Nuclear data generation & implementation for analog Monte Carlo simulation

Nuclear data is in constant evolution as more experimental data is gathered, computational capabilities increase, and evaluators verify its validity by means of stochastic and deterministic simulations. The focus here is on the analog Monte Carlo simulation of nuclear reactions that produce more than two particles in the outgoing channel, which needs specific considerations to ensure the correlations between the particles and thus the conservation of energy and of translational and angular momenta. It is possible to adapt nuclear data and its exploitation to implement realistic reactions from the phenomenological point of view (as opposed to the historical need of variance reduction techniques), which increases computation time but allows the expansion of the transport codes capabilities. Simulation anomalies were found concerning the kinematical calculations of photon energies emitted from neutron-induced inelastic scattering (n,n’γ), as well as concerning the photon multiplicity of 155Gd(n,γ) due to the presence of a rotational band in 156Gd. Recommendations are given for potential solutions for both anomalies

Improvement of Geant4 Neutron-HP package: From methodology to evaluated nuclear data library

International audienceAn accurate description of interactions between thermal neutrons (below 4 eV) and materials is key to simulate the transport of neutrons in a wide range of applications such as criticality-safety, reactor physics, compact accelerator-driven neutron sources, radiological shielding or nuclear instrumentation, just to name a few. While the Monte Carlo transport code Geant4 was initially developed to simulate particle physics experiments, its use has spread to neutronics applications, requiring evaluated cross-sections for neutrons and gammas between 0 and 20 MeV (the so-called neutron High Precision – HP – package), as well as a proper offline or on-the-flight treatment of these cross-sections. In this paper we will point out limitations affecting the Geant4 (version 10.07.p01) thermal neutron treatment and associated nuclear data libraries, by using comparisons with the reference Monte Carlo neutron transport code Tripoli-4®, version 11, and we will present the results of various modifications of the Geant4 Neutron-HP package, required to overcome these limitations. Also, in order to broaden the support of nuclear data libraries compatible with Geant4, a nuclear processing tool has been developed (the code is available on a GitLab repository) and validated allowing the use of the code together with ENDF/B-VIII.0 and JEFF-3.3 libraries for example. These changes will be taken into account in an upcoming Geant4 release

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

This 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

Current status of the verification and processing system GALILÉE-1 for Evaluated Data

This paper describes the current status of GALILÉE-1 [1], 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 independently of the format, to diagnose inconsistencies in the evaluated data and to provide continuous-energy and multigroup data as well as probability tables for particle transport and depletion codes. All these components are written in C++ language and share the same objects. In this paper, we detail themain advances made in GALILÉE-1 : Unresolved Resonance Range (URR) treatment, Thermal Scattering Law (TSL) processing and anisotropy calculations from resonance parameters

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

GALILÉE-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

FIFRELIN – TRIPOLI-4® coupling for Monte Carlo simulations with a fission model. Application to shielding calculations

TRIPOLI-4® Monte Carlo transport code and FIFRELIN fission model have been coupled by means of external files so that neutron transport can take into account fission distributions (multiplicities and spectra) that are not averaged, as is the case when using evaluated nuclear data libraries. Spectral effects on responses in shielding configurations with fission sampling are then expected. In the present paper, the principle of this coupling is detailed and a comparison between TRIPOLI-4® fission distributions at the emission of fission neutrons is presented when using JEFF-3.1.1 evaluated data or FIFRELIN data generated either through a n/g-uncoupled mode or through a n/g-coupled mode. Finally, an application to a modified version of the ASPIS benchmark is performed and the impact of using FIFRELIN data on neutron transport is analyzed. Differences noticed on average reaction rates on the surfaces closest to the fission source are mainly due to the average prompt fission spectrum. Moreover, when working with the same average spectrum, a complementary analysis based on non-average reaction rates still shows significant differences that point out the real impact of using a fission model in neutron transport simulations