PhD Topics

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prof. Jerzy Cetnar

  1. Investigation of alternative option of HTR configuration dedicated for mixed fuel cycle with thorium utilization
    HTR is a versatile reactor in terms of utilised fuel. As thorium-uranium cycle can be an attractive alternative to uranium-plutonium one it has its challenges due to lack of U233 in nature. Th-U cycle in existing solutions involves thorium irradiation in a reactor blanket and then its reprocessing after discharge from the reactor. As the thorium fuel reprocessing brings many challenges an alternative options that are based on once-through cycle or applying simplified separation (fission products removal only) will be examined in terms of reactor core design and separation feasibility.

Part III of exam topics proposals:

  1. Lead cooled reactor characteristics
  2. Basics of reactor fuel cycle
  3. Radiotoxicity

prof. Konrad Czerski

  1. High-temperature corrosion of ceramic materials – includes corrosion oven construction
    One of the main problem of the high temperature reactors is to have very good refractory materials which could be resistant to corrosion and other mechanical and embrittlement factors. These materials should be very resilient in ambient temperatures up to 1100 C and even up to 1700 C for short times. The best known candidates are carbides SiC, ZrC, TiC and their compounds ZrC-TiC. As for the fluids some eutectics made of natural uranium U238 for fuel loop and natural lead (Pb) for cooling loop should be tested. This Ph.D. topic will contain the problem of creating a high temperature facility – the corrosion oven, which could make some precise remote-handled mechanical testing. The program contains: oven construction, electronic control system construction, long-period studies under high temperature, stirring equipment preparation (rotating up to 300 rpm to 1000 rpm depending on bath size) to simulate Pb coolant velocity, decontamination of samples, some diagnostic experiments at nanoscale. The final output will be the choice of best construction material chosen for the metallic DFR/m. The facility will also allow to measure the physical and chemical parameters simulated by prepared independently computer model of the reactor.
  2. Analysis and development of an MHD Pump for the DFR
    Construction and testing of the laboratory-scale Mini-Demonstrator, consisting of two small loops: fuel (depleted Uranium) and coolant (Lead), would require some important components: at least two MHD pumps, vane pumps, melting plugs, ceramic tubes, heating facility, TZM pipes, melting pugs, electronic measurement system, and storage tanks. The Mini-Demonstrator will be used for stability and operation tests of the DFR/m. Particularly for the lead loop a MHD pump with an equalized flux profile has to be developed. One needs to find an appropriate MHD centrifuge for fuel refining. Alternatively, classical Pb-pumps can be also tested. A proper technique for the Pb/H2O heat exchanger should be chosen. The project would concentrate on the optimization of an MHD pomp selection for the facility, first studying the theoretical background and then selecting appropriate device for the Mini-Demonstrator.

Part III of exam topics proposals:

  1. Thorium fuel cycle in thermal reactors
  2. Thorium fuel cycle in fast reactors
  3. High-temperature corrosion of ceramic materials
  4. Stability analysis of nonlinear dynamical systems

prof. Mariusz Dąbrowski

  1. Stability analysis of nonlinear dynamical calculations for the high-temperature reactors
    Within the framework of this project, the performance of special stability analysis methods, in particular the bifurcation theory approach, will be investigated and demonstrated by a specific example of complicated technical systems, the nuclear reactor. The results of the project will support both the deepening of knowledge about the safety characteristics of operating nuclear reactors and the build-up of the technical expertise of nuclear safety experts, who should be able to assess the nuclear safety of the global operating reactors and, in particular, of the high temperature nuclear reactors which are supposed to be erected in Poland.
    Four scientific issues of general meaning for nonlinear dynamical systems describing nuclear reactors are of particular interest:
    • Application of local and global bifurcation theory in the framework of stability analysis of nonlinear dynamical reactor systems.
    • Application of actual mathematically justified Model Order Reduction (MOR) methods with the goal of developing fast running stability analysis codes. This stability analysis approach is also very well suited to get an overview of the stability behavior of nonlinear dynamical systems in a wide parameter range.
    • Based on the experience that was achieved in the development of nuclear reactor Reduced Order Models (ROMs), the development of ROMs of GEN IV reactors (HTGR, DFR) in the rigorous MOR theory sense is to be prepared and realized.
    • A comprehensive assessment of the ROM analysis concerning reactor transient analyses (application potential in reactor dynamics).

Part III of exam topics proposals:

  1. Energy Returned on Invested (EROI) definition and applications
  2. Molten salt reactor characteristics
  3. Generations of nuclear reactors – classification and basic features
  4. The laws of thermodynamics in nuclear engineering

prof. Wacław Gudowski
(in collaboration with USNC – Ultra-Safe Nuclear Corporation)

  1. Modeling Fully Ceramic Microencapsulated (FCM) fuel
    The TRISO particle behaviour under irradiation in normal and accident conditions have already been modelled and it appears from comparison of experimental results on TRISO failures and calculations that failure mechanisms are rather well understood.  The objective of the PhD is to develop a model of the whole FCM pellet integrating a model of TRISO particles. The simplest solution would be to get access to an existing TRISO particle computer code, e.g. the INL code PARFUME and to develop a model of the behaviour of the SiC matrix and of the interactions between the matrix and the particles. If it is not possible to get access to a particle code, there is sufficient available literature to develop a reasonable one.  There is an experimental part of the work, to get the mechanical properties of as manufactured SiC matrix measured. The benefit of this PhD will be to get a better understanding of the behaviour of the HTGR fuel in general and of the FCM fuel in particular, to possibly optimise the design and manufacturing of this fuel.
  2. Very high temperature molten salt technology
    Solar molten salt can be used as heat carrier up to 570˚C. Above this temperature there are problems of stability of this molten salt. The objective of the PhD is to select molten salts that are stable at higher temperature and materials for the IHX that are compatible both with these molten salts and with impure helium atmosphere, in order to be able to increase the operating temperature of Micro Modular Reactor (MMR). Corrosion phenomena in impure He and in the molten salt will have to be addressed, as well as the mechanical properties of the materials, which will be considered, at operating temperature. The benefit of this PhD is to enable efficient very high temperature heat transport (> 600˚C), without which any step forward of HTGR technology towards VHTR would be useless for application to industrial process heat supply.
  3. Assessment of the source term – Reserved
    HTGR fuel has very low fission product release during operation and even in case of an accident. Nevertheless the fission products accumulating in the primary system during the whole lifetime of the reactor form the source term that can be released in the environment in case of a break in the primary containment. It is important for safety to have an assessment of the global activity that can be released into the environment. But even if measuring the circulating activity is easy and if some of the plated out activity could also be measured (e.g. the activity plated out in the Intermediate Heat Exchanger – IHX), part of the activity in the primary circuit remains inaccessible (e.g. activity adsorbed in graphite).  The objective of the PhD is to define a strategy for assessing continuously the global activity in the primary system through a limited number of accessible measurements (e.g. only the circulating activity or the circulating activity + the plated-out activity in the IHX). For this purpose, existing data and models on fission product transport, plate-out and lift-off should be revisited and the possibility to get new data on existing HTGR test reactors (HTTR, HTR-10) should be investigated, to improve the knowledge on the distribution of fission products in the primary system. An assessment of the uncertainty of the evaluation of the global activity should be made. The result of this PhD will be very useful for the licensing of the FOAK reactor.
  4. Impact assessment of introducing HTGRs in the Polish energy system
    HTGRs can supply industrial facilities with heat and electricity, and possibly basic raw materials like hydrogen and oxygen. Introducing them will have a large impact on the overall Polish energy system such as: reduction of CO2 emissions of the Polish industry, improving human health through better air quality, limiting the energy dependency and geostrategic risk of fuel supply disruption, limiting the consumption of local coal resources with a social impact or enabling the valorisation of unused coal resources into added value chemicals, generating low-carbon electricity, providing price certainty, improving local science, technology and industrial skills and capabilities, etc. To enable policy-making, these positive and negative impacts could be thoroughly assessed with a PhD, to make an overall balance of the interest of introducing HTGRs in Poland, learning from previous cases such as France. It could run Monte Carlo simulations to generate long-term scenarios. Collaboration with an energy model developer in Poland or abroad could be also fruitful to model the overall energy system in Poland (including non-industrial applications like buildings, services and transport). The PhD would feed policy-making with relevant data on HTGRs. USNC has experts with experience in EU energy policy-making and impact assessment, as well as cost-benefit analyses that could supervise a PhD. USNC could also carry out similar assessments for other countries where it operates.

Part III of exam topics proposals:

  1. Temperature feedbacks in [HTGR/DFR/MSR/SFR/LFR] reactor
  2. Fuel separation/reprocessing options for [HTGR/DFR/MSR/SFR/LFR] reactor
  3. Comparison of LWR and [HTGR/DFR/MSR/SFR/LFR] fuel cycle
  4. Capabilities and limitations of CFD turbulence models for HTGR modeling

prof. Tomasz Kozłowski

  1. Multi-Physics Uncertainty Analysis of High Temperature Gas Cooled Reactor
    The development of the High Temperature Gas Cooled Reactors (HTGRs) requires verification of HTGR design and safety features with reliable high fidelity physics models and robust, efficient, and accurate codes. The predictive capability of coupled neutronics/thermal-hydraulics simulations for reactor design and safety analysis can be assessed with sensitivity analysis (SA) and uncertainty analysis (UA) methods. Uncertainty originates from errors in physical data, manufacturing uncertainties, and modelling and computational algorithms. SA is helpful to partition the prediction uncertainty to various contributing sources of uncertainty and error. SA and UA is required to address cost, safety, and licensing needs and should be applied to all aspects of reactor multi-physics simulation. Current SA and UA rely either on derivative based methods, stochastic sampling methods, or on generalized perturbation theory to obtain sensitivity coefficients. The proposed project will develop and new hybrid multi-physics uncertainty method to quantify the impact of different sources of uncertainties on the design and safety parameters of HTGRs. Only a parallel effort in advanced simulation and in nuclear data improvement will be able to provide designers with more general and well-validated calculation tools to meet design target accuracies.

Part III of exam topics proposals:

  1. Validation process the nuclear thermal-hydraulics codes/models
  2. Validation process the nuclear reactor physics codes/models

prof. Rafael Macian-Juan

  1. Technical safety-related assessment of transmutation plant with liquid fuel (DFR) – Reserved
    In the framework of this PhD work, safety-critical issues on the transient and accident behavior of transmutation plants are being examined in detail with the aid of the DFR system code(s) developed at TUM and NCBJ.  Based on event trees that take into account all significant component failures, transient analyzes, including startup and shutdown simulations and stability analyzes are performed, including consideration of the mechanical integrity of the component materials as a result of abnormal occurrence or accidents, i.e. possible (component) consequential damages of abnormal occurrence or accidents are to be estimated. The aim is to prove that the plant fulfills all safety requirements that are set within the scope of the licensing procedure. The work within the PhD project will be carried out in the following four steps.
    1. Compilation of the principles of safety design of the MSR/DFRs. Development of event trees, which are the basis for the transient analyzes to be carried out in the next work step.
    2. Performing transient analyzes based on the event trees in 1. These are simulations with the calculation codes provided by TUM/NCBJ.
    3. Critical analysis of the simulation results from 2. and estimation of possible consequential damages.
    4. Development of a clear presentation of the safety characteristics and comparison of the different reactor variants.

Part III of exam topics proposals:

  1. Minor actinide transmutation behavior in thermal reactors
  2. Minor actinide transmutation behavior in fast reactor