HTGR Reactor Physics and Burnup Calculations Using the Serpent Monte Carlo Code

Jaakko Leppänen, M. DeHart

Research output: Contribution to journalArticleScientificpeer-review

19 Citations (Scopus)

Abstract

One of the main advantages of the continuous-energy Monte Carlo method is its versatility and the capability to model any fuel or reactor configuration without major approximations. This capability becomes particularly valuable in studies involving innovative reactor designs and next-generation systems, which often lie beyond the capabilities of deterministic light water reactor transport codes. In this study, a conceptual prismatic high-temperature gas-cooled reactor (HTGR) fuel assembly was modeled using the Serpent Monte Carlo reactor physics burnup calculation code, under development at VTT Technical Research Centre of Finland since 2004. A new geometry model was developed for the Serpent code in order to account for the heterogeneity effects encountered in randomly dispersed particle fuels. The results are compared to other Monte Carlo and deterministic transport codes, and the study also serves as a test case for the modules and methods in SCALE 6.
Original languageEnglish
Pages (from-to)782-784
Number of pages3
JournalTransactions of the American Nuclear Society
Volume101
Publication statusPublished - 2009
MoE publication typeA1 Journal article-refereed

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reactor physics
high temperature gas cooled reactors
light water reactors
reactor design
Finland
nuclear fuels
versatility
Monte Carlo method
assembly
modules
reactors
geometry
configurations
approximation

Cite this

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HTGR Reactor Physics and Burnup Calculations Using the Serpent Monte Carlo Code. / Leppänen, Jaakko; DeHart, M.

In: Transactions of the American Nuclear Society, Vol. 101, 2009, p. 782-784.

Research output: Contribution to journalArticleScientificpeer-review

TY - JOUR

T1 - HTGR Reactor Physics and Burnup Calculations Using the Serpent Monte Carlo Code

AU - Leppänen, Jaakko

AU - DeHart, M.

PY - 2009

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AB - One of the main advantages of the continuous-energy Monte Carlo method is its versatility and the capability to model any fuel or reactor configuration without major approximations. This capability becomes particularly valuable in studies involving innovative reactor designs and next-generation systems, which often lie beyond the capabilities of deterministic light water reactor transport codes. In this study, a conceptual prismatic high-temperature gas-cooled reactor (HTGR) fuel assembly was modeled using the Serpent Monte Carlo reactor physics burnup calculation code, under development at VTT Technical Research Centre of Finland since 2004. A new geometry model was developed for the Serpent code in order to account for the heterogeneity effects encountered in randomly dispersed particle fuels. The results are compared to other Monte Carlo and deterministic transport codes, and the study also serves as a test case for the modules and methods in SCALE 6.

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JO - Transactions of the American Nuclear Society

JF - Transactions of the American Nuclear Society

SN - 0003-018X

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