The Similarity/Transposition theory to Assess Accurately PWR-MOx 15x15 Used Fuel Inventory with Darwin2.3

International audienceThe DARWIN2.3 package, dedicated to the characterization of spent fuels from reactors, benefits from a broad experimental validation database for the isotopic inventory of 17x17 PWR mixed oxide (MOx) fuels. However, a lack is to notice for fuels in 15x15 configurations indeed, no data are available in the post irradiation examination (PIEs) database for these fuels. Under those circumstances, the CEA has decided to study the possibility to make use of the experimental validation available for MOx fuels to assess accurately 15x15 PWR-MOx fuels depletion calculation results. This paper focuses on preliminary investigations on the use of the similarity/transposition approach on 17x17 PWR-MOx fuel rod depletion calculations to use in 15x15 PWR-MOx fuel rod characterization

Application of the bias transposition method on PWR decay heat calculations with the DARWIN2.3 package

International audienceAn accurate estimation of the decay heat associated with controlled bias and uncertainty is of paramount importance for the design and the operation of future and current reactors as well as for the back- end cycle. The DARWIN2.3 package, dedicated to the characterization of spent fuels from reactors, benefits from the Verification, Validation and Uncertainty Quantification process. The goal of this paper is to illustrate how to exploit the corpus of integral experimental values for the decay heat though the bias transposition method in order to improve the accuracy of the decay heat calculations for industrial PWR reactors. In particular, parametric studies are conducted on the decay heat time dependence and discharge burn-up

How to get an enhanced extended uncertainty associated with decay heat calculations of industrial PWRs with the DARWIN2.3 package

International audienceThe decay heat is a crucial issue for in-core safety after reactor shutdown and back-end cycle. An accurate computation of its value is done at the CEA within the DARWIN2.3 package. The DARWIN2.3 package benefits from a Verification, Validation and Uncertainty Quantification (VVUQ) process. The VVUQ ensures that the parameters of interest computed with the DARWIN2.3 package have been validated over experimental measurements and that biases and uncertainties have been quantified for a particular domain. For the parameter decay heat, there are few integral experiments available to ensure the experimental validation over the whole range of parameters needed to cover the French reactor fleet (fissile content, burnup, fuel, cooling time). The experimental validation currently covers PWR UOX fuels for cooling times only between 45 minutes and 42 days, and between 13 and 23 years. Therefore the uncertainty quantification step is of paramount importance in order to increase the reliability and accuracy of decay heat calculations. This paper focuses on the strategy that could be used to answer this issue with the complement and the exploitation of the DARWIN2.3 experimental validation

NUCLEAR DATA UNCERTAINTY QUANTIFICATION FOR THE DECAY HEAT OF PWR MOX FUELS USING DATA ASSIMILATION OF ELEMENTARY FISSION BURSTS

Currently there is no integral experimental data for code validation regarding the decay heat of MOX fuels, excepted fission burst experiments (for fission products contributions at short cooling times) or post-irradiated experiments on nuclide inventories (restricted number of nuclide of interest for decay heat). The uncertainty quantification mainly relies on uncertainty propagation of nuclear data covariances. In the recent years, the transposition method, based on the data assimilation theory, was used in order to transpose the experiment-to-calculation discrepancies at a given set of parameters (cooling time, fuel burnup) to another set of parameters. As an example, this method was used on the CLAB experiments and the experiment-to-calculation discrepancies at 13 years were transposed to an UOX fuel between 5 and 27 years and for burnups from 10 to 50 GWd/t. The purpose of this paper is to study to what extent the transposition method could be used for MOX fuels. In particular, the Dickens fission burst experiment of 239Pu was considered for MOX fuels at short cooling times (< 1h30) and low burnup (< 10 GWd/t). The impact of fission yields (FY) correlations was also discussed. As a conclusion, the efficiency of the transposition process is limited by the experimental uncertainties larger than nuclear data uncertainties, and by the fact that fission burst experiments would only be representative to the FY contribution of the decay heat uncertainty of an irradiated reactor fuel. Nevertheless, this method strengthens the decay heat uncertainties at very short cooling times, previously based only on nuclear data covariance propagation through computation

Work plan for improving the DARWIN2.3 depleted material balance calculation of nuclides of interest for the fuel cycle

International audienceDARWIN2.3 is the reference package used for fuel cycle applications in France. It solves the Boltzmann and Bateman equations in a coupling way, with the European JEFF-3.1.1 nuclear data library, to compute the fuel cycle values of interest. It includes both deterministic transport codes APOLLO2 (for light water reactors) and ERANOS2 (for fast reactors), and the DARWIN-PEPIN2 depletion code, each of them being developed by CEA-DEN with the support of its industrial partners. The DARWIN2.3 package has been experimentally validated for pressurized and boiling water reactors, as well as for sodium fast reactors; this experimental validation relies on the analysis of post-irradiation experiments (PIE). The DARWIN2.3 experimental validation work points out some isotopes for which the depleted concentration calculation can be improved. Some other nuclides have no available experimental validation, and their concentration calculation uncertainty is provided by the propagation of a priori nuclear data uncertainties. This paper describes the work plan of studies initiated this year to improve the accuracy of the DARWIN2.3 depleted material balance calculation concerning some nuclides of interest for the fuel cycle

Integral data assimilation of the MERCI-1 experiment for the nuclear data associated with the PWR decay heat computation

Nuclear decay heat is a crucial issue for PWR in-core safety after reactor shutdown and back-end cycle. It is a dimensioning parameter for safety injection systems (SIS) to avoid a dewatering of the reactor core. The decay heat uncertainty needs to be controlled over the largest range of applications. The assimilation of the MERCI-1 experiment was studied to provide feedbacks on nuclear data. This experiment consisted in the measurement of the decay heat of a PWR UOX fuel sample irradiated in the OSIRIS reactor, for cooling times between 45 minutes and 42 days. More specifically, the consideration of several experimental values of MERCI-1 at different cooling times was tested. This raised issues about correlations to consider between different measurements. Besides, the impact of considering correlations between independent fission yields in covariance matrices on the decay heat uncertainty calculation and on the feedbacks on nuclear data is discussed

Assessment of the 153^{153}Eu and 154^{154}Eu neutron capture cross sections from the Integral Data Assimilation of spent nuclear fuel experiments

International audienceThe purpose of this paper is to make the most of a large set of used nuclear fuel experiments for determining integral trends on the $^{153}$Eu(n,$\gamma$) and $^{154}$Eu(n,$\gamma$) reactions cross sections. The assimilation of the integral trends is based on a linear least-square fitting procedure relying on the Bayes’ theorem.Realistic uncertainties are obtained thanks to a marginalization technique. The method is applied to the $^{153}$Eu and $^{154}$Eu capture cross sections recommended in the evaluated nuclear data library JEFF-3.1.1. Our results indicate an underestimation of both capture cross-sections of approximately 4.9% and 7.1%, respectively. For $^{153}$Eu, our study suggests to increase the capture resonance integral from I$_0$=1409 barns to I$_0$=1502 $\pm$ 68 barns. This trend is consistent with the capture resonance integral I$_0$=1560 barns obtained from recent time-of-flight measurements carried out at the Renssealer Polytechnic Institute. The trend obtained in this work on $^{154}$Eu capture cross section is also consistent with a recent analysis carried out on ENDF/B-VII.1, for which $^{154}$Eu capture cross section is very similar to JEFF-3.1.1, therefore confirming its underestimation

Experimental validation of photon-heating calculation for the Jules Horowitz Reactor

International audienc