Correlation ν̅ p − σ − χ in the fast neutron range via integral information

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Abstract. This paper presents a study of the impact of the nuclear data (cross sections, neutron emission and spectra) on different quantities for spent nuclear fuels (SNF) from Swiss power plants: activities, decay heat, neutron and gamma sources and isotopic vectors. Realistic irradiation histories are considered using validated core follow-up models based on CASMO and SIMULATE. Two Pressurized and one Boiling Water Reactors (PWR and BWR) are considered over a large number of operated cycles. All the assemblies at the end of the cycles are studied, being reloaded or finally discharged, allowing spanning over a large range of exposure (from 4 to 60 MWd/kgU for ≃9200 assembly-cycles). Both UO 2 and MOX fuels were used during the reactor cycles, with enrichments from 1.9 to 4.7% for the UO 2 and 2.2 to 5.8% Pu for the MOX. The SNF characteristics presented in this paper are calculated with the SNF code. The calculated uncertainties, based on the ENDF/B-VII.1 library are obtained using a simple Monte Carlo sampling method. It is demonstrated that the impact of nuclear data is relatively important (e.g. up to 17% for the decay heat), showing the necessity to consider them for safety analysis of the SNF handling and disposal.

Section: Correlationmentioning

confidence: 99%

Uncertainties for Swiss LWR spent nuclear fuels due to nuclear data

Rochman

Vasiliev

Dokhane

et al. 2018

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“…One of those reasons is that integral data cannot provide a unique set of nuclear data that represent the measured data. To again consider the example of Jezebel, many combinations of PFNS, v and s f of 239 Pu would be consistent with the measured data, leading to correlations [2] not taken into account in current evaluations. Smaller effects, such as impurities of 240 Pu, would also impact the result.…”

Section: Use Of All Informationmentioning

confidence: 99%

“…The Jezebel 239 Pu critical assembly shown above is widely used by nuclear data evaluators to constrain their evaluations of neutron-induced reactions on 239 Pu, creating hidden correlations between different quantities such as v, PFNS and s f , as discussed by Bauge and Rochman[1] and Rochman et al[2].…”

mentioning

confidence: 99%

Evaluating nuclear data and their uncertainties

Talou

2018

In the last decade or so, estimating uncertainties associated with nuclear data has become an almost mandatory step in any new nuclear data evaluation. The mathematics needed to infer such estimates look deceptively simple, masking the hidden complexities due to imprecise and contradictory experimental data and natural limitations of simplified physics models. Through examples of evaluated covariance matrices for the soon-to-be-released U.S. ENDF/B-VIII.0 library, e.g., cross sections, spectrum, multiplicity, this paper discusses some uncertainty quantification methodologies in use today, their strengths, their pitfalls, and alternative approaches that have proved to be highly successful in other fields. The important issue of how to interpret and use the covariance matrices coming out of the evaluated nuclear data libraries is discussed. *

“…The standard deviation of this distribution represents the nuclear data propagated uncertainty for the reaction rate. Note that for the Bayesian Monte Carlo (BMC) [6,17] assimilation step all the individual reaction rates for each ACE file are required.…”

Section: Total Monte Carlomentioning

confidence: 99%

Uncertainty propagation for the design study of the PETALE experimental programme in the CROCUS reactor

Laureau

Lamirand

Rochman

et al. 2020

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The PETALE experimental programme in the CROCUS reactor intends to provide integral measurements to constrain stainless steel nuclear data. This article presents the tools and the methodology developed to design and optimize the experiments, and its operating principle. Two acceleration techniques have been implemented in the Serpent2 code to perform a Total Monte Carlo uncertainty propagation using variance reduction and correlated sampling technique. Their application to the estimation of the expected reaction rates in dosimeters is also discussed, together with the estimation of the impact of the nuisance parameters of aluminium used in the experiment structures. This is an Open Access article distributed under the terms of the Creative Commons Attribution License (http://creativecommons.org/licenses/by/4.0), which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited. ,