Monte Carlo perturbation calculation for geometry change in fixed source problems with the perturbation source method

In the reactor design and analysis, there is often a need to calculate the effects caused by perturbations of temperature, components and even structure of reactors on reactivity. And in sensitivity studies, uncertainty analysis of target quantities and unclear data adjustment, perturbation calculations are also widely used. To meet the need of different types of reactors (complex, multidimensional systems), Monte Carlo perturbation methods have been developed. In this paper, several kinds of perturbation methods are investigated. Specially, differential operator sampling method and correlated tracking method are discussed in details. MCNP’s perturbation calculation capability is discussed by calculating certain problems, from which some conclusions are obtained on the capabilities of the differential operator sampling method used in the perturbation calculation model of MCNP. Also, a code using correlated tracking method has been developed to solve certain problems with cross-section changes, and the results generated by this code agree with the results generated by straightforward Monte Carlo techniques.

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A first collision source method for coupling Monte Carlo and discrete ordinates for localized source problems

Monte-Carlo Methods and Applications in Neutronics, Photonics and Statistical Physics - Lecture Notes in Physics ◽

10.1007/bfb0049062 ◽

2006 ◽

pp. 352-366 ◽

Cited By ~ 3

Author(s):

Raymond E. Alcouffe

Keyword(s):

Monte Carlo ◽

Discrete Ordinates ◽

Localized Source ◽

Source Method

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Monte Carlo shielding design with a large perturbation calculation method

Annals of Nuclear Energy ◽

10.1016/j.anucene.2019.107186 ◽

2020 ◽

Vol 138 ◽

pp. 107186

Author(s):

Wanbin Tan ◽

Guangyao Sun ◽

Pengcheng Long ◽

Bin Wu ◽

Jing Song

Keyword(s):

Monte Carlo ◽

Calculation Method ◽

Perturbation Calculation ◽

Large Perturbation ◽

Shielding Design

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EFFECT OF THE UNIFORM FISSION SOURCE METHOD ON LOCAL POWER VARIANCE IN FULL CORE SERPENT CALCULATION

EPJ Web of Conferences ◽

10.1051/epjconf/202124704024 ◽

2021 ◽

Vol 247 ◽

pp. 04024

Author(s):

Yurii Bilodid ◽

Jaakko Leppänen

Keyword(s):

Monte Carlo ◽

Power Distribution ◽

Reactor Core ◽

Monte Carlo Code ◽

Local Power ◽

Axial Distribution ◽

Source Method ◽

Power Variance ◽

Full Core ◽

Statistical Variance

One of challenges of the Monte Carlo full core simulations is to obtain acceptable statistical variance of local parameters throughout the whole reactor core at a reasonable computation cost. The statistical variance tends to be larger in low-power regions. To tackle this problem, the Uniform-Fission-Site method was implemented in Monte Carlo code MC21 and its effectiveness was demonstrated on NEA Monte Carlo performance benchmark. The very similar method is also implemented in Monte Carlo code Serpent under the name Uniform Fission Source (UFS) method. In this work the effect of UFS method implemented in Serpent is studied on the BEAVRS benchmark which is based on a real PWR core with relatively flat radial power distribution and also on 3x3 PWR mini-core simulated with thermo-hydraulic and thermo-mechanic feedbacks. It is shown that the application of the Uniform Fission Source method has no significant effect on radial power variance but equalizes axial distribution of variance of local power.

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