Utilization of high-density fuel and beryllium elements for the neutron flux enhancement in typical MTR type research reactors

Abstract

Pakistan Research Reactor-1 (PARR-1) is a typical MTR type swimming pool reactor utilizing low enriched uranium (LEU), i.e. 19.99% enriched in 235U, silicide dispersion fuel of density 3.28 gU/cm 3. This simulation study was conducted by employing the standard reactor physics simulation codes WIMS-D/4, CITATION, and a burnup analysis code FCAP along with a reactor thermal hydraulic simulation code PARET. The present study shows that by directly substituting LEU silicide dispersion fuel of density 4.8 gU/cm 3 in place of fuel currently in use in PARR-1 and by loading beryllium elements at the unused 9 × 6 position of PARR-1 grid plate, a smaller equilibrium core can be designed that can provide 82% higher neutron fluxes at the central flux trap position at 14% lower cost than the existing core. Fuel cycle length of this core is also two days larger than the existing core and this core can be operated safely at the existing power of 10 MW with the existing coolant flow rate of 1000 m 3/h. A possible use of LEU U–Mo monolithic fuel of density 15.3 gU/cm 3 with some adjustment in fuel to moderator ratio and use of Be reflector would provide 89% higher neutron flux in PARR-1 at 29% lower cost. Fuel cycle length of this core will be five days shorter than the existing core and it will require 48% more coolant flow rate for its safe operation at 10 MW.