Kinetic and Stochastic Modeling and Simulation of Fission Product Activity in Primary Loop of a Typical PWR

Abstract

Muhammad Javed Iqbal, PhD, Department of Physics & Applied Mathematics, PIEAS, May, 2008."Kinetic and Stochastic Modeling and Simulation of Fission Product Activity in Primary Loop of a Typical PWR”; Supervisor: Dr. Nasir M. Mirza; Co-Supervisor: Dr. Sikander M. Mirza; Department of Physics & Applied Mathematics, PIEAS, Nilore 45650, Islamabad. In comparison with emerging power systems, the Pressurized Water Reactors (PWRs) have many times higher post shutdown radiation levels, originating partly from the fission products released to the primary coolant from defective fuel pins in the core. This results in prolonging the maintenance schedule and translates into substantial economic costs. To minimize the plant maintenance time and to reduce the radiation dose for plant operators and general public, a detailed knowledge of radioactivity buildup and its kinetics is essential. In this work, a detailed methodology has been developed for modeling and simulation of kinetics of fission product activity in primary coolant loops of typical PWRs under steady-state and power transients. For this purpose, a detailed three-stage methodology has been developed and implemented in the computer code FPCART, which uses LEOPARD and ODMUG codes as its subroutines. It has been coded in Fortran-77 and uses adaptive Runge-Kutta-Fehlberg algorithm as its base ODE- solver. Mathematical model is based on a coupled system of first order, ordinary differential equations governing the kinetics of dominant fission products within the fuel, fuel-clad gap, and the primary coolant loops. Code is capable of handling power transients, and takes into account the effects of purification system as well. Simulation of fission product activity in primary coolant under flow-rate transients have also been performed by using a two-stage model from fuel to fuel-clad gap and then from gap to primary coolant region. A one-dimensional nodal-scheme has been developed for modeling the behavior of fission products in the primary circuit. For normal constant power operation, results of over 39 fission products show that activity due to fission products in the fuel region of PWRs is dominated by 134 I and is followed by 134 Te and 133 I. The value of the fission product activity in fuel region predicted by FPCART code has been found to agree with-in 0.36% range with the corresponding values found by using the ORIGEN-2.0 code. The predictions of FPCART code for primary coolant activity have been found in good agreement with corresponding values of ANS-18.1 Standard as well as with some power plant measured data with 2.4% deviation in the value of specific activity of the dominating fission product 134 I. Similarly, xviifor constant power operation and constant flow rate, results for 15 major fission products show that the activity in the primary coolant circuit of PWRs is dominated by 133 Xe and it is followed by 135 Xe, 131M Xe and 129 Te which contribute 40%, 12.9%, 11% and 8.2%, respectively, to the total fission product activity. These simulations indicate a strong dependence of saturation values of specific activity on primary coolant flow rate. For pump coast-down having a characteristic time t p ~ 2000 h, an 8.6% increase has been observed in the value of total specific activity due to fission products. For increasing t p values, the value of maximum specific activity due to fission products shows a rise followed by an approach towards a saturation value. The simulation of primary coolant activity due to 85 Kr, 87 Kr and 135 Xe chains, have been carried out using classic Runge-Kutta (RK4), adaptive Runge-Kutta-Fehlberg (RKF), Adams-Bashforth-Moulton (ABM) and Semi-Implicit-Extrapolation (SIA), with later two as stiff solvers. Deviations were observed between the corresponding predictions between the lumped and un-lumped systems, especially, during the initial phase of the simulations. Finally, a stochastic model has been developed for simulation of fuel failure time sequences by sampling time dependant intensity functions. Then the three stage model based deterministic methodology of FPCART code has been extended to FPCART-ST, which simulate the random fuel failure sequences followed by burst release of radioactive contents present in fuel-clad gap at that time, into primary coolant coupled with power transients. The value of the 131 I activity in primary coolant predicted by FPCART-ST code has been found in good agreement with the corresponding values measured at ANGRA-1 nuclear power plant. The predictions of FPCART-ST code with constant release option have also been found in good agreement with corresponding experimental values for time dependant 135 I, 135 Xe and measured during EDITHMOX-1 experiments.

Kr concentrations in primary coolant