Neutronics/thermal-hydraulics coupling calculations and new coupling cases studies for the PWR MOX/UO2 benchmark

다국어 초록 (Multilingual Abstract)

To better support research on neutronics/thermal-hydraulics (N-TH) coupling numerical methods and associated software development for pressurized water reactor (PWR) cores, a coupled calculation and cases extension study was conducted for the PWR MOX/UO2 transient rod ejection benchmark. The newly developed pin-by-pin SP3 neutron transport code CORCA-SPn by the Nuclear Power Institute of China (NPIC) and the open-source advanced nodal code KOMODO were employed to perform coupled calculation modeling and validation for the Part II to Part IV cases of the PWR MOX/UO2 benchmark. Additionally, eight new steady-state and eight transient neutronics-thermal hydraulics coupled conditions were designed, and all 16 new cases were analyzed using CORCA-SPn and KOMODO. The following conclusions were drawn: 1) For all 19 coupled cases, good agreement was observed between CORCA-SPn and KOMODO results, indicating comparable accuracy. 2) Compared with internationally published results for the three existing conditions (Part II–IV), deviations from CORCA-SPn and KOMODO fell within acceptable ranges. Compared to PARCS result based on 2-group crosssection libraries, the transient core relative power peak deviations were 20.1 % and 14.1 %, respectively; compared to Serpent/SCF, these values were 􀀀 4.75 % and 􀀀 9.50 %. 3) For the eight new steady-state cases, the maximum deviations between KOMODO and CORCA-SPn were 􀀀 2.7 pcm (critical boron concentration), 0.439 % (assembly-wise relative power), 􀀀 1.88 ◦C (average fuel Doppler temperature), and 􀀀 0.10 ◦C (average coolant temperature). 4) For the eight new transient cases, the maximum deviations were 􀀀 3.77 % (transient core relative power peak), 􀀀 0.004$ (reactivity peak), 􀀀 2.24 ◦C (average fuel Doppler temperature at 1.0 s), and 􀀀 0.09 ◦C (average coolant temperature at t = 1.0s).