Uncertainty and sensitivity analysis of PWR mini-core transients in the presence of nuclear data uncertainty using non-parametric tolerance limits

The impact of nuclear data uncertainties is studied for the reactor power and the reactivity during control rod withdrawal transients with reactivity insertions of 0.5 $and 0.97$ , respectively, for a PWR mini-core model. Multi-group cross sections, the multiplicities of both prompt and delayed neutrons, and fission spectra are varied by the application of the random sampling-based method XSUSA with covariance data of SCALE 6.1 supplemented by JENDL-4.0. The varied multi-group data are used by TRITON/NEWT to generate varied 2-group cross sections, which are then applied in neutron-kinetic/thermal-hydraulic calculations with DYN3D-ATHLET. A significant impact on both the reactivity uncertainty and the power uncertainty is observed. Since the distributional properties of the output time series vary across the problem time, the distribution-free Wilks tolerance limit is applied as a robust uncertainty measure to complex time series patterns. The most contributing nuclide reactions to the power uncertainty are identified via sensitivity analysis. (C) 2019 Elsevier Ltd. All rights reserved.

Preliminary analysis of the PETALE program and comparison to simulations

Thomas Jean-François Ligonnet

This Master thesis focus on a preliminary analysis of the experimental data obtained during the PETALE program in the CROCUS reactor at EPFL. The objective of PETALE is to validate neutron nuclear data for stainless steel and its main elements –namely iron, nickel, and chromium– and to contribute to the reduction of their uncertainties. For this purpose, transmission experiments were performed using activation dosimetry in separate reflectors made of the four materials. The manuscript is separated in three main parts. In the first part, after a presentation of the experimental setup, the measured spectra of the irradiated dosimeters are analysed. The results obtained with the methodology developed at EPFL are compared with those obtained by the collaboration partner CEA. The second part details two supplementary studies. First an experiment was performed and analysed to assess the impact of position uncertainties of core centre dosimeters. It was estimated to be 0.14 % for a position uncertainty of ±5 mm. The second study dealt with the estimation of PETALE’s monitors dead time. A dead time correction model was applied on a stable period experiment, and validated against activation dosimetry experiments. The power underestimation was estimated at 5 % at 80 W. The last part presents a comparison between experimental results and Serpent simulations of PETALE’s experiments using the JEFF-3.3 nuclear data library. In total twenty experiments were modelled and compared with their respective experiment. In this preliminary study, the results show a general agreement in the thermal range. An underestimation of the transparency of nickel and chromium for neutron with an energy of 2 MeV is observed. At higher energies, around 8 MeV, the chromium transparency is strongly overestimated. Finally, only a slight overestimation of the transparency to 2 MeV neutron is observed for iron and stainless steel reflectors. Outlooks are provided for the next steps of the analysis and validation, as well as for prospects on a longer term.

2021

Etudes numériques et expérimentales de caractéristiques d'un système rapide sous-critique alimenté par une source externe

Michaël Plaschy

One of the principal research axes related to nuclear energy concerns the management of nuclear wastes. In this context, certain innovative pathways are being explored to provide options which could lead to a major reduction of the quantities of radioactive nuclides requiring long-term geological storage. An important aspect is the development of advanced nuclear systems which allow the transmutation of long-lived minor actinides, the ADS (accelerator driven system) concept often being proposed for the purpose. Various important developments are necessary for these new systems, one of the principal needs being the qualification of calculational tools to enable the reliable design of an ADS demonstrator. In the present research, related numerical and experimental studies have been carried out in the framework of the MUSE4 programme, which was developed on the MASURCA facility at Cadarache, France. This programme consists of investigations of a reference critical configuration (MOX fuel with sodium), and several subcritical configurations driven by an external neutron source. The source is obtained with the help of a deuton accelerator, called GENEPI, which produces neutrons via a nuclear reaction, either with a tritium target (14.1MeV neutrons), or with a deuterium target (2.7MeV neutrons). The experimental techniques implemented in this research address the study of spectral variations via foil activation measurements near interfaces characteristic of ADSs. The additional value of such experimental results, compared to standard fission rate traverses obtained using fission chambers (which were also analysed), has been highlighted. An important novel set of integral parameters has thus been made available for the validation of different numerical codes and nuclear data. The activation and fission rate measurements carried out, as also experimental results for other parameters such askeff, βeff and φ*, have been interpreted using various calculational codes (mainly ERANOS and MCNP) and several different nuclear data libraries (JEF-2.2, ERALIB1, ENDF/B6v2). Supplementary information on the performance of the calculational methods has been obtained in the context of the MUSE4 benchmark exercise, conducted by the NEA. The present investigations have led to the resolution of various problems encountered during the earlier MUSE3 programme at MASURCA, viz. uncertainties related to description of the intrinsic source description, difficulties in interpreting fission traverses near the external source, etc. A detailed study has thus been made possible of the important factors which characterise subritical fast systems driven by an external source, i.e. influence of the subcriticality level and of the external source intensity. The various analyses carried out have permitted to establish several recommendations concerning the modeling of the external source and the accelerator tube, as well as the most appropriate numerical methods and data sets to be applied. In this context, the specificity of subcritical configurations, as regards the prediction of parameters such as spatial indices and absolute reaction rates, has been clearly brought out. Certain difficulties (interpretation of high threshold reactions, absolute flux determination, etc.) have thereby been highlighted, pointing at the need for further numerical and experimental investigations. Finally, some recent ERANOS developments have allowed a study to be carried out of the representativity of MUSE4 configurations with respect to alternative ADS concepts proposed in the framework of the international PDS-XADS project. In brief, the representativity of the MUSE4 configurations is found to be quite satisfactory for XADS-Na and XADS-gas. However, in the case of the XADS-Pb/Bi system, effects of the significant uncertainties associated with the nuclear data of lead and bismuth have been clearly highlighted, indicating the desirability of new, more specific experiments for XADS-Pb/Bi validation.

EPFL2004

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

Understanding the time-dependent behaviour of a nuclear reactor following an intended or unintended change of the reactor conditions is of crucial importance to the safe operation of nuclear reactors. Safety evaluations of nuclear reactors involve the analysis of normal and abnormal reactor operation states with adequate and validated simulation tools. There is a growing interest in supplementing the results of best-estimate reactor calculations with uncertainties. A major source of uncertainty in reactor calculations is the nuclear data. The propagation of uncertainties allows to study their impact on response values of reactor calculations and to establish safety margins in nuclear reactor safety evaluations. For the simulation of reactor transients, the two-step approach consisting of the generation of group constants in infinite lattice geometry and the full-core calculation using a neutron-kinetic/thermal-hydraulic code system has been state-of-the-art. Nowadays, Monte Carlo codes allow for the determination of group constants based on a three-dimensional full-core model. Thus, effects from neighbouring fuel assemblies can be considered. In this thesis, the applicability of the neutron-kinetic code DYN3D coupled with the thermal-hydraulic system code ATHLET for the simulation of transients in small reactor cores was assessed. Experimental reactivity accident tests carried out at the SPERT III research reactor were simulated. Appropriate group constants were generated with the Monte Carlo code Serpent. The overall good agreement between the calculated and the experimental results prove the applicability of the applied codes for the analysis of such reactor transients. Nuclear data uncertainties were propagated through the simulations of two control rod withdrawal transients in a mini-core model of a pressurised water reactor. The random sampling-based method XSUSA in combination with the SCALE code package was used to determine varied two-group constants. Based on these two-group constants, 1,000 DYN3D-ATHLET calculations were performed and statistically analysed to obtain uncertainty information for the reactor power and the reactivity. Since the assumption of normality could not be maintained for the reactor power, distribution-free Wilks tolerance limits were applied. By sensitivity analyses, the number of delayed neutrons per fission of U-235 and the scattering cross sections of U-238 were identified as the major contributors to the reactor power uncertainty. For a potential improvement of transient simulations, an approach was developed that was comprised of the determination of group constants using a Serpent full-core model at selected times during a transient simulation and the updating of the original infinite lattice group constants with these full-core group constants. Various challenges and possible solutions were identified. The applicability of the group constants obtained by a full-core model was assessed by their application in full-core diffusion calculations.

EPFL2021