Preliminary analysis of the PETALE program and comparison to simulations

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.

Future experimental programmes in the CROCUS reactor

Pavel Frajtag, Mathieu Hursin, Vincent Pierre Lamirand, Oskari Ville Pakari, Andreas Pautz

CROCUS is a teaching and research zero-power reactor operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology (EPFL). Three new experimental programmes are scheduled for the forthcoming years. The first programme consists in an experimental investigation of mechanical noise induced by fuel rods vibrations. An in-core device has been designed for allowing the displacement of up to 18 uranium metal fuel rods in the core periphery. The vibration amplitude will be 6 mm in the radial direction (±3 mm around the central position), while the frequency can be tuned between 0.1 and 5 Hz. The experiments will be used to validate computational dynamic tools currently under development, which are based on DORT-TD and CASMO/S3K code systems. The second programme concerns the measurement of in-core neutron noise for axial void profile reconstruction. Simulations performed at Chalmers University have shown how the void fraction and velocity profiles can be reconstructed from noise measurements. The motivation of these experiments is to develop an experimental setup to validate in-core the method in partnership with Chalmers University. The third experimental programme aims at continuing the validation effort on the nuclear data required in the calculation of GEN-III PWR reactors with heavy steel reflectors. This is a collaboration with CEA Cadarache that extends the results of the PERLE experiments carried out in the E reactor at CEA. Scattering cross sections at around 1 MeV will be studied separately by replacing successively the water reflector by sheets of stainless steel alloy and pure metals – iron, nickel, and chromium. Data will be extracted from the measured flux attenuation using foils in the metal reflector and from the criticality effects of these reflectors. In parallel to the three reactor experiments, we develop in-core detectors and measurement systems. Following the last development of a neutron noise measurement station in pulse mode, a second neutron noise station in current mode is being designed. In current mode the reactor can be used at higher power without dead time effects. It allows faster measurement time or lower results uncertainties. Finally, a joint development of a full new detection system based on chemical vapour deposited (sCVD) diamond has been started with the CIVIDEC instrumentation start-up company. A first prototype has been tested in November 2015 in CROCUS. One of the main purposes is to work on the discrimination of gammas, thermal and fast neutrons for demonstrating the interest of this detector type in a mixed neutron-gamma field.

2016

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

Development of the control assembly pattern and dynamic analysis of the generation IV large gas-cooled fast reactor (GFR)

Gaëtan Girardin

During the past ten years, different independent factors, such as the rapidly increasing worldwide demand in energy, societal concerns about greenhouse gas emissions, and the high and volatile prices for fossil fuels, have contributed to the renewed interest in nuclear technology. It is in this context that the Generation IV international forum (GIF) launched the initiative, in 2000, to collaborate on the research and development (R&D) efforts needed for the next generation, i.e. Generation IV, of nuclear reactors. These advanced systems will be ideally deployed beyond the year 2030, following the Generation III or III+ nuclear power plants, which are mainly based on light water technology and are currently entering deployment. A particular goal set for Generation IV systems is closure of the nuclear fuel cycle. Thus, apart from improvements in safety, they are expected to offer a better utilization of natural resources, as also a minimization of long-lived radioactive wastes. Among the systems selected by the GIF, the Gas-cooled Fast Reactor (GFR) is a highly innovative system with advanced fuel geometry and materials (fuel pellets of mixed uranium-plutonium carbide within a plate-type, honeycomb structure made of SiC). It is in the context of the large, 2400 MWth reference GFR design that the present doctoral research has been conducted, the principal aim having been to develop and qualify the control assembly (CA) pattern and corresponding CA implementation scheme for this system. The work has been carried out in three successive and complementary phases: (1) validation of the neutronics tools, (2) the CA pattern development and related static analysis, and (3) dynamic core behavior studies for hypothetical CA driven transients. The deterministic code system ERANOS and its associated nuclear data libraries for fast reactors were developed and validated for the previous generation of sodium-cooled reactors. The validation of ERANOS for GFR applications was, therefore, the first task to be realized in the present research. This has entailed a systematic reanalysis of the GFR-relevant, integral data generated at PSI during the GCFR-PROTEUS experimental program of the 1970's. Thus, during the first phase of the thesis, the reference PROTEUS test lattice from these experiments has been analyzed with ERANOS-2.0 and its associated, adjusted nuclear data library ERALIB1, in order to derive a reference computational scheme to be used later for the GFR analysis. Additionally, benchmark calculations were performed with the Monte Carlo code MCNPX, allowing one to both check the deterministic results and to analyze the sensitivity to different modern data libraries. It has been found that, for the main reaction rate ratios, the new analysis of the GCFR-PROTEUS reference lattice generally yields good agreement – within 1σ measurement uncertainty – with experimental values and with the Monte Carlo simulations. As shown by the analysis, the predictions were in somewhat better agreement in the case of the adjusted ERALIB1 library. The applicability of ERANOS-2.0/ERALIB1 as the reference neutronics tool for the GFR analysis could thus be demonstrated. Furthermore, neutronics aspects related to the novel features of the GFR, for which new experimental investigations are needed, were highlighted. In the second phase of the research, the CA pattern was developed for the GFR, based on iterative neutronics and thermal-hydraulics calculations, 2D and 3D neutronics models for the reactor core having first been set up using the reference ERANOS-2.0/ERALIB1 computational scheme. For the thermal-hydraulics analysis, the CEA code COPERNIC was used. This design work was followed by the study of an appropriate CA implementation scheme (number of CAs and corresponding positions within the core). Detailed neutronics studies revealed the existence of large CA interaction effects, so-called shadowing/anti-shadowing effects leading to an amplification/reduction of the CA worth. The interactions between the absorber pins within a CA, and between the CAs themselves, were investigated in detail, with the goal to optimize the CA efficiency, in terms of the absorber fraction and minimization of the associated heterogeneity effects. The proposed CA pattern consists of 54 absorber pins placed in a triangular lattice. Each absorber pin is a stainless-steel tube filled with highly enriched 10B boron carbide pellets. As a result of the detailed investigations, the absorber pin diameter could be chosen such as to minimize the pin-to-pin influence within the assembly. In particular, a central part of the CA was designed without any absorber pins (zone filled with stagnant helium). A final reduction of the heterogeneity effect (difference between homogeneous and heterogeneous treatments) to 13% was achieved through this feature. Of special importance, the neutronics investigations performed for the reference GFR core ("2004-Core"), especially those related to the CA interactions, have directly contributed to a new core design ("2007-Core"), with the height-to-diameter ratio having been increased to 0.6, compared to 0.3 for the reference core. During the third phase, detailed coupled, 3D neutron-kinetics (NK) and 1D thermal-hydraulics (TH) models were developed for the GFR core, the aim being to arrive at an in-depth understanding of the 3D core behavior during CA driven transients, especially from the viewpoint of spatial effects. The coupled models were developed using the PARCS code for the 3D NK and the TRACE code for the 1D TH modeling. Particular attention was paid to have each individual fuel sub-assembly and CA represented, in order to allow the analysis of local deformations of the 3D distributions of power and safety related parameters, such as the coolant, cladding and fuel temperatures. The validation of the coupled full-core models was performed against reference ERANOS-VARIANT calculations. In particular, the CA worth and reactivity feedbacks were benchmarked, the discrepancies being shown to be relatively low (

EPFL2009