Neutron kinetics models

This paper presents a model development of the Dalat nuclear research reactor (DNRR) with the Serpent 2 Monte Carlo code. The purpose is to prepare the DNRR Serpent 2 model for performing fuel burnup calculations of the DNRR as well as for generating multi-group neutron cross sections to be further used in the kinetics calculations of the DNRR with a 3D reactor kinetics code.

9p vicedric 08-02-2023 3 3   Download

This paper presents an extension to a point kinetics model of fissile solution undergoing a transient through the development and addition of correlations which describe neutronics and thermal parameters and physical models.

9p vironald 15-12-2022 7 4   Download

This paper presents a model development of the Dalat Nuclear Research Reactor (DNRR) loaded with low enriched uranium (LEU) fuel using the Serpent 2 Monte Carlo code. The purpose is to prepare the DNRR Serpent 2 model for performing fuel burnup calculations of the DNRR as well as for generating multi-group neutron cross sections to be further used in the kinetics calculations of the DNRR with a 3D reactor kinetics code.

9p trinhthamhodang1218 14-03-2021 5 1   Download

These studies are performed in the general framework of transient coupled calculations with accurate neutron kinetics models. This kind of application requires a modeling of the influence on the neutronics of the macroscopic cross-section evolution.

8p christabelhuynh 30-05-2020 3 0   Download

The studies presented in this paper are performed in the general framework of transient coupled calculations with accurate neutron kinetics models able to characterize spatial decoupling in the core. An innovative fission matrix interpolation model has been developed with a correlated sampling technique associated to the Transient Fission Matrix (TFM) approach.

14p christabelhuynh 30-05-2020 7 0   Download

In this work, the reactor kinetics capability is used to compute the design safety parameters in a PWR due to complete loss of coolant flow during protected and unprotected accidents. A thermal-hydraulic code coupled with a point reactor kinetic model are used for these calculations; where kinetics parameters have been developed from the neutronic SRAC code to provide inputs to RELAP5-3D code to calculate parameters related to safety and guarantee that they meet the regulatory requirements. In RELAP5-3D the reactivity feedback is computed by both separable and tabular models.

12p minhxaminhyeu3 25-06-2019 22 0   Download