Fission Spectrum Related Uncertainties

The paper presents a preliminary uncertainty analysis related to potential uncertainties on the fission spectrum data. Consistent results are shown for a reference fast reactor design configuration and for experimental thermal configurations. However the results obtained indicate the need for further analysis, in particular in terms of fission spectrum uncertainty data assessment

Copper benchmark experiment for the testing of JEFF-3.2 nuclear data for fusion applications

A neutronics benchmark experiment on a pure Copper block (dimensions 60 × 70 × 70 cm3) aimed at testing and validating the recent nuclear data libraries for fusion applications was performed in the frame of the European Fusion Program at the 14 MeV ENEA Frascati Neutron Generator (FNG). Reaction rates, neutron flux spectra and doses were measured using different experimental techniques (e.g. activation foils techniques, NE213 scintillator and thermoluminescent detectors). This paper first summarizes the analyses of the experiment carried-out using the MCNP5 Monte Carlo code and the European JEFF-3.2 library. Large discrepancies between calculation (C) and experiment (E) were found for the reaction rates both in the high and low neutron energy range. The analysis was complemented by sensitivity/uncertainty analyses (S/U) using the deterministic and Monte Carlo SUSD3D and MCSEN codes, respectively. The S/U analyses enabled to identify the cross sections and energy ranges which are mostly affecting the calculated responses. The largest discrepancy among the C/E values was observed for the thermal (capture) reactions indicating severe deficiencies in the 63,65Cu capture and elastic cross sections at lower rather than at high energy. Deterministic and MC codes produced similar results. The 14 MeV copper experiment and its analysis thus calls for a revision of the JEFF-3.2 copper cross section and covariance data evaluation. A new analysis of the experiment was performed with the MCNP5 code using the revised JEFF-3.3-T2 library released by NEA and a new, not yet distributed, revised JEFF-3.2 Cu evaluation produced by KIT. A noticeable improvement of the C/E results was obtained with both new libraries

Benchmarking comparison and validation of MCNP photon interaction data

The objective of the research was to test available photoatomic data libraries for fusion relevant applications, comparing against experimental and computational neutronics benchmarks. Photon flux and heating was compared using the photon interaction data libraries (mcplib 04p, 05t, 84p and 12p). Suitable benchmark experiments (iron and water) were selected from the SINBAD database and analysed to compare experimental values with MCNP calculations using mcplib 04p, 84p and 12p. In both the computational and experimental comparisons, the majority of results with the 04p, 84p and 12p photon data libraries were within 1σ of the mean MCNP statistical uncertainty. Larger differences were observed when comparing computational results with the 05t test photon library. The Doppler broadening sampling bug in MCNP-5 is shown to be corrected for fusion relevant problems through use of the 84p photon data library. The recommended libraries for fusion neutronics are 84p (or 04p) with MCNP6 and 84p if using MCNP-5

Recommendations for MYRRHA relevant cross section data to the JEFF project

Within the framework of Work Package 10 of the EC FP7 CHANDA project, nuclear data of importance for the operation of MYRRHA, a lead-bismuth cooled accelerator driven reactor under development at SCK•CEN (BE), were studied. Based on data in the main nuclear data libraries, i.e. JEFF, JENDL, ENDF/B and BROND, and in the TENDL and CIELO libraries and on experimental data reported in the literature, recommendations to the JEFF project were made for several nuclides of interest to the MYRRHA reactor.JRC.G.2-Standards for Nuclear Safety, Security and Safeguard

OECD/NEA Data Bank Scientific and Intergral Experiments Databases in Support of Knowledge Preservation and Transfer

The OECD/Nuclear Energy Data Bank was established by its member countries as an institution to allow effective sharing of knowledge and its basic underlying information and data in key areas of nuclear science and technology. The activities as regards preserving and transferring knowledge consist of the: — Acquisition of basic nuclear data, computer codes and experimental system data needed over a wide range of nuclear and radiation applications. — Independent verification and validation of these data using quality assurance methods, adding value through international benchmark exercises, workshops and meetings and by issuing relevant reports with conclusions and recommendations, as well as by organising training courses to ensure their qualified and competent use. — Dissemination of the different products to authorised establishments in member countries and collecting and integrating user feedback. Of particular importance has been the establishment of basic and integral experiments databases and the methodology developed with the aim of knowledge preservation and transfer. Databases established thus far include: — IRPhE – International Reactor Physics Experimental Benchmarks Evaluations, — SINBAD – a radiation shielding experiments database (nuclear reactors, fusion neutronics and accelerators), — IFPE – International Fuel Performance Benchmark Experiments Database, — TDB – The Thermochemical Database Project, — ICSBE – International Nuclear Criticality Safety Benchmark Evaluations [1], — CCVM – CSNI Code Validation Matrix of Thermal-hydraulic Codes for LWR LOCA and Transients [2]. This paper will concentrate on knowledge preservation and transfer concepts and methods related to some of the integral experiments and TDB

NUCLEAR DATA NEEDS FOR ADVANCED REACTOR SYSTEMS. A NEA NUCLEAR SCIENCE COMMITTEE INITIATIVE.

The Working Party on Evaluation Cooperation (WPEC) of the OECD Nuclear Energy Agency Nuclear Science Committee has established an International Subgroup to perform an activity in order to develop a systematic approach to define data needs for Gen-IV and, in general, for advanced reactor systems. A methodology, based on sensitivity analysis has been agreed and representative core configurations for Sodium, Gas and Lead cooled Fast Reactors (SFR, GFR, LFR) have been defined as well as a high burn-up VHTR and a high burn-up PWR. In the case of SFRs, both a TRU burner (called in fact SFR) and a core configuration with homogeneous recycling of not separated TRU (called EFR) have been considered