Development of a PARCS/Serpent model for neutronics analysis of the Dalat nuclear research reactor

Cross-sectional homogenization for full-core calculations of small and complex reactor configurations,such as research reactors,has been recently recognized as an interesting and challenging topic.This paper presents the development of a PARCS/Serpent model for the neutronics analysis of a research reactor type TRIGA Mark-II loaded with Russian VVR-M2 fuel(known as the Dalat Nuclear Research Reactor or DNRR).The full-scale DNRR model and a supercell model for a shim/safety rod and its surrounding fuel bundles with the Monte Carlo code Serpent 2 were proposed to generate homogenized fewgroup cross sections for full-core diffusion calculations with PARCS.The full-scale DNRR model with Serpent 2 was also utilized as a reference to verify the PARCS/Serpent calculations.Comparison of the effective neutron multiplication factors,radial and axial core power distributions,and control rod worths showed a generally good agreement between PARCS and Serpent 2.In addition,the discrepancies between the PARCS and Serpent 2 results are also discussed.Consequently,the results indicate the applicability of the PARCS/Serpent model for further steady state and transient analyses of the DNRR.

Determination of fuel burnup distribution of a research reactor based on measurements at subcritical conditions

This paper presents the determination of the fuel burnup distribution of the Dalat nuclear research reactor(DNRR) using a method of measurements at subcritical conditions. The method is based on the assumption of linear dependence of the reactivity on the burnup of fuel bundles and the measurements at subcritical conditions.The measurements were taken for seven selected fuel bundles in two different measuring sequences. The measured burnup values have also been compared with the calculations for verifying the method and the measurement procedure. The results obtained with the three detectors have a good agreement with each other with a discrepancy less than 1.0%. The errors of the measured burnup values are within 6%. Comparison between the calculated and measured burnup values shows that the discrepancy of the C/E ratio is within 9% compared to unity. The results indicate that the method of measurements at subcritical conditions could be well applied to determine the relative burnup distribution of the DNRR.