The main object of this diploma thesis is solving problems concerning heat transfer in disposal cannister for spent nuclear fuel. In forepart possibilities of conceptual solving according of disposal cannister to particular states are reviwed. On the basis of this a variant of possible protect of a nuclear fuel repository in the Czech republic has been chosen for calculationof a simplified model. Second part is computational solving that was divided into two parts. The first deals with calculation of heat transfer in disposal canister and is done by an analytical method. In the second part is calculation is done by numerical model. In this way region in near surroundings of this model of disposal cannister is analysed. Last part those diploma thesis deals with design of the storage of spacing among disposal canisters as well as optimum placing in underground part of nuclear fuel repository.

Dedicated sensors are widely used throughout many industries to monitor everyday operations, maintain safety, and report performance characteristics. In order to adopt a more sustainable solution, much research is being applied to self-powered sensing, implementing solutions which harvest wasted ambient energy sources to power these dedicated sensors. The adoption of not only wireless sensor nodes, but also self-powered capabilities in the nuclear energy process is critical as it can address issues in the overall safety and longevity of nuclear power. The removal of wires for data and power transmission can greatly reduce the cost of both installation and upkeep of power plants, while self-powered capabilities can further reduce effort and money spent in replacing batteries, and importantly may enable sensors to work even in losses to power across the plant, increasing plant safety. This thesis outlines three harvesting opportunities in the nuclear energy process from: thermal, vibration, and radiation sources in the main structure of the power plant, and from thermal and radiation energy from spent fuel in dry cask storage. Thermal energy harvesters for the primary and secondary coolant loops are outlined, and experimental analysis done on their longevity in high-radiation environments is discussed. A vibrational energy harvester for large rotating plant machine vibration is designed, prototyped, and tested, and a model is produced to describe its motion and energy output. Finally, an introduction to the design of a gamma radiation and thermal energy harvester for spent nuclear fuel canisters is discussed, and further research steps are suggested. / Master of Science

Solving radiation shielding problems, i.e. deep penetration problems, is a challenging task from both computation time and resource aspects in field of nuclear engineering. This is mainly because of the complexity of the governing equation for neutral particle transport - Linear Boltzmann Equation (LBE). The LBE includes seven independent variables with presence of integral and differential operators. Moreover, the low successive rate of radiation shielding problem is also challenging for solving such problems. In this dissertation, the Detector Response Function (DRF) methodology is proposed and developed for real-time and accurate radiation shielding calculation. The real-time capability of solving radiation shielding problem is very important for: (1) Safety and monitoring of nuclear systems; (2) Nuclear non-proliferation; and (3) Sensitivity study and Uncertainty quantification. Traditionally, the difficulties of solving radiation problem are: (1) Very long computation time using Monte Carlo method; (2) Extremely large memory requirement for deterministic method; and (3) Re-calculations using hybrid method. Among all of them, the hybrid method, typically Monte Carlo + deterministic, is capable of solving radiation shielding problem more efficiently than either Monte Carlo or deterministic methods. However, none of the aforementioned methods are capable of performing "real-time" radiation shielding calculation. Literature survey reveals a number of investigation on improving or developing efficient methods for radiation shielding calculation. These methods can be categorized by: (1) Using variance reduction techniques to improve successive rate of Monte Carlo method; and (2) Developing numerical techniques to improve convergence rate and avoid unphysical behavior for deterministic method. These methods are considered clever and useful for the radiation transport community. However, real-time radiation shielding calculation capability is still missing although the aforementioned advanced methods are able to accelerate the calculation efficiency significantly. In addition, very few methods are "Physics-based" For example, the mean free path of neutrons are typically orders of magnitude smaller than a nuclear system, i.e. nuclear reactor. Each individual neutron will not travel too far before its history is terminated. This is called the "loosely coupled" nature of nuclear systems. In principle, a radiation shielding problem can be potentially decomposed into pieces and solved more efficient. In the DRF methodology, the DRF coefficients are pre-calculated with dependency of several parameters. These coefficients can be directly coupled with radiation source calculated from other code system, i.e. RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system. With this arrangement, detector/dosimeter response can be calculated on the fly. Thus far, the DRF methodology has been incorporated into the RAPID code system, and applied on four different benchmark problems: (1) The GBC-32 Spent Nuclear Fuel (SNF) cask flooded with water with a $^3$He detector placed on the cask surface; (2) The VENUS-3 experimental Reactor Pressure Vessel (RPV) neutron fluence calculation benchmark problem; (3) RPV dosimetry using the Three-Mile Island Unit-1 (TMI-1) commercial reactor; and (4) A Dry storage SNF cask external dosimetry problem. The results show that dosimeter/detector response or dose value calculations using the DRF methodology are all within $2sigma$ relative statistical uncertainties of MCNP5 + CADIS (Consistent Adjoint Driven Importance Sampling) standard fixed-source calculation. The DRF methodology only requires order of seconds for the dosimeter/detector response or dose value calculations using 1 processor if the DRF coefficients are appropriately prepared. The DRF coefficients can be reused without re-calculations when a model configuration is changed. In contrast, the standard MCNP5 calculations typically require more than an hour using 8 processors, even using the CADIS methodology. The DRF methodology has enabled the capability of real-time radiation shielding calculation. The radiation transport community can be greatly benefited by the development of DRF methodology. Users can easily utilize the DRF methodology to perform parametric studies, sensitivity studies, and uncertainty quantifications. The DRF methodology can be applied on various radiation shielding problems, such as nuclear system monitoring and medical radiation facilities. The appropriate procedure of DRF methodology and necessary parameters on DRF coefficient dependency will be discussed in detail in this dissertation. / Doctor of Philosophy / Since the beginning of nuclear era, enormous amount of radiation applications have been proposed, developed, and applied in our daily life. The radiation is useful and beneficial when they are under control. However, there will be some "unwanted radiation" from these applications, which have to be shielded. For this, radiation shielding has become a very important task. To effectively shield the unwanted radiations, studying the thickness and design of the shields is important. Instead of directly performing experiments, computation is a more affordable and safer approach. The radiation shielding computation is typically an extremely difffficult task due to very limited "communication" between the radiation within the shield and detector outside the shield. In general, it is impractical to simulate the radiation shielding problems directly because the extremely expensive computation resources. Most of interactions of radiation are within the shield while we are only interested in how many of them penetrate through the shield. This is typically called "deep penetration" problems in the radiation transport community.

In a nuclear fuel cask, the heat generating spent fuel rods are packed in a housing and the resulting bundle is placed inside a cask of thick outer shell made of materials like lead or concrete. The cask presents a wide variation in geometrical dimensions ranging from the diameter of the rods to the diameter of the cask. To make the problem tractable, first the heat generating rod bundle alone is considered for analysis and the effective thermal conductance of the bundle is correlated in terms of the relevant parameters. In the second part, the bundle is represented as a solid of equivalent thermal conductance and the attention is focused on the modelling of the cask. The first part, dealing with the effective thermal conductance is solved using Fluent software, considering coupled conduction, natural convection and surface radiation in the heat generating rod bundle encased in a hexagonal sheath. Helium, argon, air and nitrogen are considered as working media inside the bundle. A correlation is obtained for the critical Rayleigh number which signifies the onset of natural convection. A correlation is also developed for the effective thermal conductance of the bundle, considering all the modes of transport, in terms of the maximum temperature in the rod bundle, pitch-to-diameter ratio, bundle dimension (or number of rods), heat generation rate and the sheath temperature. The correlation covers pitch-to-diameter ratios in the range 1.1-2, number of rods ranging from 19 to 217 and the heat generation rates encountered in practical applications. The second part deals with the heat transfer modeling of the cask with the bundle represented as a solid of effective (or equivalent) thermal conductance. The mathematical model describes two-dimensional conjugate natural convection and its interaction with surface radiation in the cask. Both Boussinesq and non-Boussinesq formulations have been considered for convection. Numerical solutions are obtained on a staggered mesh with a pressure correction method using a custom-made Fortran code. The surface radiation is coupled to the conduction and convection at the solid-fluid interfaces. Steady-state results are obtained using time-marching. Results for various quantities of interest, namely, the flow and temperature distributions, Nusselt numbers, and interface temperatures, are presented. The Grashof number based on the volumetric heat generation and gap width is varied from 105 to 5 ×109. The emissivities of the interfaces are varied from 0.2-0.8 for the radiative calculations. The solid-to-fluid thermal conductivity ratio for the inner cylinder is varied in the range 5-20 in the parametric studies. Simulations are also performed with thermal conductivity calculated in an iterative manner from bundle parameters. The dimensionless outer wall conductivity ratio is chosen to correspond to cask walls made of lead or concrete. The dimensionless thickness (with respect to gap width) of the outer shell is in the range of 0.0825-1, while the inner cylinder dimensionless radius is 0.2. Air is the working medium in the cask for which the Prandtl number is 0.71. Correlations are obtained for the average temperatures and Nusselt numbers at the inner interface in terms of the parameters. The radiation heat transfer is found to contribute significantly to the heat dissipation.