Burnup calculation

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

Application of Gd2O3 in form of burnable poison particles (BPPs) has been investigated for improving neutronics performance of VVER-1000 fuel assembly. Numerical calculations have been conducted for optimizing BPP parameters using the MVP code and the JENDL-3.3 data library.

7p visnape 30-01-2023 9 4   Download

In this study, analysis of the complete operational history of the “Joˇzef Stefan” Institute (JSI) TRIGA reactor was performed. Reactor power changes, core configurations and weekly excess reactivity measurements were analysed to obtain the needed data for fuel burnup calculations.

21p vironald 15-12-2022 13 4   Download

Validating boiling water reactor (BWR) spent nuclear fuel inventory calculations is challenging due to the complexity of BWR assembly designs, the lack of publicly available radiochemical assay measurements, and limited access to documentation on fuel design and operating conditions.

15p vironald 15-12-2022 9 3   Download

"The journal of Nuclear science and technology - Volume 8/Number 4, 2018" present reactivity induced transient analysis when the occurrence of leakage in the dry irradiation channels of the Dalat Nuclear Research Reactor; burnup calculation of the OECD VVER-1000 LEU benchmark assembly using MCNP6 and SRAC2006; particle identification for neutron rich nuclei 63,65Cr from knockout reactions...

48p trinhthamhodang1218 14-03-2021 13 1   Download

The present work aims to perform burnup calculation of the OECD VVER-1000 LEU (low enriched uranium) computational benchmark assembly using the Monte Carlo code MCNP6 and the deterministic code SRAC2006.

10p trinhthamhodang1218 14-03-2021 4 1   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

"The journal of Nuclear science and technology - Volume 9/Number 2, 2019" present on burnup modelling issues associated with vver–440 fuels; processing of the multigroup cross-sections for MCNP calculations; low-energy experiments at the s3 spectrometer; dosimetric characteristics of 6 MV photons from truebeam STX medical linear accelerator simulation and experimental data...

56p trinhthamhodang1218 14-03-2021 22 2   Download

The paper investigates various computational modelling issues associated with VVER-440 fuel depletion, relevant to burnup credit. The SCALE system and the TRITON sequence are used for the calculations. The effects of variations in depletion parameters and used calculation methods on the isotopic vectors are investigated. The burnup behaviour of Gadolinium is quite important in actual core analysis, but its behaviour is somewhat complicated, requiring special treatment in numerical modelling and calculations.

9p trinhthamhodang1218 14-03-2021 9 1   Download

A comparison between stochastic and deterministic depletion calculations based on a graphite-filled MOX fuel assembly configuration is presented in this paper. The infinite multiplication factors and isotope inventory changes as a function of burnup obtained by Monte Carlo method module SCALE/KENO and deterministic method module SCALE/NEWT are compared with those obtained by deterministic code HELIOS. The impact in calculation results by using different nuclear data library is also investigated.

10p trinhthamhodang9 10-12-2020 10 0   Download

In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1) and 233U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode.

6p minhxaminhyeu5 30-06-2019 32 0   Download

This reutilization option is a potential candidate technique to make better use of the nuclear resources. Standard two step method is used to calculate node i.e. fuel assembly average burnup and then pin by pin h values are reconstructed to ascertain the residual reactivity in the used fuel pins. Fuel pins with h >1:0 are used to reconstruct to-be-reused fuel assemblies.

11p minhxaminhyeu3 12-06-2019 10 0   Download

The burnup equation of nuclides is one of the most equations in nuclear reactor physics, which is generally coupled with transport calculations. The burnup equation describes the variation of the nuclides with time. Because of its very stiffness and the need for large time step, this equation is solved by special methods, for example transmutation trajectory analysis (TTA) or the matrix exponential methods where the matrix exponential is approximated by CRAM.

5p minhxaminhyeu3 12-06-2019 7 0   Download

The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS.

10p minhxaminhyeu3 25-06-2019 19 0   Download

The aim of this book is to disseminate state-of-the-art research and advances in the area of nuclear reactors technology. The book was divided in two parts.

0p phoebe75 19-02-2013 65 8   Download