Farzad Rahnema

Abstract:

COMET is based on an incident flux response expansion method in which nuclear reactor core heterogeneities are explicitly modeled without any spatial homogenization. The method/code performs steady state criticality analysis in water reactors (BWR, PWR and CANDU) as well as prismatic reactor cores typical of fast and thermal gas and liquid metal cores. The method consists of local transport calculations of response functions, a two-level sweeping technique to determine the global eigenvalue and construction of the global fuel pin-power distribution. A stochastic neutron transport method pre-calculates local response expansion coefficients for each unique assembly/block. The use of the stochastic method for computation of the local response functions allows the heterogeneity and geometric complexity of fuel assemblies typical of any reactor to be fully modeled without using homogenized cross sections. Extensive benchmarking work in prismatic fast and thermal reactors as well as operating water reactors (BWR, CANDU and PWR) has demonstrated that COMET has similar accuracy and geometric flexibility as the stochastic method while having computational efficiency that is significantly (2-3 orders of magnitude) faster than conventional fine mesh deterministic transport and whole core stochastic methods. For example, the COMET method has been shown to accurately perform steady state (criticality) analysis in stylized 3-dimensional water reactor cores typical of BWR, PWR, and CANDU. In these cases, COMET determined the eigenvalue of each core with an error on the order of less than 100 pcm from the benchmark Monte Carlo solution, and the normalized pin fission density with an average error of less than 1%. These accurate results have been achieved regardless of the reactor core type and geometry. In this work, we will describe the latest advancements in the COMET method and demonstrate its computational efficiency and accuracy by comparing its results to those of a direct stochastic reference method for calculating the fuel temperature reactivity coefficient and control rod worth in a new integral inherently safe light water (I2S-LWR) reactor concept. The demonstration will highlight the capability of COMET to accurately calculate reactivity coefficient and control rod worth in addition to providing a detailed solution in the whole reactor core, including individual fuel pin power distribution, in a short period of time without requiring supercomputing resources as is the case with any other neutron transport method/code.