scholarly journals ANALYSIS OF FAST NEUTRON TRANSPORT IN CHLORIDE SALTS USING MONTE CARLO METHOD

2020 ◽  
Vol 28 ◽  
pp. 42-49
Author(s):  
Ondřej Šťastný ◽  
Karel Katovský ◽  
Dušan Král ◽  
Antonín Krása ◽  
Kamil Števanka
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Fast Neutron ◽  
Nuclear Reactor ◽  
Neutron Transport ◽  
Fissile Material ◽  
Experimental Assembly ◽  
Project Objectives ◽  
Chloride Salts ◽  
International Research Project

The aim of this paper is to present results of fast neutron behavior analysis within the chloride salts environment using simulations based on Monte Carlo method (MCNP 6.2). Three non-fueled salts (NaCl, KCl, MgCl2) and two salts containing fissile material (UCl3, ThCl4) were studied. Results of this theoretical study will be used for design of an experimental assembly, which will serve to achieve goals of the international research project ADAR (Accelerator Driven Advanced Reactor). One of the project objectives is to investigate chloride salts as potential coolant and a dissolved fuel carrier of advanced nuclear reactor cooled by molten salts and driven by an accelerator.

3D Calculations of PWR Vessels Neutron Fluence With EFLUVE 3D Code

2013 ◽  
Author(s):  
Cécile-Aline Gosmain ◽  
Sylvain Rollet ◽  
Damien Schmitt
Keyword(s):  
Monte Carlo ◽  
Fast Neutron ◽  
Pressure Vessel ◽  
Reaction Rate ◽  
Neutron Fluence ◽  
Neutron Transport ◽  
Reaction Rates ◽  
Surveillance Program ◽  
Monte Carlo Code ◽  
Neutron Sources

In the framework of surveillance program dosimetry, the main parameter in the determination of the fracture toughness and the integrity of the reactor pressure vessel (RPV) is the fast neutron fluence on pressure vessel. Its calculated value is extrapolated using neutron transport codes from measured reaction rate value on dosimeters located on the core barrel. EDF R&D has developed a new 3D tool called EFLUVE3D based on the adjoint flux theory. This tool is able to reproduce on a given configuration the neutron flux, fast neutron fluence and reaction rate or dpa results of an exact Monte Carlo calculation with nearly the same accuracy. These EFLUVE3D calculations does the Source*Importance product which allows the calculation of the flux, the neutronic fluence (flux over 1MeV integrated on time) received at any point of the interface between the skin and the pressure vessel but also at the capsules of the pressurized water reactor vessels surveillance program and the dpa and reaction rates at different axial positions and different azimuthal positions of the vessel as well as at the surveillance capsules. Moreover, these calculations can be carried out monthly for each of the 58 reactors of the French current fleet in challenging time (less than 10mn for the total fluence and reaction rates calculations considering 14 different neutron sources of a classical power plant unit compared to more than 2 days for a classic Monte Carlo flux calculation at a given neutron source). The code needs as input: - for each reaction rate, the geometric importance matrix produced for a 3D pin by pin mesh on the basis of Green’s functions calculated by the Monte Carlo code TRIPOLI; - the neutron sources calculated on assemblies data (enrichment, position, fission fraction as a function of evolution), pin by pin power and irradiation. These last terms are based on local in-core activities measurements extrapolated to the whole core by use of the EDF core calculation scheme and a pin by pin power reconstruction methodology. This paper presents the fundamental principles of the code and its validation comparing its results to the direct Monte Carlo TRIPOLI results. Theses comparisons show a discrepancy of less than 0,5% between the two codes equivalent to the order of magnitude of the stochastic convergence of Monte Carlo results.

Simulation of Nuclear Reactor Kinetics by the Monte Carlo Method

2017 ◽  
Vol 80 (8) ◽  
pp. 1399-1407
Author(s):  
E. A. Gomin ◽  
V. D. Davidenko ◽  
A. S. Zinchenko ◽  
I. K. Kharchenko
Keyword(s):  
Monte Carlo ◽  
Monte Carlo Method ◽  
Nuclear Reactor ◽  
Reactor Kinetics ◽  
The Monte Carlo Method ◽  
Nuclear Reactor Kinetics

scholarly journals Solution methods of neutron transport equation in nuclear reactors

2013 ◽  
Vol 14 (2) ◽  
pp. 59
Author(s):  
Mohamad Ali Shafii
Keyword(s):  
Monte Carlo ◽  
Transport Equation ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Neutron Transport ◽  
Collision Probability ◽  
Computer Memory ◽  
Neutron Transport Equation ◽  
Solution Methods ◽  
Transport Approximation

A few numerical methods that usually used to solve neutron transport equation in nuclear reactor are SN dan PN method, Monte Carlo, Collision Probability and Methods of Characteristics . First two methods have been developed using diffusion approach, and last three methods suitable are applicated for transport approximation. Those of three methods have important role in the desain of nuclear reactors. In addition to follow the development of advanced reactor designs, the three methods were chosen because they do not use diffusion approach these are more accurate methods, as well as less need considerable computer memory. Of all the existing methods, the CP method has several advantages among the others. Keywords : Neutron transport, SN, PN, CP, MOC, MC

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