Abstract
This paper examines the accuracy and performance of the RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system for the simulation of a used nuclear fuel (UNF) cask. RAPID is capable of determining eigenvalue, subcritical multiplication, and pin-wise, axially-dependent fission density throughout a UNF cask. We study the source convergence based on the analysis of the different parameters used in an eigenvalue calculation in the MCNP Monte Carlo code. For this study, we consider a single assembly surrounded by absorbing plates with reflective boundary conditions. Based on the best combination of eigenvalue parameters, a reference MCNP solution for the single assembly is obtained. RAPID results are in excellent agreement with the reference MCNP solutions, while requiring significantly less computation time (i.e., minutes vs. days). A similar set of eigenvalue parameters is used to obtain a reference MCNP solution for the whole UNF cask. Because of time limitation, the MCNP results near the cask boundaries have significant uncertainties. Except for these, the RAPID results are in excellent agreement with the MCNP predictions, and its computation time is significantly lower, 35 second on 1 core versus 9.5 days on 16 cores.
Highlights
To ensure criticality safety and maintain a high level of security of Used Nuclear Fuel (UNF), it is necessary to perform detailed Particle Transport simulations
The RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system [4] is based on the Multi-stage Response-function Transport (MRT) methodology [5], and uses the fission matrix (FM) formulation to determine eigenvalue, subcritical multiplication and fission density distribution of nuclear systems
This section is divided into four main parts: in Section 5.1 the parametric study of the single-assembly problem to assess adequate eigenvalue calculation parameters to establish a referene MCNP solution is presented; Section 5.2 addresses the cycle-to-cycle correlation issue; while in Sections 5.3 and 5.4 the results of the RAPID code are compared to the MCNP reference calculation for the single assembly and full cask models, respectively
Summary
To ensure criticality safety and maintain a high level of security of Used Nuclear Fuel (UNF), it is necessary to perform detailed Particle Transport simulations. The RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system [4] is based on the Multi-stage Response-function Transport (MRT) methodology [5], and uses the FM formulation to determine eigenvalue, subcritical multiplication and fission density distribution of nuclear systems. We performed a detailed sensitivity analysis on a single-assembly model in order to determine the MCNP eigenvalue parameters necessary to obtain a reliable reference solution. These results provided a guide for the selection of an appropriate set of parameters for the eigenvalue calculation for the whole cask
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