Student Research Experience Openings

GENTLE Student Research Experiences (SRE) applications will be accepted until August 31st, 2016. SREs will have to be completed by December 31st, 2016, with no exceptions.

 

EC Joint Research Centre Geel

(Geel, Belgium)

Evaluation of neutron cross section data for MYRRHA

SCK•CEN, the Belgian Nuclear Research Centre in Mol has been working for several years on the design of the multi-purpose irradiation facility MYRRHA. This flexible fast spectrum research reactor (50-100 MWth) is conceived as an accelerator driven system (ADS), able to operate in sub-critical and critical modes.

For the development and safety assessment of MYRHHA accurate nuclear data are required. Therefore, the quality of the main nuclear data libraries, e.g. ENDF/B, JEFF and JENDL, has to be assessed with an emphasis on nuclear data for MYRRHA-relevant elements and isotopes. The assessment includes a verification of the cross sections based on experimental data that are available in the literature and recommendations for improvements will be given as input for an updated version of the JEFF library. This library will be used as a reference library for the licensing of MYRRHA.

Keywords: MYRRHA, neutron, cross section, nuclear data, evaluation, verification, licensing, safety

Willy Mondelears, JRC Geel, willy.mondelaers@gentleproject.eu

Peter Schillebeeckx, JRC Geel, peter.schillebeeckx@gentleproject.eu

Jan Heyse, JRC Geel, jan.heyse@gentleproject.eu

 

Integral data

An accurate and complete knowledge of nuclear data for reactor dosimetry is essential for improving assessments of the service life of reactor pressure vessels and other components in nuclear power plants. Recent dosimetry evaluations have revealed that the uncertainty on spectrum averaged (SPA) cross sections, used to benchmark evaluated files, is often very large. In addition, no evaluated experimental SPA data are available for a number of reactions.

The Standards for Nuclear Safety, Security and Safeguards (SN3S) unit of the JRC Geel is involved in the improvement of neutron cross-section data for reactor dosimetry and measurement of SPA cross sections in a 252Cf(s.f.) neutron field as requested by the IAEA/NDS and the OECD/NEA. A selected number of neutron induced reactions will be measured at the NPL’s neutron metrology department and at CIEMAT’s newly commissioned neutron metrology lab. This project aims at developing and commissioning the experimental set-up and defining the measurement strategy.

Keywords: reactor, dosimetry, neutron, cross section, SPA, 252Cf, measurement, safety

Willy Mondelears, JRC Geel, willy.mondelaers@gentleproject.eu

Peter Schillebeeckx, JRC Geel, peter.schillebeeckx@gentleproject.eu

Jan Heyse, JRC Geel, jan.heyse@gentleproject.eu

 

 

EC Joint Research Centre Karlsruhe

(Karlsruhe, Germany)

Design, Modelling and Synthesis of Accident Tolerant Fuel Pellets

After the events in the Nuclear Power plant of Fukushima Daiichi in 2011, research efforts have been started worldwide in order to develop innovative accident tolerant fuels for existing light water reactors. These fuels would fit into the existing core design of a LWR but would provide more robustness during accidents with extended loss of core cooling. The main safety benefits are an increased safety margin to allowable temperatures and more available reaction time before it comes to core meltdown. Current proposals include advanced fuel claddings but also innovative fuel pellets. Most approaches for advanced claddings are based on substituting the zircaloy cladding of the fuel rods by ceramic materials (e.g. SiC) or stainless steel in order to increase robustness with respect to high temperatures but also to reduce the potential for hydrogen generation.
Advanced fuel pellets can be based on materials with superior properties compared to uranium dioxide but also on particle fuel embedded in an inert matrix. Important matrix properties are high heat conductivity, high melting point, good mechanical stability, radiation resistance and high retention of fission products inside the fuel matrix. The Institute for Transuranium Elements (JRC Karlsruhe) has developed an advanced method to produce small beads of fuel with designed properties. This method constitutes an ideal basis to synthesize accident tolerant fuel pellets based on UO2 but also MOX fuel.
The candidate will work in the Nuclear Fuels Unit of JRC Karlsruhe and will participate in the design of advanced accident tolerant fuel based on a silicon carbide matrix with embedded fuel particles of uranium and plutonium dioxide. The irradiation behaviour will be modelled and finally a number of fuel pellets will be synthesized and characterized using advanced methods such as ceramography, SEM, XRD, radiography and others. The produced fuel pellets are an ideal basis for future irradiation experiments.
The work plan includes the production of uniform beads of uranium and/or MOX and/or plutonium dioxide (the Pu should be a good differentiator for JRC Karlsruhe compared to US) (~ 100 µm to ~ 500 µm) and to embed them directly into a SiC matrix. Swelling is to be compensated by a defined bead porosity. Therefor a high density of the beads has not to be achieved in order to accommodate for fission gas build-up and solid swelling. By this way the SiC matrix should stay without cracks during irradiation. The advantages would be a lower fuel temperature (SiC has better heat conductivity as UO2) and an additional fission product barrier (SiC matrix).
The detailed steps of the work plan are: synthesis and characterisation of MOX (UO2) particles in the required size range and their embedding in a SiC matrix followed by investigations on properties such as thermal conductivity and examinations on thermal stability behaviour.

Keywords: MOX fuels, sol gel synthesis, SiC matrix.

Contact Person / Supervisor:

Daniel FREIS, daniel.freis@gentleproject.eu

Proposed starting date: in 2014.

Proposed duration: 12 months.

 

 

SCK•CEN

Knudsen cell effusion measurements of tellurium release from liquid lead-bismuth eutectic

The Belgian Nuclear Centre (SCK•CEN ) is currently developing the MYRRHA irradiation facility. MYRRHA is an accelerator driven nuclear system that is able to operate in subcritical mode and will be used to study waste transmutation. The coolant that will be used is liquid lead bismuth eutectic (LBE), which has excellent neutronic, heat transfer, and radiation shielding properties. A drawback of using LBE is that polonium is formed by activation of bismuth. Under normal operating conditions, polonium is very well retained in LBE, but evidence has been found that under off-normal conditions polonium may evaporate significantly from LBE which jeopardizes the safety of the installation. The physicochemical processes that underlie this enhanced evaporation are currently not well understood. In support of the design of MYRRHA and to establish the licensing case, a better understanding of this enhanced evaporation is required.
The experimental study of polonium is very challenging because of its high radioactivity and low availability. In the present project, we will circumvent these limitations by studying the lighter chemical homologue of polonium, tellurium. By measuring the release of tellurium from LBE, we expect to gain valuable insight into the behaviour of polonium under similar conditions. Tellurium release will be measured by Knudsen effusion mass spectrometry (KEMS). The temperature and concentration dependence of tellurium release will be determined in inert, oxidizing and humid atmospheres. Significant effort will be devoted to the analysis of mass spectra in order to identify the chemical speciation of the evaporated molecules. The majority of the work will be performed at JRC Karlsruhe in the frame of a collaboration.

Keywords: MYRRHA, lead bismuth eutectic, polonium, tellurium, Knudsen effusion mass spectrometry

Alexander Aerts, SCK-CEN, alexander.aerts@sckcen.be

 

 

EC Joint Research Centre Geel

(Geel, Belgium)

Improve the accuracy of capture cross section measurements

High resolution cross section measurements are performed at the neutron time-of-flight facility GELINA, which is operated by the Standards for Nuclear Safety, Security and Safeguards (SN3S) unit of the JRC Geel. These measurements are performed to improve the nuclear data for neutron induced reactions, which are important in various fields of nuclear technology.

A continuous effort is made to enhance the accuracy of the experimental data by improving the experimental techniques and data analysis procedures.

This project aims at a better understanding of the background contribution to the observed response in order to reduce bias effects. More in particular measurements will be performed to assess the contribution due to neutron scattering in the sample and disentangle the components due to room return and direct scattering into the detector. The project involves cross section measurements at GELINA and the use of data reduction and analysis codes, including a nuclear reaction model code.

Keywords: neutron, cross section, time-of-flight, GELINA, accuracy, background, scattering, measurement, analysis

Willy Mondelears, JRC Geel, willy.mondelaers@gentleproject.eu

Peter Schillebeeckx, JRC Geel, peter.schillebeeckx@gentleproject.eu

Jan Heyse, JRC Geel, jan.heyse@gentleproject.eu

Development of an active background shield for neutron detection

The project focusses on the development of an intelligent shield to reduce background contributions for low-rate neutron counting applications. The shield consists of plastic scintillator acting as a veto detector. When the intelligent shield detects muons produced in the atmosphere or neutrons from an external source, a veto signal will temporarily interrupt the primary neutron detection system thus reducing the neutron background count and consequently improve the detection limit. It is expected that the veto shield will reduce the background by about one order of magnitude and therefore increases the sensitivity for various experiments with low neutron yields such as low-level nuclear waste measurements or for more fundamental experiments to study exotic nuclei or low-yield nuclear reactions.

The purpose is to investigate if atmospheric muons, or an external source of neutrons, can be identified by their interactions (ionization and elastic scattering, respectively) in the plastic scintillation detectors. We expect that tracking the time of energy depositions, and the signal pulse height, in multiple scintillation layers can identify when a muon event, or an external neutron source, can be expected to impinge on the neutron detector assembly. A positive identification will cause a veto signal to temporarily inhibit the acquisition in the detector assembly, and consequently reduce the measured neutron background.

The main objective of the project is to define an optimum active shield. i.e. number, size and geometry of the plastic scintillators. This will be realized by Monte Carlo simulations which are validated by experimental results. The experiments include measurements at the VdG facility of the EC-JRC Geel using mono-energetic neutron beams and measurements with radionuclide neutron sources at the JRC Geel and JRC Ispra.

https://ec.europa.eu/jrc/en/page-related-facilities/all/272/28131

Characterization of melted fuel by neutron resonance analysis

JRC in collaboration with the Japanese Atomic Energy Agency (JAEA) have developed a method to characterize particle-like debris of melted nuclear fuel from the damaged Fukushima Daiichi reactor cores. The method referred to a Neutron Resonance Densitometry is based on the presence of resonance structures in neutron induced reaction cross sections. NRD relies on the principles of Neutron Resonance Transmission Analysis (NRTA) and Neutron Resonance Capture Analysis (NRCA), which are two methods that have been developed at the JRC’s facility GELINA (GEel LINear Accelerator). The project focusses on measurements at GELINA to validate the method by results of experiments at GELINA using samples of nuclear material.

https://ec.europa.eu/jrc/en/training-programmes
https://ec.europa.eu/jrc/en/page-related-facilities/all/272/28131
https://ec.europa.eu/jrc/en/news/japan-and-jrc-demonstrate-method-quantify-u-and-pu-melted-fuel

EC Joint Research Centre Karlsruhe

(Karlsruhe, Germany)

Karlsruhe Experimental studies of Corium material systems: high-temperature behaviour of two -three crucial compositions in the U-Pu-Zr-Fe-O system (3-5 months, D. Manara)

During hypothetical meltdown accidents in nuclear reactors, the fuel is foreseen to liquefy and react with the liquid cladding and the vessel materials, forming a “lava-like” melt commonly called “corium”. Knowledge of the corium behaviour during a severe accident has great importance for the risk and safety assessment of a nuclear reactor.

In this framework, the present study will focus on the applicability of Raman spectroscopy to the study of some crucial corium sub-systems, previously laser-heated beyond melting. The key in-vessel system includes UO2 fuel, zirconium-based alloys for the cladding and ferrous-based reactor internal structures, in various possible oxidising conditions. Once the vessel fails, and the corium interacts with the ex-vessel structures, then more complex systems will be formed, containing also concrete decomposition products: oxides of Si, Ca, Al, Na, K, Mg, etc.

Raman spectroscopy is a particularly advantageous technique thanks to its remote, non-destructive and local character. Such features, combined with a minimal sample preparation, are particularly attractive when deal with radioactive materials. A sensitivity assessment of this vibrational spectroscopy approach to the identification of different corium phases will constitute the main goal of the present internship.

Keywords: Corium, Meltdown accident, Laser heating

Contact Person / Supervisor:

Dario Manara, JRC Karlsruhe, Dario.manara@gentleproject.eu

 

Investigation of the structural and electronic properties of U1-yAmyO2±x

The Nuclear Fuels Unit of JRC Karlsruhe conducts research programmes related to the safety of minor actinides bearing nuclear fuels, such as U1-yAmyO2±x. As these materials are considered as promising compounds for the efficient recycling of Am, this internship project aims at extending our knowledge of their structural and electronic properties. The oxidation states of U and Am will be of particular interest as controversial results have been published these past few years.UO2, UO2+x, AmO2, AmO2-x, U1-yAmyO2±x will be synthesized by sol-gel before the starting date of the internship. The candidate will however participate to the pressing and the sintering of the materials. Several sintering atmosphere will be employed so as to reach different O stoichiometries and cation valence states. The candidate will actively participate to the structural characterization with X-Ray Diffraction (XRD) and to the electronic characterization through X-ray Absorption Near Edge Spectroscopy (XANES) and X-ray Photoelectron Spectroscopy (XPS). XRD will provide structural informations on the synthesized materials (lattice parameter, crytallinity, etc.). XPS and XANES at M edges will be performed on UO2, UO2+x, AmO2, AmO2-x, U1-yAmyO2±x to study the oxidation states of U and Am cations. A beamtime has already been allocated for the XANES in the framework of the TALISMAN project but no precise date has been fixed yet. If possible, the candidate will participate to this experiment. In concrete terms, the candidate will learn one of the synthesis processes of nuclear materials. Therefore, he will also learn the security rules relative to the work in a nuclear facility. As the main focus of the internship is the characterization by XRD, XPS and XANES, the applicant will participate, as much as possible, to the sample preparation and to the experiments themselves. The candidate will learn the basis of the analysis and interpretation of XRD, XPS and XANES data.

Keywords: Actinide dioxides, Americium burning, X-ray Absorption Spectroscopy.

Contact Person / Supervisor:

Damien PRIEUR, damien.prieur@gentleproject.eu

Proposed starting date: April 2014.

Proposed duration: 6 months.

 

New sol-gel route for nuclear carbide preparation and fuel pellet fabrication by SPS: pre-tests using cerium and neodymium

The project is aimed to the usage of metal carbides as nuclear fuels. Alternative routes for carbide fuel preparation based on low temperature thermal decomposition of suitable reactive metal precursors shall be investigated. Sol gel methods for their preparation will be applied based on already well established procedures. The obtained carbides will be used for reparation of final pellets in a spark plasma sintering device. The first work is focused on the cold pre-tests using Ce and Nd salts in order to map the possible feasibility of such method for UC and PuC preparation.

The work plan includes the synthesis, characterisation of gel consisting of organic acid and metal ions as precursors for the metal carbides. The application of the sol gel methods ensures a constant and fixed carbon to metal ratio at molecular scale. Thermal decomposition at relatively low temperatures and under inert conditions then leads via decomposition to the formation of nano-scaled products metal carbides. Preparation of the final pellets will take place spark plasma sintering device.

In detail the work plan foresees the steps of synthesis of the precursors by sol gel methods, their spectroscopic characterisation, the thermal decomposition step, analyses of the resulting ceramics, further and final handling of the ceramics due to the production of pellets and all this based on variations in the substrates involved in the synthesis and different synthesis conditions.

Keywords: Actinide carbides, thermal decomposition, alternative fuels

Contact Person / Supervisor: Vaclav TYRPEKL, vaclav.tyrpekl@gentleproject.eu

Proposed starting date: April 2014.

Proposed duration: 6 months.

 

Ligands for decommissioning

The project is aimed to the usage of ligands for decommissioning. Once when an apparatus or a surface is contaminated by radioactive nuclides which are mainly heavy metals the contamination in principal can be removed from the surface by coordination of the metal ions to suitable ligands under formation of stable and soluble complexes. These complexes can be dissolved in an appropriate solvent and the solution be removed from the surface leaving it back as clean.

According to the project scopes synthesis, spectroscopic and structural characterisation as well as determination of complex formation constants are objectives of the project. Complexes of lanthanides may act as models for complexes from fission products on the one hand, on the other hand they represent an attractive simulate for the corresponding actinide compounds before they are to be synthesised as well.

Expected side effects of this proposal are knowledge on the influence of the ligand constitution on the coordination geometry of the complexes formed including possible other application of these ligands such as in partitioning or medicinal applications.

The project is specially designed for students who want to perform their master thesis in this field.

Keywords: Actinide complexes, lanthanide complexes, synthesis, spectroscopy, complex formation constant

Contact Person / Supervisor: Olaf WALTER, olaf.walter@gentleproject.eu

Proposed starting date: April 2014.

Proposed duration: 4-6 months.

 

Karlsruhe Institute of Technology – INR

(Karlsruhe, Germany)

Validation of the ASTEC V2.1 code with the KIT CORA-13 experiment

The Fukushima accidents in 2011 have shown that further improvements on Severe Accident Management (SAM) are still necessary. Among the different strategies to stop the core degradation process, one of the most important is the injection of water into the overheated core (e.g. core reflooding). However, several experimental studies (e.g. the QUENCH and CORA test facilities) have shown that before the water manages to cool down the overheated core, the oxidation process of the Zircaloy-4 claddings can be intensified. As a result, the measure may backfire in terms of enhanced fission product release and enhanced hydrogen generation, this fact jeopardizing the containment at an early phase of the accident.

The availability of state-of-the art validated numerical simulation tools is relevant to estimate the current knowledge on severe accidents, and to assess if further efforts on R&D are necessary to gain a deeper understanding of certain phenomena. The reference European code ASTEC, jointly developed by the “Institut de Radioprotection et de Sûreté Nucléaire (IRSN)” and the “Gesellschaft für Anlagen und Reaktorsicherheit (GRS)” is able to simulate the progression of a severe accident from the initiating event till the release of radioactive to the environment.

Nevertheless, the ASTEC code requires further validation against experimental facilities handling key physical phenomena in a severe accident sequence, and further application to nuclear power plants. Within the European project CESAM (2013-2017), embedded within the framework program FP7, ASTEC capabilities concerning SAM will be improved both via validation against reactor-scalable experiments and plant calculations on European NPPs.

The integral CORA facility at the Forschungszentrum Karlsruhe (FZK), now Karlsruhe Institute of Technology (KIT), is an optimal calibration data for code validation, since it investigated the different phenomena contributing to Light Water Reactor (LWR) fuel damage destruction in severe accident situations (i.e. bundle temperatures ranging between 1200 °C up to 2400 °C).

In the frame of the present work, the ASTEC V2.1 code will be validated against the CORA-13 experiment (OECD International Standard Problem 31), this test studying the influence of water injection into an overheated core, in terms of hydrogen generation and core damage propagation. As a first step, the basics of the CORA facility and the ASTEC code will be acquainted. As a second step, an input deck for the CORA-13 experiment will be developed in ASTECV2.1. As a third step, the ASTECV2.1 model of CORA-13 will be validated against experimental data. Finally, sensitivity calculations will be performed and an outlook of these investigations will be given.

Pre-requisites: Mechanical engineering, Reactor Physics, programming skills
Welcoming skills: Linux OS
Duration: 4-6 Months
Location: KIT Campus North, Institute for Neutron Physics and Reactor Technology (INR), Building 521

Contact persons:

  • Ignacio Gómez García-Toraño (ignacio.torano@kit.edu,+49 (0)721 608-29281)
  • Dr. Victor Sanchez (victor.sanchez@kit.edu, +49 (0)721 608-22283)

 

Validation of the ASTEC code with the KIT experiment QUENCH-06, and assessment of modelling uncertainties using the URANIE software

The Fukushima accidents in 2011 have shown that further improvements on Severe Accident Management (SAM) are still necessary. Among the different strategies to stop the core degradation process, one of the most important is the injection of water into the overheated core (e.g. core reflooding). However, several experimental studies (e.g. the KIT-QUENCH and CORA facilities) have shown that before the water manages to cool down the overheated core, the oxidation process of the Zircaloy-4 claddings can be intensified. As a result, the measure may backfire in terms of enhanced fission product release and enhanced hydrogen generation, this fact jeopardizing the containment at an early phase of the accident.

The availability of state-of-the art validated numerical simulation tools is relevant to estimate the current knowledge on severe accidents, and to assess if further efforts on R&D are necessary to gain a deeper understanding of certain phenomena. The reference European code ASTEC, jointly developed by the “Institut de Radioprotection et de Sûreté Nucléaire (IRSN)” and the “Gesellschaft für Anlagen und Reaktorsicherheit (GRS)” is able to simulate the progression of a severe accident from the initiating event till the release of radioactive to the environment.

Nevertheless, the ASTEC code requires further validation against experimental facilities handling key physical phenomena in a severe accident sequence, and further application to nuclear power plants. Within the European project CESAM (2013-2017), embedded within the framework program FP7, ASTEC capabilities concerning SAM will be improved both via validation against reactor-scalable experiments and plant calculations on European NPPs.

In the frame of the present work the ASTECV2.1 code will be validated against the QUENCH Nr-06 experiment, this test studying the impact of the reflooding of an overheated core on the hydrogen generation and core degradation progression. As a second step, an assessment of the ASTECV21 modelling uncertainties of the QUENCH-06 will be performed by use of the URANIE software, and the main sources of uncertainty will be identified.

Pre-requisites: Mechanical engineering, Physics, Reactor Physics, programming skills
Welcoming skills: use of Linux OS
Duration: 4-6 Months
Location: KIT Campus North, Institute for Neutron Physics and Reactor Technology (INR), Building 521

Contact persons:

  • Ignacio Gómez García-Toraño (ignacio.torano@kit.edu,+49 (0)721 608-29281)
  • Dr. Victor Sanchez (victor.sanchez@kit.edu, +49 (0)721 608-22283)