Program - Brookhaven National Laboratory

Uncertainty Assessment for fast Reactors Based on Nuclear Data Adjustment
T. Ivanova, E. Ivanov, F. Ecrabet
Institut de Radioprotection et de Surete Nucleaire
For the purpose of safety assessment, IRSN studies issues of use of high-accurate codes and methods for fast
reactor simulation and uncertainty study. The unbiased Monte Carlo methods (MCNP code and SCALE
sequences) are used together with deterministic methods (ERANOS code) for computation of fast system
neutronics parameters. The cross-section adjustment technique has been recently implemented in the
IRSN’s in-house BERING validation tool in order to establish values for the robust biases and uncertainties
in these parameters. The adjustment technique is used for consolidation of a prior set of integral parameters
measured in benchmark experiments and a corresponding set of calculated parameters obtained with
mentioned computational codes. A better understanding of the performance of the adjustment methodology
is needed in order to improve the confidence concerning its validity and applicability for uncertainty
assessment. The benchmark exercise, established by the OECD/NEA Working Party on International
Nuclear Data Evaluation Co-operation (WPEC) Subgroup 33, offers a very good opportunity to learn the
drawbacks and advantages of the adjustment and to test BERING performance. The benchmark exercise
has been proposed with the aim to test different methods of nuclear data adjustment and different sets of
covariance data, so as to reduce the design uncertainties of a particular type of sodium-cooled fast reactor.
The exercise uses a single, limited set of integral experiments and measurements, which include Keff for
Jezebel, Jezebel- 240 Pu, Flattop- 240 Pu, ZPR6-7, ZPR6-7 with high 240 Pu content, ZPPR-9 and Joyo MK-1;
reaction rates for Jezebel, Flattop- 240 Pu, ZPR6-7 and ZPPR-9; and sodium void effects for ZPPR-9. The
final results are tested on a model of the Advanced Fast Burner Reactor (ABR) with plutonium oxide fuel
and a model of the Fast Breeder Reactor (FBR) core. Every participant to the benchmark is supposed
to use the same integral experiment values and uncertainties, but their own calculated values, sensitivity
coefficients, and adjustment method. Own or same initial cross sections and nuclear data covariances can
be used depending on step of the exercise. The full paper will present IRSN’s results of the benchmark
exercise generated using different sets of input data: integral parameters and sensitivity coefficients computed
by the deterministic ERANOS code and the Monte Carlo and deterministic SCALE6.0 sequences;
ENDF/B-VI.8 and ENDF/B-VII cross-section data; COMMARA 2.0 and TENDL-2011 covariances; with
and without integral correlations provided by JAEA. The outcomes of BERING will be analyzed and
compared in order to demonstrate whether the results of the adjustment converge when using different
input cross-section covariances. The impact of different calculation methods’ approximation bias will also
be shown.
OB 4 2:40 PM
New JEFF-3.2 Sodium Neutron Induced Cross-Sections Evaluation for Neutron Fast
Reactors Applications: from 0 to 20 MeV
P. Archier, G. Noguère, C. De Saint Jean
CEA, DEN, DER, SPRC, LEPh, Cadarache, F-13108 Saint-Paul-lez-Durance, France
A.J.M. Plompen, C. Rouki
JRC-IRMM, Retieseweg, 2440 Geel, Belgium
The current JEFF-3.1.1 sodium evaluation shows large discrepancies with the microscopic measurements
available on the EXFOR database. Furthermore, no cross-sections covariance matrices are available in this
file, which are of interest for fast reactor applications. In the framework of the ASTRID project (French
sodium fast reactor), a new sodium evaluation, from 0 to 20 MeV, have been carried out using the CONRAD
code. This file contains both re-evaluated nuclear data and covariances and is divided in two energy
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