Current status of the verification and processing system GALILEE-1 for evaluated data

International audienceThis paper describes the current status of GALILÉE-1 that is the new verification and processing system for evaluated data, developed at CEA. It consists of various components respectively dedicated to read/write the evaluated data whatever the format is, to diagnose inconsistencies in the evaluated data and to provide continuous-energy and multigroup data as well as probability tables for transport and depletion codes. All these components are written in C++ language and share the same objects. Cross-comparisons with other processing systems (NJOY, CALENDF or PREPRO) are systematically carried out at each step in order to fully master possible discrepancies. Some results of such comparisons are provided

Comments on jeff-33t1 library

International audienceComments on JEFF-33T1 Librar

Tripoli-4 benchmarking onjeff-3.3t1 library

International audienc

Use of probability tables for propagating uncertainties in neutronics

International audienceProbability tables are a generic tool that allows representing any random variable whose probability density function is known. In the field of nuclear reactor physics, this tool is currently used to represent the variation of cross-sections versus energy (neutron transport codes TRIPOLI4, MCNP, APOLLO2, APOLLO3, ECCO-ERANOS...). In the present article we show how we can propagate uncertainties, thanks to a probability table representation, through two simple physical problems an eigenvalue problem(neutron multiplication factor) and a depletion problem

Probability Tables: a generic tool for representing and propagating uncertainties

Probability tables are a generic tool that allows representing any random variable whose probability density function is known. In the field of Nuclear Reactor Physics, this tool is currently used to represent the variation of cross-sections versus energy (TRIPOLI4®, MCNP, APOLLO2, ECCO/ERANOS, …). In the present article we show how we can propagate uncertainties, thanks to a probability table representation, through two very simple mathematical problems: an eigenvalue problem (neutron multiplication factor, …) or a depletion problem

Unresolved resonance range processing and probability tables generation in the GAIA2 system

The objective of this paper is to present the work carried out at the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) to process nuclear data in the unresolved resonance range (URR). Recently, a great deal of effort has been devoted at IRSN to develop an independent nuclear data processing system, GAIA2. First, a nuclear data storing object, independent from the ENDF-6 format, has been implemented in order to transmit information between the components of a module-based scheme. Then, the generation of probability tables in the URR has been added as an independent module named TOP (as Table Of Probability), and tested on a selected set of benchmarks. The methods used and the results are discussed, and some limitations in the manner to construct the tables are pointed out

Unresolved resonance range processing and probability tables generation in the GAIA2 system

The objective of this paper is to present the work carried out at the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) to process nuclear data in the unresolved resonance range (URR). Recently, a great deal of effort has been devoted at IRSN to develop an independent nuclear data processing system, GAIA2. First, a nuclear data storing object, independent from the ENDF-6 format, has been implemented in order to transmit information between the components of a module-based scheme. Then, the generation of probability tables in the URR has been added as an independent module named TOP (as Table Of Probability), and tested on a selected set of benchmarks. The methods used and the results are discussed, and some limitations in the manner to construct the tables are pointed out

GALILEE-1 a validation and processing system for ENDF-6 and GND evaluations

International audienceGALILÉE-1 is the new validation and processing system for evaluated data, developed at CEA. This system can handle evaluations stored either in the ENDF-6 format or in the new General Nuclear Data (GND) format. It consists of various components respectively dedicated to read/write the evaluated data whatever the format is, to diagnose inconsistencies in the evaluated data and to provide continuous-energy and multigroup data as well as probability tables for transport and depletion codes. All these components are written in C++ language and share the same objects. This paper describes the state of progress of the various parts of the system and gives some illustrations

Methodology for an efficient statistical cross sections sampling in the unresolved resonance range

International audienceCross sections are crucial nuclear data that describe the probability for a particular nuclear reaction to occur between an incident particle and a target nuclide. In the context of neutron transport, accurate cross section calculations are notably crucial in that they provide the input for reactor physics and criticality safety calculations. Cross sections are computed from a large variety of models that sometimes involve experimental data, in particular to describe in detail the resonant shape of the cross sections at low energy. In this paper, the statistical contributions of compound nucleus resonances to the cross sections are investigated. This is particularly useful when artificial sets of statistical resonances must be sampled, eg. in the framework of the ladder method used to compute probability tables in the unresolved resonance range, where resonances are experimentally indistinguishable. The ladder method is a Monte Carlo based technique in which sets of resonances (called ladders) are sampled around reference energies in the unresolved resonance range from average resonance parameters. In this paper, a methodology is proposed to estimate the statistical weight of the resonances for cross sections calculations in the unresolved resonance range. This provides practical insights to determine the minimal number of resonances to be sampled in the unresolved resonance range, and the needed number of Monte Carlo histories. The methods of the present article can be extended to any physical problem based on a statistical sampling of nuclear resonances. In particular, the conclusions can be directly applied to nuclear data processing codes, to some evaluation techniques that require resonances sampling, and to Monte Carlo transport codes that handle the unresolved resonance range on-the-fly