The issue of fission source convergence in Monte Carlo eigenvalue calculations is of interest because of the potential consequences of erroneous criticality safety calculations. In this work, the authors compare two different techniques to improve the source convergence behavior of standard Monte Carlo calculations applied to challenging source convergence problems. The first method, super-history powering, attempts to avoid discarding important fission sites between generations by delaying stochastic sampling of the fission site bank until after several generations of multiplication. The second method, stratified sampling of the fission site bank, explicitly keeps the important sites even if conventional sampling would have eliminated them...
In 1995, at a conference on criticality safety, a special session was devoted to the Monte Carlo ''E...
This paper presents a new response matrix based solver implemented in the Serpent 2 Monte Carlo code...
Abstract: In Monte Carlo (MC) criticality simulation, fission neutrons accumulated in a generation n...
The fission source convergence of a very loosely coupled array of 36 fuel subassemblies with slightl...
We compare nominal efficiencies, i.e. variances in power shapes for equal running time, of different...
In order to avoid calculation bias and increase calculation efficiency, convergence of the fission s...
This paper discusses the source convergence of Monte Carlo calculations for α-eigenvalues. Compared ...
Fission source convergence in Monte Carlo criticality calculations can be difficult for some types o...
We propose a fission source convergence acceleration method for Monte Carlo criticality simulation. ...
Because of the accuracy and ease of implementation, the Monte Carlo methodology is widely used in th...
In Monte Carlo (MC) criticality calculations, source error propagation through the stationary cycles...
Criticality calculations use the source iteration method and serve an increasingly prominent role in...
In this work we describe a new method for parallelizing the source iterations in a Monte Carlo criti...
The use of Monte Carlo, random number sampling, for neutron transport has been used for about half a...
Nuclear criticality calculations with Monte Carlo codes are normally done using a power iteration me...
In 1995, at a conference on criticality safety, a special session was devoted to the Monte Carlo ''E...
This paper presents a new response matrix based solver implemented in the Serpent 2 Monte Carlo code...
Abstract: In Monte Carlo (MC) criticality simulation, fission neutrons accumulated in a generation n...
The fission source convergence of a very loosely coupled array of 36 fuel subassemblies with slightl...
We compare nominal efficiencies, i.e. variances in power shapes for equal running time, of different...
In order to avoid calculation bias and increase calculation efficiency, convergence of the fission s...
This paper discusses the source convergence of Monte Carlo calculations for α-eigenvalues. Compared ...
Fission source convergence in Monte Carlo criticality calculations can be difficult for some types o...
We propose a fission source convergence acceleration method for Monte Carlo criticality simulation. ...
Because of the accuracy and ease of implementation, the Monte Carlo methodology is widely used in th...
In Monte Carlo (MC) criticality calculations, source error propagation through the stationary cycles...
Criticality calculations use the source iteration method and serve an increasingly prominent role in...
In this work we describe a new method for parallelizing the source iterations in a Monte Carlo criti...
The use of Monte Carlo, random number sampling, for neutron transport has been used for about half a...
Nuclear criticality calculations with Monte Carlo codes are normally done using a power iteration me...
In 1995, at a conference on criticality safety, a special session was devoted to the Monte Carlo ''E...
This paper presents a new response matrix based solver implemented in the Serpent 2 Monte Carlo code...
Abstract: In Monte Carlo (MC) criticality simulation, fission neutrons accumulated in a generation n...