Abstract
Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized in nodal diffusion codes, PARCS and RAST-K. The applicability of the methodology is quantified on the sodium fast reactor (SFR) ABR-1000 design with a metallic fuel from the OECD/NEA SRF benchmark. The few-group XSs generated by MCS with a two-dimensional (2D) fuel assembly geometry are well consistent with those of SERPENT 2. Furthermore, the simulation of beginning-of-cycle (BOC) steady-state three-dimensional (3D) whole-core problem with PARCS and RAST-K is conducted using the generated 24-group XSs by MCS. The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS. Overall, the code-to-code comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-mean-square (RMS) error in assembly power less than 1.15%. Therefore, it is successfully demonstrated that the employment of the MG XSs generation by MCS for nodal diffusion codes is feasible to accurately perform analyses for fast reactors.
Highlights
A Monte Carlo (MC) code or a deterministic code is to be employed to simulate the nuclear reactor
The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS
These XSs data is converted into the compatible database that is able to be used in nodal diffusion codes PARCS and RAST-K
Summary
A Monte Carlo (MC) code or a deterministic code is to be employed to simulate the nuclear reactor. Deterministic code’s capability is to provide an adequately accurate result, which requires less computational demand Their disadvantage is the simplicity in geometry and transport/diffusion physics. The combination of both stochastic and deterministic codes has become more and more attractive in the framework of establishing the basic viability of the advanced fast reactor. This approach ensures the superposition of the benefits of these methods. The generation of the multigroup (MG) cross sections (XSs) data by MC method is implemented in the in-house UNIST MC code MCS. The simulation of beginning-of-cycle (BOC) steady-state threedimensional (3D) whole-core problem, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS
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