Alain Santamarina

Trends in Nuclear Data Derived from Integral Experiments in Thermal and Epithermal Reactors

The calculation of reactor design parameters with ever higher accuracy requires constant improvement in basic nuclear data and computational techniques. Integral experiments are a very powerful means of assessing the quality of nuclear data for specific applications, nicely complementing differential measurements. In France, an extensive programme of validation of the neutronic code APOLLO2 and its associated JEF2.2 based library CEA93 is underway. The results obtained allows us to demonstrate the interest of coupling information extracted from integral experiments and differential measurements to improve nuclear data evaluations. This study has strongly contributed to the improvement and selection of the nuclear data evaluations which will be introduced in the new JEFF3.0 library.

Priority Nuclear Data Needs for Industrial Applications

The aim of this paper is to review the activity of the French Committee on Nuclear Data (CFDN) regarding priority of data needs for R&D programs and industrial applications in France under investigation at FRAMATOME, EDF, COGEMA, and CEA. The target precisions on key integral parameters and the associated milestones are surveyed and a long-term strategy to evaluate quantitatively the corresponding data needs, including the trends for data changes resulting from various validation studies using JEF-2 data is presented.

A Modeling of BWR-MOX Assemblies Based on the Characteristics Method Combined with Advanced Self-Shielding Models

Abstract Calculations based on the characteristics method and different self-shielding models are presented for 9 × 9 boiling water reactor (BWR) assemblies fully loaded with mixed-oxide (MOX) fuel. The geometry of these assemblies was recovered from the BASALA experimental program. We have focused our study on three configurations simulating the different voiding conditions that an assembly can undergo in a BWR pressure vessel. A parametric study was carried out with respect to the spatial discretization, the tracking parameters, and the anisotropy order. Comparisons with Monte Carlo calculations in terms of keff, radiative capture, and fission rates were performed to validate the computational tools. The results are in good agreement between the stochastic and deterministic approaches. The mutual self-shielding model recently introduced within the framework of the Ribon extending self-shielding method appears to be useful for this type of assemblies. Indeed, in the calculation of these MOX benchmarks, the overlapping of resonances, especially between 238U and 240Pu, plays an important role due to the spectral strengthening of the flux as the voiding percentage is increased. The method of characteristics is shown to be adequate to perform accurate calculations handling a fine spatial discretization.

First burnup credit application including actinides and fission products for transport and storage cask by using French experiments

The burnup credit (BUC) methodology for a transport and storage cask application, including actinides and fission products, is implemented at AREVA TN using the French BUC calculation route for pressurized water reactor (PWR) UO2 used fuel. The methodology is based on the connection of the French depletion code DARWIN2 and the French criticality safety package CRISTAL V1. The BUC methodology includes the experimental validation of the computation codes dedicated to the calculation of the used fuel inventory calculations. Indeed, the results of the comparison calculation–experiment (C−E)/E allow to determine either a set of isotopic correction factors (ICFs) for the BUC nuclides considered in the criticality calculation or keff-penalty terms directly used for the definition of the keff-acceptance criterion for the criticality assessment of the transport and storage cask. These ICFs or keff-penalty terms are one of the key of the BUC method to guarantee the conservativeness of the fuel reactivity in safety-criticality calculations using BUC approach. A French BUC program has been developed at CEA/Cadarache in the framework of the CEA–AREVA collaboration in order to validate fuel inventory calculations. This program involves two kinds of experiments: chemical analyses and microprobe measurements of PWR irradiated fuel pins (French PIE program) on one hand, and reactivity worth measurements of the BUC nuclides in the MINERVE reactor on the other hand. This paper highlights, through a first industrial AREVA TNs application of the BUC method, including fission products, that the French PIE program and reactivity worth measurements in MINERVE reactor are suitable for the implementation of BUC in transport and storage cask applications loaded with PWR UO2 used fuels assemblies.