accident scenario assumes a double-guillotine break of the concentric inlet and outlet coolant legs, with the coolant consequently blowing down into the containment building. After the initial flow and pressure transient, the core is cooled by natural circulation of the coolant gas as it is thermo-siphoned from the containment building. Analyses of the peak fuel, reactor vessel, and containment atmosphere temperatures are used to quantify the reactors heat rejection performance during the accident scenario and to suggest methods of improving the decay heat removal system, e.g., He primary loop with CO2 injection during the pressure transient to improve heat transfer. (authors)