Below you find all possible topics, but please take into account that the priority have the topics selected for further admission to NCBJ Graduate School in June 2020 which are:
Prof. Mariusz Dąbrowski
Topic 2A. Stability Integrated risk assessment for combined nuclear and chemical facilities
This project concerns the most demanding problems on risk assessment for the chemical installations coupled with High or Very High-Temperature Reactors (HTR/VHTR) optimized to the process heat production. Replacement of the coal or gas boilers by HTRs within the chemical plants requires, among others, considering the interactions between nuclear and non-nuclear parts of the installation, combined together in one complex system. Independently developed risk models for chemical and nuclear parts of such facilities would not describe properly the real state of the whole system and thus need to be integrated. The aim of the PhD work is to develop the overall risk assessment framework for combined nuclear and chemical facilities by taking into account the events posing a challenge for safety of the nuclear installation after failure of one or more chemical systems and vice versa. Within this framework there is a need for identification of the critical elements, the failure of which could lead to severe accidents as well as the accident sequences. Significant improvement of the current status of knowledge is expected in characterization of the risk associated with the physical phenomena and chemical processes applied in one part of the system (chemical or nuclear) in the context of fragility of the other one. This includes also proposition of the risk metrics for such installation. The framework for an integrated risk assessment will also provide methods for assessing the risks associated with external hazards as the potential threats for one or both (nuclear and chemical) parts of the installation. The project might be developed in collaboration with Japan Atomic Energy Agency (JAEA) operating the High-Temperature Test Reactor (HTTR).
prof. Tomasz Kozłowski
Topic 13A: 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.
prof. Wacław Gudowski
Topic 16A: Nuclear Fuel Cycle studies of nuclear reactor fleet consisting of LWRs, HTGRs and other advanced reactors – development of the Nuclear Fuel Cycle Simulator – FANCSEE
In order to study material flow, material inventory, radioactivity and radiotoxicity of the final wastes in a nuclear power program it is obligatory to have a reliable and validated tool performing all necessary simulations. This PhD project is focused on development of an existing Nuclear Fuel Cycle Simulator FANCSEE which enables such simulations, but still requires a significant development. A candidate for such PhD project should have advanced skills in C++ programming and having good understanding of reactor physics and nuclear technology. Finally this simulator will be used for studying different scenarios of nuclear program development in Poland and other countries in order to optimize the costs of the back-end of the nuclear fuel cycle.
prof. Rafael Macian-Juan
Topic 17A: Technical safety-related assessment of transmutation plant with liquid fuel (DFR)
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.
List of all possible future reference topics including exam topics:
prof. Jerzy Cetnar
- Investigation of an alternative option of HTR configuration dedicated for mixed fuel cycle with thorium utilization
HTR is a versatile reactor in terms of utilized fuel. As the 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 a 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:
- Lead cooled reactor characteristics
- Basics of the reactor fuel cycle
- Radiotoxicity
prof. Konrad Czerski
Part III of exam topics proposals:
- Thorium fuel cycle in thermal reactors
- Thorium fuel cycle in fast reactors
- High-temperature corrosion of ceramic materials
- Stability analysis of nonlinear dynamical systems
prof. Mariusz Dąbrowski
- Integrated risk assessment for combined nuclear and chemical facilities (topic selected to NCBJ Graduate School admission 2020)
This project concerns the most demanding problems on risk assessment for the chemical installations coupled with High or Very High-Temperature Reactors (HTR/VHTR) optimized to the process heat production. Replacement of the coal or gas boilers by HTRs within the chemical plants requires, among others, considering the interactions between nuclear and non-nuclear parts of the installation, combined together in one complex system. Independently developed risk models for chemical and nuclear parts of such facilities would not describe properly the real state of the whole system and thus need to be integrated. The aim of the PhD work is to develop the overall risk assessment framework for combined nuclear and chemical facilities by taking into account the events posing a challenge for safety of the nuclear installation after failure of one or more chemical systems and vice versa. Within this framework there is a need for identification of the critical elements, the failure of which could lead to severe accidents as well as the accident sequences. Significant improvement of the current status of knowledge is expected in characterization of the risk associated with the physical phenomena and chemical processes applied in one part of the system (chemical or nuclear) in the context of fragility of the other one. This includes also proposition of the risk metrics for such installation. The framework for an integrated risk assessment will also provide methods for assessing the risks associated with external hazards as the potential threats for one or both (nuclear and chemical) parts of the installation. The project might be developed in collaboration with Japan Atomic Energy Agency (JAEA) operating the High-Temperature Test Reactor (HTTR). - Start-up and shut-down procedure and safety calculations for DFR variants
One of the main novelty of the DFR is the fact that no control rods are necessary for steering and shutting-down the reactor. This is mainly due to the negative temperature coefficient of the reactor which is basically the depression of the neutron flux after the rise of temperature as a consequence of density decrease of both fuel and coolant. However, the start-up and especially the shut-down process seem to be requiring the control rods for some “traditionally thinking” nuclear reactor designer. The task of this project would be to make detailed calculations of the start-up and shut-down processes which involve pre-heating both fuel and coolant and pumping them from sub-critical tanks into the reactor core via fuel and coolant channels and finally emptying the reactor core to the same sub-critical tanks. In the whole process the role is played by the Pyrochemical Processing Unit (PPU) which provides the appropriate fuel mixture which can be critical inside the core and this should also be investigated. The safety issues (gaining criticality, fuel processing in PPU, emergency shut-down, etc.) will be considered for various DFR models (variants of molten salt and metallic fuels, different power units).
Part III of exam topics proposals:
- Energy Returned on Invested (EROI) definition and applications
- Molten salt reactor characteristics
- Generations of nuclear reactors – classification and basic features
- The laws of thermodynamics in nuclear engineering
prof. Wacław Gudowski
- Nuclear Fuel Cycle studies of nuclear reactor fleet consisting of LWRs, HTGRs and other advanced reactors – development of the Nuclear Fuel Cycle Simulator – FANCSEE” (topic selected to NCBJ Graduate School admission 2020)
In order to study material flow, material inventory, radioactivity and radiotoxicity of the final wastes in a nuclear power program it is obligatory to have a reliable and validated tool performing all necessary simulations. This PhD project is focused on development of an existing Nuclear Fuel Cycle Simulator FANCSEE which enables such simulations, but still requires a significant development. A candidate for such PhD project should have advanced skills in C++ programming and having good understanding of reactor physics and nuclear technology. Finally this simulator will be used for studying different scenarios of nuclear program development in Poland and other countries in order to optimize the costs of the back-end of the nuclear fuel cycle.
- Back end of the nuclear fuel cycle of HTGR- solutions for the spent HTGR fuel disposal This PhD project will be focused on studies of the back-end of the nuclear fuel cycle of HTGR in particular on finding good and cost-efficient solution for the final disposal of the TRISO spent fuel and irradiated graphite blocks. The project will pose and answer some of the important questions like: is the geological disposal model of LWR’s spent fuel a suitable option for HTGR fuel? Are there any possible reprocessing options for TRISO fuel? Will it be economically justified? What are the best options for the final disposal of the HTGR spent fuel?
Part III of exam topics proposals:
- Temperature feedbacks in [HTGR/DFR/MSR/SFR/LFR] reactor
- Fuel separation/reprocessing options for [HTGR/DFR/MSR/SFR/LFR] reactor
- Comparison of LWR and [HTGR/DFR/MSR/SFR/LFR] fuel cycle
- Capabilities and limitations of CFD turbulence models for HTGR modeling
prof. Tomasz Kozłowski
- Multi-Physics Uncertainty Analysis of High Temperature Gas Cooled Reactor (topic selected to NCBJ Graduate School admission 2020)
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. -
Physical models uncertainty quantification of thermal-hydraulics models HTGR safety analysis
The statistical uncertainty methodologies, such as the ones used by DAKOTA and URANIE, are well developed and commonly used in research and industry. The initial step of such uncertainty methodology is selection of the uncertain parameters and their probability distribution functions (PDFs). Since most of the physical models implement in thermal-hydraulics codes are correlations with best-fit coefficients, it is vital that proper uncertainty quantification of these physical models is carried out to represent uncertainty of the simulation results. The proposed project consists of two parts. First, estimate PDFs (uncertainties) of the physical models used in thermal-hydraulics models used for HTGR safety analysis. Ideally, the PDFs should be calculated based on the original experimental data used to derive each physical model/correlation used in the thermal-hydraulics code. However, if the original data is lost or unavailable, a Bayesian-based inverse method can be used with other relevant experimental measurements. The second part of the project is to apply the statistical uncertainty methodology already implemented in DAKOTA (or similar) code to calculate the uncertainty of HTGR thermal-hydraulics prediction of safety significance, such as hot spots of pebble bed core.
Part III of exam topics proposals:
- Validation process the nuclear thermal-hydraulics codes/models
- Validation process the nuclear reactor physics codes/models
prof. Rafael Macian-Juan
- Technical safety-related assessment of transmutation plant with liquid fuel (DFR) (topic selected to NCBJ Graduate School admission 2020)
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:
- Minor actinide transmutation behavior in thermal reactors
- Minor actinide transmutation behavior in fast reactor
