Estimation of neutron source uncertainties in pressure vessel fluence calculations

One of the major issues facing the aging nuclear power plants is radiation embrittlement of the plant structures, especially the weldings of the pressure vessel (PV). Embrittlement of the PV is estimated based on the PV fluence, which has to remain below a specified limit. To determine the fluence, multidimensional, multigroup transport calculations are performed. The results are benchmarked based on the in-vessel capsule (front of PV) or, more recently, the ex-vessel cavity dosimeter (back of PV). Since the flux/fluence shape and spectrum within the PV is only obtained via calculations, whereas experiments are done outside the PV, it is necessary to improve the calculational accuracy and to establish the range of uncertainties. In recent years, the LEPRICON package was developed to correct for known uncertainties, thereby improving the confidence level of calculational results. LEPRICON, however, may not be the complete answer because core designers are changing fuel management strategies to improve neutron economy, e.g., through longer cycle cores with low neutron leakage. Higher burnups of peripheral assemblies lead to higher concentrations and contributions of plutonium isotopes to fission source. So, certain assumptions, issues, uncertainties, and previously developed formulations related to neutron source calculations have to be revisited, which is the subject of this paper. The multigroup neutron source distribution within the core is determined as Sig = xigCiP, where x is the source spectrum, the C factor is the power-to-source conversion factor (C = v/ER; ER is recoverable energy per fission), P is the cycle-averaged power distribution throughout the reactor core, and i and g refer to spatial mesh and energy group, respectively. The pinwise power distribution is determined by performing CASMO/SIMULATE criticality calculations