### Skapa referens, olika format (klipp och klistra)

**Harvard**

Hernández-Solís, A., Demazière, C. och Ekberg, C. (2013) *Uncertainty Analyses Applied to the UAM/TMI-1 Lattice Calculations Using the DRAGON (Version 4.05) Code and Based on JENDL-4 and ENDF/B-VII.1 Covariance Data*.

** BibTeX **

@article{

Hernández-Solís2013,

author={Hernández-Solís, Augusto and Demazière, Christophe and Ekberg, Christian},

title={Uncertainty Analyses Applied to the UAM/TMI-1 Lattice Calculations Using the DRAGON (Version 4.05) Code and Based on JENDL-4 and ENDF/B-VII.1 Covariance Data},

journal={Science and Technology of Nuclear Installations},

issn={1687-6075},

abstract={The OECD/NEA Uncertainty Analysis in Modeling (UAM) expert group organized and
launched the UAM benchmark. Its main objective is to perform uncertainty analysis in light
water reactor (LWR) predictions at all modeling stages. In this paper, multigroup microscopic
cross-sectional uncertainties are propagated through the DRAGON (version 4.05) lattice code in
order to perform uncertainty analysis on kinf and 2-group homogenized macroscopic cross-sections.
The chosen test case corresponds to the Three Mile Island-1 (TMI-1) lattice, which is
a 15*15 pressurized water reactor (PWR) fuel assembly segment with poison and at full
power conditions. A statistical methodology is employed for the uncertainty assessment, where
cross-sections of certain isotopes of various elements belonging to the 172-group DRAGLIB
library format are considered as normal random variables. Two libraries were created for such
purposes, one based on JENDL-4 data and the other one based on the recently released
ENDF/B-VII.1 data. Therefore, multigroup uncertainties based on both nuclear data libraries
needed to be computed for the different isotopic reactions by means of ERRORJ. The
uncertainty assessment performed on kinf and macroscopic cross-sections, that is based on
JENDL-4 data, was much higher than the assessment based on ENDF/B-VII.1 data. It was
found that the computed Uranium 235 fission covariance matrix based on JENDL-4 is much
larger at the thermal and resonant regions than, for instance, the covariance matrix based on
ENDF/B-VII.1 data. This can be the main cause of significant discrepancies between different
uncertainty assessments.},

year={2013},

}

** RefWorks **

RT Journal Article

SR Electronic

ID 173347

A1 Hernández-Solís, Augusto

A1 Demazière, Christophe

A1 Ekberg, Christian

T1 Uncertainty Analyses Applied to the UAM/TMI-1 Lattice Calculations Using the DRAGON (Version 4.05) Code and Based on JENDL-4 and ENDF/B-VII.1 Covariance Data

YR 2013

JF Science and Technology of Nuclear Installations

SN 1687-6075

AB The OECD/NEA Uncertainty Analysis in Modeling (UAM) expert group organized and
launched the UAM benchmark. Its main objective is to perform uncertainty analysis in light
water reactor (LWR) predictions at all modeling stages. In this paper, multigroup microscopic
cross-sectional uncertainties are propagated through the DRAGON (version 4.05) lattice code in
order to perform uncertainty analysis on kinf and 2-group homogenized macroscopic cross-sections.
The chosen test case corresponds to the Three Mile Island-1 (TMI-1) lattice, which is
a 15*15 pressurized water reactor (PWR) fuel assembly segment with poison and at full
power conditions. A statistical methodology is employed for the uncertainty assessment, where
cross-sections of certain isotopes of various elements belonging to the 172-group DRAGLIB
library format are considered as normal random variables. Two libraries were created for such
purposes, one based on JENDL-4 data and the other one based on the recently released
ENDF/B-VII.1 data. Therefore, multigroup uncertainties based on both nuclear data libraries
needed to be computed for the different isotopic reactions by means of ERRORJ. The
uncertainty assessment performed on kinf and macroscopic cross-sections, that is based on
JENDL-4 data, was much higher than the assessment based on ENDF/B-VII.1 data. It was
found that the computed Uranium 235 fission covariance matrix based on JENDL-4 is much
larger at the thermal and resonant regions than, for instance, the covariance matrix based on
ENDF/B-VII.1 data. This can be the main cause of significant discrepancies between different
uncertainty assessments.

LA eng

DO 10.1155/2013/437854

LK http://dx.doi.org/10.1155/2013/437854

LK http://publications.lib.chalmers.se/records/fulltext/173347/local_173347.pdf

OL 30