Sub-channel analysis of Pb-Bi cooled fast breeder reactor PEACER fuel assembly using MATRA

Full text of publication follows: The nuclear power is the one of the realistic means that can solve the shortage of usable energies due to depletion of fossile fuel and due to the environmental contamination. However, since the nuclear fission yields a kind of fission fragment as a by-product, a radiological hazard of spent fuel is now a major problem. To overcome this difficulty, a number of studies are being performed and planned. One of the key solutions to this problem is to eliminate spent fuel by nuclear transmutation. According to this research, the most significant long-lived isotopes in spent fuels of the current power reactors can be transmuted into short-lived ones by using fast neutron spectrum with localized thermal traps. For this reason, liquid metal cooled fast breeder reactor is widely chosen as the answer to solve the problem. In Korea, PEACER(Proliferation-resistant, Environment-friendly, Accident-tolerant, Continual and Economical Reactor) is under study to work out this issue. PEACER core is designed to produce about 1560 MW of the thermal output with electric output up to 550 MW which efficiency is about 0.35. PEACER uses control rode made of B4C to perform reactivity control and Pb-Bi liquid metal is adopted as a coolant for primary system. In nuclear power plants it is important to keep the temperature of the reactor core structures under certain criteria in order to prevent damage of fuel materials which can advance to severe situations such as radiation leakage, and even meltdown of the fuel. This study was intended to see the liquid metal coolant behavior along the PEACER fuel channels and to find out whether the given heat flux profiles and geometrical arrangement of the fuel rods yields reasonable fluid dynamic distribution under nominal operation by using subchannel approach. The subchannel analysis of the fuel assembly under nominal operation condition was performed using MATRA (Multi-channel Analyzer for Transient and steady-state in Rod Arrays). The MATRA is a thermal hydraulic analysis code, which is developed at KAERI (Korea Atomic Energy Research Institute) based on the subchannel approach for calculating the enthalpy and flow distribution in nuclear fuel rod bundle elements for both steady-state and transient conditions. A small computer program based on basic heat transfer concepts was developed to predict fuel material and coolant temperature distribution of the reactor along axial and radial direction and the results were compared with that of MATRA. The result showed us that the input data based on the current design of PEACER core yielded reliable results, under limits which the deterioration of fuel structure may happen. Typical results obtained include hydrodynamic properties of Pb-Bi liquid metal coolant in core subchannels and the thermodynamic properties of core structures. The results obtained by each program showed little difference and it was mainly due to different approaches of which they have adopted. (authors)