Results 1 to 10 of about 1,076,657 (296)

Calculation and experimental analysis of benchmark experiments with a fast neutron spectrum and models of sodium and lead cooled fast reactors using different evaluated nuclear data libraries [PDF]

Nuclear Energy and Technology, 2021
The paper presents the results of a comparative analysis of criticality calculations using a Monte-Carlo code with the BNAB-93 and BNAB-RF neutron group constants, as well as with evaluated neutron data files from the Russian ROSFOND evaluated nuclear ...
Olga N. Andrianova, Yury Ye. Golovko, Gleb B. Lomakov, Yevgeniya S. Teplukhina, Gennady M. Zherdev   +4 more
doaj   +3 more sources

Precision neutronic calculations of experiments on the neutron transmission through the reflector layers at the BFS critical facilities for expanding the verification database to justify lead cooled fast reactor designs [PDF]

Nuclear Energy and Technology, 2020
The paper presents the results of the efforts concerned with expanding the verification database and estimating the calculation uncertainty of the power density in the steel reflector of lead cooled fast reactor designs based on experiments performed in ...
Olga N. Andrianova, Evgeniya S. Teplukhina, Gennady M. Zherdev, Zhanna V. Borovskaya, Andrey P. Zhirnov   +4 more
doaj   +3 more sources

One-step coupled calculations (Serpent-OpenFOAM) for a fuel rod of the IPR-R1 triga reactor

Brazilian Journal of Radiation Sciences, 2021
In this work, a single step of coupled calculations for a fuel rod of IPR-R1 TRIGA was performed. The used me-thodology allowed to simulate the fuel pin behavior in steady-state mode for different power levels.
Tiago Augusto Santiago Vieira, Rebeca Cabral Gonçalves, Izabella Cristina de Paiva Machado, Guilherme Augusto Moura Vidal, Higor Fabiano Pereira Castro, Nataly Lamounier Ribeiro, Marcos Pais Barroso Filho, Wilker Gustavo Ferreira Santos, Daniel de Almeida Magalhães Campolina, Graiciany de Paula barros, Vitor Vasconcelos Araújo Silva, André Augusto Campagnole dos Santos   +11 more
doaj   +1 more source

NEUTRONIC MODELLING OF THE FFTF CONTROL ROD WORTH MEASUREMENTS WITH DIFFUSION CODES [PDF]

EPJ Web of Conferences, 2021
This paper presents an assessment of three deterministic core simulators with the focus on the neutronic performance in steady-state calculations of small Sodium cooled Fast Reactor cores.
Nikitin E., Fridman E., Mikityuk K., Radman S., Fiorina C.   +4 more
doaj   +1 more source

Investigation of discretization uncertainty in Monte Carlo neutron transport simulations of the Molten Salt Fast Reactor (MSFR)

Brazilian Journal of Radiation Sciences, 2023
In the present work, an assessment of the Neutronic Benchmark of the Molten Salt Fast Reactor (MSFR) was performed using mesh based Monte Carlo Neutron Transport (MCNT) calculations with numerical uncertainty quantification due to discretization in ...
Tiago Augusto Santiago Vieira, Felipe Reis Campanha Ribeiro, Yasmim Martins Carvalho, Vitor Vasconcelos Araújo Silva, Graiciany de Paula Barros, Andre Augusto Campagnole dos Santos   +5 more
doaj   +1 more source

Consideration of heterogeneous effects in preparing neutron multigroup constants in the CONSYST/BNAB-RF system [PDF]

Nuclear Energy and Technology, 2020
The need for building mutually self-agreed computational models for high-precision and engineering neutronic codes is defined by requirements to certification and verification of software products and nuclear data in accordance with the Regulations for ...
Olga N. Andrianova, Gennady M. Zherdev, Gleb B. Lomakov, Yevgeniya S. Teplukhina   +3 more
doaj   +3 more sources

FAST REACTOR MULTIPHYSICS AND UNCERTAINTY PROPAGATION WITHIN WIMS [PDF]

EPJ Web of Conferences, 2021
For liquid metal-cooled fast reactors (LMFRs), improved predictive modelling is desirable to facilitate reactor licensing and operation and move towards a best estimate plus uncertainty (BEPU) approach.
Lindley Ben, Tollit Brendan, Smith Peter, Charles Alan, Mason Robert, Ware Tim, Perry Ray, Lavarenne Jean, Davies Una, Gregg Robert   +9 more
doaj   +1 more source

Coupled neutronics/thermal-hydraulic analysis of ANTS-100e using MCS/RAST-F two-step code system

Nuclear Engineering and Technology, 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.
Tung Dong Cao Nguyen, Tuan Quoc Tran, Deokjung Lee   +2 more
doaj   +1 more source

Using effective temperature as a measure of the thermal scattering law uncertainties to UOX fuel calculations from room temperature to 80°C

EPJ Nuclear Sciences & Technologies, 2022
The effective temperature Teff is an important physical quantity in neutronic calculations. It can be introduced in a Free Gas Model to approximate crystal lattice effects in the Doppler broadening of the neutron cross sections. In the last decade, a few
Noguere Gilles, Xu Shuqi
doaj   +1 more source

Loading pattern optimization of a PWR reactor fuel assemblies using PESA-II algorithm by considering neutronic and thermal-hydraulic parameters [PDF]

مجله علوم و فنون هسته‌ای, 2020
The multi-objective optimization methods are widely used in various areas of science and engineering. In this paper, a multi-objective optimization algorithm has been used to design the optimal placement of fuel assemblies in the KWU reactor with the ...
R. Gharari, B. Salmasian, O. Safarzadeh
doaj   +1 more source