• Nuclear Engineering and Technology
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• 제54권7호
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• pp.2635-2649
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• 2022

CEFR is a small core-size sodium-cooled fast reactor (SFR) using high enrichment fuel with stainless-steel reflectors, which brings a significant challenge to the deterministic methodologies due to the strong spectral effect. The neutronic simulation of the start-up experiments conducted at the CEFR have been performed with a deterministic code system RAST-F, which is based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo (MC) code and a multi-group nodal diffusion solver. The RAST-F results were compared against the measurement data. Moreover, the characteristic of neutron spectrum in the fuel rings, and adjacent reflectors was evaluated using different models for generation of accurate nuclear libraries. The numerical solution of RAST-F system was verified against the full core MC solution MCS at all control rods fully inserted and withdrawn states. A good agreement between RAST-F and MCS solutions was observed with less than 120 pcm discrepancies and 1.2% root-mean-square error in terms of k_eff and power distribution, respectively. Meanwhile, the RAST-F result agreed well with the experimental values within two-sigma of experimental uncertainty. The good agreement of these results indicating that RAST-F can be used to neutronic steady-state simulations for small core-size SFR, which was challenged to deterministic code system.

• Nuclear Engineering and Technology
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• 제55권12호
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• pp.4491-4503
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• 2023

This study aimed to verify and validate the transient simulation capability of the hybrid code system RAST-F for fast reactor analysis. For this purpose, control rod (CR) drop experiments involving eight separate CRs and six CR groups in the China Experimental Fast Reactor (CEFR) start-up tests were utilized to simulate the CR drop transient. The RAST-F numerical solution, including the neutron population, time-dependent reactivity, and CR worth, was compared against the measurement values obtained from two out-of-core detectors. Moreover, the time-dependent reactivity and CR worth from RAST-F were verified against the results obtained by the Monte Carlo code Serpent using continuous energy nuclear data. A code-to-code comparison between Serpent and RAST-F showed good agreement in terms of time-dependent reactivity and CR worth. The discrepancy was less than 160 pcm for reactivity and less than 110 pcm for CR worth. RAST-F solution was almost identical to the measurement data in terms of neutron population and reactivity. All the calculated CR worth results agreed with experimental results within two standard deviations of experimental uncertainty for all CRs and CR groups. This work demonstrates that the RAST-F code system can be a potential tool for analyzing time-dependent phenomena in fast reactors.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4048-4056
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• 2023

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1789-1803
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• 2022

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group cross-section calculation schemes are employed to improve the agreement between the nodal and reference solutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated by collision probability code TULIP. A good agreement between MCS/RAST-F and reference solution is observed with less than 120 pcm discrepancy in k_eff and less than 1.2% root-mean-square error in power distribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the three-dimensional simulation of reactor cores with fast spectrum.