This paper presents the application and evaluation of a deterministic truncation of Monte Carlo (DTMC) solution method in a whole core reactor problem based on a continuous energy transport calculation. The DTMC method has been studied and developed as a systematic way to truncate the high-fidelity Monte Carlo (MC) solution to reduce the computational cost without compromising the essential reliability of the solution. Its fea-sibility and capability were preliminarily validated in several benchmark problems using a multi-group energy MC code. In this paper, further study has been conducted in the more practical application with the continuous-energy based MC calculation. The con-cept of the DTMC method is briefly described. Improvements to...
The research team has developed a practical, high-order, discrete-ordinates, short characteristics n...
For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes ...
Thesis: Ph. D. in Computational Nuclear Science and Engineering, Massachusetts Institute of Technolo...
This thesis investigates the feasibility of Monte Carlo (MC) neutronic calculations that combine mul...
This paper presents a new hybrid method of continuous energy Monte Carlo (MC) and multi-group Method...
The accurate modeling of nuclear reactors is essential to the safe and economic operation of current...
A considerable speed-up in continuous-energy Monte Carlo neutron transport calculation can be achiev...
This paper is a general overview of the Serpent Monte Carlo reactor physics burnup calculation code....
Full-core reactor dynamics calculation involves the coupled modelling of thermal hydraulics and the ...
A whole-core depletion analysis of OPR-1000 using MCS coupled with FRAPCON is presented in this pape...
Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditi...
In many reactor calculations, high fidelity, high accuracy results are required only in a small spat...
Over the past two decades, criticality safety analysts have come to rely to a large extent on Monte ...
M.Sc.Due to the demand for medical isotopes, new Materials Testing Reactors (MTR's) are being consid...
This paper presents an evaluation of a Monte Carlo (MC) neutron transport code as a tool to generate...
The research team has developed a practical, high-order, discrete-ordinates, short characteristics n...
For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes ...
Thesis: Ph. D. in Computational Nuclear Science and Engineering, Massachusetts Institute of Technolo...
This thesis investigates the feasibility of Monte Carlo (MC) neutronic calculations that combine mul...
This paper presents a new hybrid method of continuous energy Monte Carlo (MC) and multi-group Method...
The accurate modeling of nuclear reactors is essential to the safe and economic operation of current...
A considerable speed-up in continuous-energy Monte Carlo neutron transport calculation can be achiev...
This paper is a general overview of the Serpent Monte Carlo reactor physics burnup calculation code....
Full-core reactor dynamics calculation involves the coupled modelling of thermal hydraulics and the ...
A whole-core depletion analysis of OPR-1000 using MCS coupled with FRAPCON is presented in this pape...
Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditi...
In many reactor calculations, high fidelity, high accuracy results are required only in a small spat...
Over the past two decades, criticality safety analysts have come to rely to a large extent on Monte ...
M.Sc.Due to the demand for medical isotopes, new Materials Testing Reactors (MTR's) are being consid...
This paper presents an evaluation of a Monte Carlo (MC) neutron transport code as a tool to generate...
The research team has developed a practical, high-order, discrete-ordinates, short characteristics n...
For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes ...
Thesis: Ph. D. in Computational Nuclear Science and Engineering, Massachusetts Institute of Technolo...