Monte Carlo neutronics benchmarks on nuclear fuel depletion: A review

Sean P. Martinson; Sunil S. Chirayath

doi:10.1016/j.anucene.2021.108441

Monte Carlo neutronics benchmarks on nuclear fuel depletion: A review

Sean P. Martinson, Sunil S. Chirayath

Research output: Contribution to journal › Review article › peer-review

8 Scopus citations

Abstract

Monte Carlo (MC) neutronics codes are used widely for academic and industrial needs. Several schemes of coupling MC neutronics code with isotope generation and depletion code exist, which are used for performing nuclear fuel depletion simulations. These simulations can estimate the inventory of isotopes in neutron irradiated nuclear reactor fuel. However, the accuracy of these simulations shall be validated through experiments. MC codes are seldom validated by isotopic benchmarks compared to criticality benchmarks. This work compiles and analyzes the fuel depletion benchmarks and validations used to analyze the performance of MC-based fuel depletion neutronics codes. Analyses of these benchmarks and validations showed that the computed concentrations of ¹³³Cs, ¹³⁵Cs, ¹³⁷Cs, ¹⁴⁸Nd, ²³⁹Pu, ²⁴⁰Pu, and ²⁴¹Pu in the irradiated fuel by the depletion codes agreed with the measured values within 10% error. However, the computed concentrations of ¹²⁵Sb, ²⁴²Cm, ²⁴³Cm, ²⁴⁴Cm, ²⁴⁵Cm, and ²⁴⁶Cm had errors more than 15% compared to the measured values. Ventina depletion code showed the most accurate predictions for the greatest number of isotope concentrations compared to ORIGEN2 and CINDER90.

Original language	English
Article number	108441
Journal	Annals of Nuclear Energy
Volume	161
DOIs	https://doi.org/10.1016/j.anucene.2021.108441
State	Published - Oct 2021
Externally published	Yes

Funding

This work was funded by the Consortium for Monitoring, Technology, and Verification under Department of Energy National Nuclear Security Administration award number DE-NA0003920. The opinions expressed in this article are the authors’ own and do not reflect the view of the National Nuclear Security Administration, the Department of Nuclear Energy, or the United States government. Permissions to use figures and tables were received from Elsevier, Dr. Mark DeHart, Dr. Hugo Dalle, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS), and the Lithuanian Journal of Physics.

Funders	Funder number
National Nuclear Security Administration	DE-NA0003920

Keywords

KENO
MCNP
MONTEBURNS
Monte Carlo neutronics codes
SERPENT

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title = "Monte Carlo neutronics benchmarks on nuclear fuel depletion: A review",

abstract = "Monte Carlo (MC) neutronics codes are used widely for academic and industrial needs. Several schemes of coupling MC neutronics code with isotope generation and depletion code exist, which are used for performing nuclear fuel depletion simulations. These simulations can estimate the inventory of isotopes in neutron irradiated nuclear reactor fuel. However, the accuracy of these simulations shall be validated through experiments. MC codes are seldom validated by isotopic benchmarks compared to criticality benchmarks. This work compiles and analyzes the fuel depletion benchmarks and validations used to analyze the performance of MC-based fuel depletion neutronics codes. Analyses of these benchmarks and validations showed that the computed concentrations of 133Cs, 135Cs, 137Cs, 148Nd, 239Pu, 240Pu, and 241Pu in the irradiated fuel by the depletion codes agreed with the measured values within 10% error. However, the computed concentrations of 125Sb, 242Cm, 243Cm, 244Cm, 245Cm, and 246Cm had errors more than 15% compared to the measured values. Ventina depletion code showed the most accurate predictions for the greatest number of isotope concentrations compared to ORIGEN2 and CINDER90.",

keywords = "KENO, MCNP, MONTEBURNS, Monte Carlo neutronics codes, SERPENT",

author = "Martinson, {Sean P.} and Chirayath, {Sunil S.}",

note = "Publisher Copyright: {\textcopyright} 2021 Elsevier Ltd",

year = "2021",

month = oct,

doi = "10.1016/j.anucene.2021.108441",

language = "English",

volume = "161",

journal = "Annals of Nuclear Energy",

issn = "0306-4549",

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TY - JOUR

T1 - Monte Carlo neutronics benchmarks on nuclear fuel depletion

T2 - A review

AU - Martinson, Sean P.

AU - Chirayath, Sunil S.

PY - 2021/10

Y1 - 2021/10

N2 - Monte Carlo (MC) neutronics codes are used widely for academic and industrial needs. Several schemes of coupling MC neutronics code with isotope generation and depletion code exist, which are used for performing nuclear fuel depletion simulations. These simulations can estimate the inventory of isotopes in neutron irradiated nuclear reactor fuel. However, the accuracy of these simulations shall be validated through experiments. MC codes are seldom validated by isotopic benchmarks compared to criticality benchmarks. This work compiles and analyzes the fuel depletion benchmarks and validations used to analyze the performance of MC-based fuel depletion neutronics codes. Analyses of these benchmarks and validations showed that the computed concentrations of 133Cs, 135Cs, 137Cs, 148Nd, 239Pu, 240Pu, and 241Pu in the irradiated fuel by the depletion codes agreed with the measured values within 10% error. However, the computed concentrations of 125Sb, 242Cm, 243Cm, 244Cm, 245Cm, and 246Cm had errors more than 15% compared to the measured values. Ventina depletion code showed the most accurate predictions for the greatest number of isotope concentrations compared to ORIGEN2 and CINDER90.

AB - Monte Carlo (MC) neutronics codes are used widely for academic and industrial needs. Several schemes of coupling MC neutronics code with isotope generation and depletion code exist, which are used for performing nuclear fuel depletion simulations. These simulations can estimate the inventory of isotopes in neutron irradiated nuclear reactor fuel. However, the accuracy of these simulations shall be validated through experiments. MC codes are seldom validated by isotopic benchmarks compared to criticality benchmarks. This work compiles and analyzes the fuel depletion benchmarks and validations used to analyze the performance of MC-based fuel depletion neutronics codes. Analyses of these benchmarks and validations showed that the computed concentrations of 133Cs, 135Cs, 137Cs, 148Nd, 239Pu, 240Pu, and 241Pu in the irradiated fuel by the depletion codes agreed with the measured values within 10% error. However, the computed concentrations of 125Sb, 242Cm, 243Cm, 244Cm, 245Cm, and 246Cm had errors more than 15% compared to the measured values. Ventina depletion code showed the most accurate predictions for the greatest number of isotope concentrations compared to ORIGEN2 and CINDER90.

KW - KENO

KW - MCNP

KW - MONTEBURNS

KW - Monte Carlo neutronics codes

KW - SERPENT

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U2 - 10.1016/j.anucene.2021.108441

DO - 10.1016/j.anucene.2021.108441

M3 - Review article

AN - SCOPUS:85108693396

SN - 0306-4549

VL - 161

JO - Annals of Nuclear Energy

JF - Annals of Nuclear Energy

M1 - 108441

ER -

Monte Carlo neutronics benchmarks on nuclear fuel depletion: A review

Abstract

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Keywords

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