Lead Cooled Fast Reactor Core Physics Parameters Calculation Based on Monte Carlo Method

2017 ◽  
pp. 783-791
Author(s):  
Yang Lyu
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Fast Reactor ◽  
Reactor Core ◽  
Parameters Calculation

Startup of “CANDLE” burnup in a Gas-cooled Fast Reactor using Monte Carlo method

2012 ◽  
Vol 50 ◽  
pp. 44-49 ◽  
Author(s):  
M. Kheradmand Saadi ◽  
A. Abbaspour ◽  
A. Pazirandeh
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Fast Reactor

scholarly journals The effect of different fuels and clads on neutronic calculations in a boiling water reactor using the Monte Carlo method

2020 ◽  
Vol 10 (1) ◽  
Author(s):  
Mehtap Düz ◽  
Selcan İnal
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Reactor Core ◽  
Nuclear Data ◽  
Boiling Water ◽  
Square Lattice ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
The Monte Carlo Method ◽  
Nuclear Data Library

AbstractIn this study, a Boiling Water Reactor (BWR) modeling was done for the reactor core divided into square lattice 8 × 8 type using the Monte Carlo Method. Each of the square lattices in the reactor core was divided into small square lattices 7 × 7 type in groups of four. In the BWR designed in this study, modeling was made on fuel assemblies at pin-by-pin level by using neptunium mixed fuels as fuel rod, Zr-2 and SiC as fuel cladding, H2O as coolant. In fuel rods were used NpO2 and NpF4 fuels at the rate of 0.2%-1% as neptunium mixed fuels. In this study, the effect on the neutronic calculations as keff, neutron flux, fission energy, heating of NpO2 and NpF4 fuels in 0.2%-1% rates, and Zr-2 and SiC clads were investigated in the designed BWR system. The three-dimensional (3-D) modelling of the reactor core and fuel assembly into the designed BWR system was performed by using MCNPX-2.7.0 Monte Carlo method and the ENDF/B-VII.0 nuclear data library.

scholarly journals 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

2020 ◽  
Vol 6 (4) ◽  
pp. 269-274
Author(s):  
Olga N. Andrianova ◽  
Evgeniya S. Teplukhina ◽  
Gennady M. Zherdev ◽  
Zhanna V. Borovskaya ◽  
Andrey P. Zhirnov
Keyword(s):  
Monte Carlo ◽  
Fast Reactor ◽  
Experimental Studies ◽  
Reactor Core ◽  
Heterogeneous Structure ◽  
Neutron Data ◽  
The Monte Carlo Method ◽  
Critical Facilities ◽  
Data Library ◽  
Neutronic Characteristics

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 different years at the BFS critical assemblies by analyzing and revising earlier calculation and experimental studies on the transmission of neutrons through the steel reflector layers. The discussion includes experiments at the BFS-66 critical assembly to model neutron and photon fluxes in the reactor core shielding compositions, as well as experiments at the BFS-64 and BFS-80-2 critical assemblies to model the transmission of neutrons and gamma quanta through the reflector layers of various materials. The information provided in earlier materials with the descriptions of the above experiments has been analyzed and expanded through respective data required to prepare precision calculation models for Monte-Carlo neutronic codes. Precision neutronic models have been developed based on actualized and updated data with a detailed description of the BFS heterogeneous structure and experimental devices, and test calculations have been carried out to confirm their efficiency. The calculations of key neutronic characteristics measured at the BFS-66, -64 and -80-2 assemblies were performed using codes based on the Monte Carlo method (MCU-BR, MCNP, MMK-RF, MMK-ROKOKO) with BNAB-RF and MDBBR50 neutron data and the ROSFOND evaluated neutron data library. The developed precision calculation neutronic models of the experiments discussed can be used to justify lead cooled fast reactor designs, to verify neutronic codes and neutron data, and to evaluate the associated uncertainties.

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