Automatic Whole Core Depletion & Criticality Calculations by MCNPX 2.7.0

Research output: Contribution to report/book/conference proceedings › In-proceedings paper › peer-review

Standard

Automatic Whole Core Depletion & Criticality Calculations by MCNPX 2.7.0. / Kalcheva, Silva; Koonen, Edgar.

Physor 2012. LaGrange Park, IL, United States, 2012.

Research output: Contribution to report/book/conference proceedings › In-proceedings paper › peer-review

Harvard

Kalcheva, S & Koonen, E 2012, Automatic Whole Core Depletion & Criticality Calculations by MCNPX 2.7.0. in Physor 2012. LaGrange Park, IL, United States, 2012 - PHYSOR, Knoxville, Tennessee, United States, 2012-04-15.

APA

Kalcheva, S., & Koonen, E. (2012). Automatic Whole Core Depletion & Criticality Calculations by MCNPX 2.7.0. In Physor 2012

Vancouver

Kalcheva S, Koonen E. Automatic Whole Core Depletion & Criticality Calculations by MCNPX 2.7.0. In Physor 2012. LaGrange Park, IL, United States. 2012

Author

Kalcheva, Silva ; Koonen, Edgar. / Automatic Whole Core Depletion & Criticality Calculations by MCNPX 2.7.0. Physor 2012. LaGrange Park, IL, United States, 2012.

Bibtex - Download

@inproceedings{cad2d9fd625a49f9bfd0fe912a933539,
title = "Automatic Whole Core Depletion & Criticality Calculations by MCNPX 2.7.0",
abstract = "Different approaches to perform automatic whole core criticality & depletion calculations in a research reactor using MCNPX 2.7.0 are presented. An approximate method is to use the existing symmetries of the burned fuel material distribution in the core, i.e., the axial, radial and azimuth symmetries around the core center, in order to significantly reduce the computation time. In this case it is not necessary to give a unique material number to each burn up cell. Cells having similar burn up and power, achieved during similar irradiation history at same initial fuel composition, will experience similar composition evolution and can therefore be given the same material number. To study the impact of the number of unique burn up materials on the computation time and utilized RAM memory, several MCNPX models have been developed. The paper discusses the accuracy of the model on comparison with measurements of BR2 operation cycles in function of the number of unique burn up materials and the impact of the used Q-value (MeV/fission) of the recoverable fission energy.",
keywords = "MCNPX 2.7.0, whole core depletion/criticality analysis, unique burn up material.",
author = "Silva Kalcheva and Edgar Koonen",
note = "Score = 3; 2012 - PHYSOR : Advances in Reactor Physics ; Conference date: 15-04-2012 Through 20-04-2012",
year = "2012",
month = may,
day = "15",
language = "English",
booktitle = "Physor 2012",
url = "http://meetingsandconferences.com/physor2012/",

}

RIS - Download

TY - GEN

T1 - Automatic Whole Core Depletion & Criticality Calculations by MCNPX 2.7.0

AU - Kalcheva, Silva

AU - Koonen, Edgar

N1 - Score = 3

PY - 2012/5/15

Y1 - 2012/5/15

N2 - Different approaches to perform automatic whole core criticality & depletion calculations in a research reactor using MCNPX 2.7.0 are presented. An approximate method is to use the existing symmetries of the burned fuel material distribution in the core, i.e., the axial, radial and azimuth symmetries around the core center, in order to significantly reduce the computation time. In this case it is not necessary to give a unique material number to each burn up cell. Cells having similar burn up and power, achieved during similar irradiation history at same initial fuel composition, will experience similar composition evolution and can therefore be given the same material number. To study the impact of the number of unique burn up materials on the computation time and utilized RAM memory, several MCNPX models have been developed. The paper discusses the accuracy of the model on comparison with measurements of BR2 operation cycles in function of the number of unique burn up materials and the impact of the used Q-value (MeV/fission) of the recoverable fission energy.

AB - Different approaches to perform automatic whole core criticality & depletion calculations in a research reactor using MCNPX 2.7.0 are presented. An approximate method is to use the existing symmetries of the burned fuel material distribution in the core, i.e., the axial, radial and azimuth symmetries around the core center, in order to significantly reduce the computation time. In this case it is not necessary to give a unique material number to each burn up cell. Cells having similar burn up and power, achieved during similar irradiation history at same initial fuel composition, will experience similar composition evolution and can therefore be given the same material number. To study the impact of the number of unique burn up materials on the computation time and utilized RAM memory, several MCNPX models have been developed. The paper discusses the accuracy of the model on comparison with measurements of BR2 operation cycles in function of the number of unique burn up materials and the impact of the used Q-value (MeV/fission) of the recoverable fission energy.

KW - MCNPX 2.7.0

KW - whole core depletion/criticality analysis

KW - unique burn up material.

UR - http://ecm.sckcen.be/OTCS/llisapi.dll/open/ezp_122070

UR - http://knowledgecentre.sckcen.be/so2/bibref/9239

M3 - In-proceedings paper

BT - Physor 2012

CY - LaGrange Park, IL, United States

T2 - 2012 - PHYSOR

Y2 - 15 April 2012 through 20 April 2012

ER -

ID: 342692