Development and validation of ALEPH Monte Carlo burn-up code

Alexey Stankovskiy; Gert Van den Eynde; Tim Vidmar; Edouard Malambu Mbala

Development and validation of ALEPH Monte Carlo burn-up code

Alexey Stankovskiy, Gert Van den Eynde, Tim Vidmar, Edouard Malambu Mbala

Research output › peer-review

Abstract

The Monte-Carlo burn-up code ALEPH is being developed in SCKCEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and "deterministic" depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH.

Original language	English
Title of host publication	Nuclear measurements, Evaluations and Applications - NEMEA-6
Place of Publication	Paris, France
Pages	161-170
State	Published - 11 Aug 2011
Event	Nuclear measurements, Evaluations and Applications - NEMEA-6 - Krakow Duration: 25 Oct 2010 → 28 Oct 2010

Conference

Conference	Nuclear measurements, Evaluations and Applications - NEMEA-6
Country/Territory	Poland
City	Krakow
Period	2010-10-25 → 2010-10-28

Cite this

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title = "Development and validation of ALEPH Monte Carlo burn-up code",

abstract = "The Monte-Carlo burn-up code ALEPH is being developed in SCKCEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and {"}deterministic{"} depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH.",

keywords = "burn-up, nuclear data, differential equations",

author = "Alexey Stankovskiy and {Van den Eynde}, Gert and Tim Vidmar and {Malambu Mbala}, Edouard",

note = "Score = 3; Nuclear measurements, Evaluations and Applications - NEMEA-6 ; Conference date: 25-10-2010 Through 28-10-2010",

year = "2011",

month = aug,

day = "11",

language = "English",

pages = "161--170",

booktitle = "Nuclear measurements, Evaluations and Applications - NEMEA-6",

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TY - GEN

T1 - Development and validation of ALEPH Monte Carlo burn-up code

AU - Stankovskiy, Alexey

AU - Van den Eynde, Gert

AU - Vidmar, Tim

A2 - Malambu Mbala, Edouard

N1 - Score = 3

PY - 2011/8/11

Y1 - 2011/8/11

N2 - The Monte-Carlo burn-up code ALEPH is being developed in SCKCEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and "deterministic" depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH.

AB - The Monte-Carlo burn-up code ALEPH is being developed in SCKCEN since 2004. Belonging to the category of shells coupling Monte Carlo transport (MCNP or MCNPX) and "deterministic" depletion codes (ORIGEN-2.2), ALEPH possess some unique features that distinguish it from other codes. The most important feature is full data consistency between steady-state Monte Carlo and time-dependent depletion calculations. Recent improvements of ALEPH concern full implementation of general-purpose nuclear data libraries (JEFF-3.1.1, ENDF/B-VII, JENDL-3.3). The upgraded version of the code is capable to treat isomeric branching ratios, neutron induced fission product yields, spontaneous fission yields and energy release per fission recorded in ENDF-formatted data files. The alternative algorithm for time evolution of nuclide concentrations is added. A predictor-corrector mechanism and the calculation of nuclear heating are available as well. The validation of the code on REBUS experimental programme results has been performed. The upgraded version of ALEPH has shown better agreement with measured data than other codes, including previous version of ALEPH.

KW - burn-up

KW - nuclear data

KW - differential equations

UR - http://ecm.sckcen.be/OTCS/llisapi.dll/open/ezp_114741

UR - http://knowledgecentre.sckcen.be/so2/bibref/8190

M3 - In-proceedings paper

SP - 161

EP - 170

BT - Nuclear measurements, Evaluations and Applications - NEMEA-6

CY - Paris, France

T2 - Nuclear measurements, Evaluations and Applications - NEMEA-6

Y2 - 25 October 2010 through 28 October 2010

ER -