The continual increase in electricity dependence for the advancement of society has led to increased demand in electricity globally. This increased demand, among other things such as global warming interventions and energy security have encouraged the need to diversify electricity generation sources. Civilian use of nuclear power dates back to the 1950s. The United States of America and France are currently leading with the highest nuclear power generation in the world, generating 101 GWe and 63 GWe, respectively. Several countries such as China and the United Arab Emirates have committed to new nuclear build in order to increase their nuclear power generation capacities. Standing against the prospects of growth of the nuclear power industry are technical and nontechnical challenges. These include proliferation risk, safety, high capital costs and high-level waste management. Most spent nuclear fuel from power reactors is currently stored in the spent fuel pools on reactor sites, and some have been reprocessed. It is estimated that about 32% (370 000 tons of Heavy Metal) of the total spent fuel generated from power reactors have been reprocessed up to date. With most of the spent fuel pools filling up, alternative interim and long term disposal of spent nuclear fuel solutions have been under investigation from as early as the 1970s. South Africa has planned an interim dry storage facility for the spent nuclear fuel to be established at the existing Koeberg power station. The interim dry storage facility will make use of HI-STAR 100 multi-purpose casks to store spent nuclear fuel until the country decides on final disposal solution. There are many aspects that are critical to safe, efficient and cost-effective long term storage of spent nuclear fuel. Some of the physics and engineering aspects concerning dry storage facilities are briefly discussed. The aspects presented here are: radiation containment, spent fuel, sub-criticality, decay heat removal, site location aspects, response to seismic events, cask corrosion, transportation infrastructure, operability and monitoring. The study of the three existing dry cask storages from the USA, Hungary and Belgium gives an overview of the dry cask technology in use today. These presentations are based on publicly available reliable information. The proposed dry storage facility at Koeberg will be in the existing power station footprint using the HI-STAR 100 casks. The decision to have the proposed dry storage facility at Koeberg will minimise related licence applications and part of security installations as the site already has some security. The location of the facility in the power station’s footprint also allows for cost-effective and safe transportation of casks from the reactor building to the proposed facility. The modularity aspect of the dry cask storage facility at MV Paks in Hungary should also be employed at Koeberg to allow for more storage. This will cater for additional casks that may need to be stored if more nuclear power plants are procured in the future. South Africa’s air traffic around the Western Cape is not as congested as Belgium’s. There is, therefore, no need for the casks to be housed in concrete buildings like Doel’s. Most of Koeberg’s high-level waste would have had a longer cooling time in the pools compared to the minimum cooling time required for the chosen cask technology. This will provide a conservative, safe approach for Koeberg’s facility. Dry cask storage technology has provided a reliable interim dry storage solution for several countries. Despite uncertainties for long term disposal options, the proposed dry cask storage facility at Koeberg is a suitable interim storage alternative for South Africa to allow continuous operation of the plant. This conclusion is based on the physics and engineering aspects that have been presented in this minor dissertation.

The PTR system allows the EPR2 fuel pool to be cooled. The evacuation of the residual power fromthe pool is ensured by several heat exchangers and pumps, which have to be dimensioned in order to meetdifferent requirements.In order to dimension them, the worst-case scenario of the components must first be determined.Sensitivity to external conditions and efficiency studies enable to propose a heat exchanger design tomeet the requirements. A parametric study then allows to study more precisely the influence of thegeometry of the exchanger on the heat transfer. This allows to guide the conception of a CFD study ofthe design on the Comsol software in order to validate it. The proposed design can then be integratedinto the PTR cooling train. The train is modeled with FloMaster, in order to compute the head losses inthe hydraulic system and to propose a pump altimetry preventing cavitation.The dimensioning case of the exchangers corresponds to the operating case of the PTR trains duringunit shutdown, while the scenario that facilitates cavitation corresponds to the boiling of the fuel pool.The temperature of the cold source RRI is a sensitive data for the operation of the exchangers. In addition,the placement of the baffles and the space between the tubes play a determining role in the heat removal.It was difficult to construct the desired exchanger geometry in CFD. A compromise model was thusidentified and studied in CFD. The FloMaster study showed that the pressure drop in the PTR network isabout 15.5 mCE at the considered flow rate. Cavitation in a main train is not a problem if the pumps arelowered by at least 1.8 meters from the pool suction point.The sizing study therefore allowed us to propose a heat exchanger design close to the specifications,but this could not be precisely studied in CFD. The pressure drop study allowed to propose a pumpaltimetry preventing cavitation. / PTR-systemet gör det möjligt att kyla bränslebassängen i EPR2. Evakueringen av den återstående energin frånfrån bassängen säkerställs av flera värmeväxlare och pumpar, som måste dimensioneras för att uppfyllaolika krav.För att kunna dimensionera dem måste man först fastställa det värsta tänkbara scenariot för komponenterna.Känslighet för yttre förhållanden och effektivitetsstudier gör det möjligt att föreslå en värmeväxlardesign somuppfyller kraven. En parametrisk studie gör det sedan möjligt att mer exakt studera påverkan avväxlarens geometri har på värmeöverföringen. Detta gör det möjligt att vägleda utformningen av en CFD-studie avav konstruktionen i programvaran Comsol för att validera den. Den föreslagna konstruktionen kan sedan integrerasi PTR-kyltåget. Tåget modelleras med FloMaster, för att beräkna huvudförlusterna ihydraulsystemet och för att föreslå en pumphöjdmätning som förhindrar kavitation.Dimensioneringsfallet för växlarna motsvarar driftsfallet för PTR-tågen under driftavställning avenhetens avstängning, medan det scenario som underlättar kavitation motsvarar kokning av bränslebassängen.Temperaturen hos den kalla källan RRI är en känslig uppgift för driften av växlarna. Dessutom måsteplaceringen av bafflarna och utrymmet mellan rören en avgörande roll för värmeavledningen.Det var svårt att konstruera den önskade växlargeometrin i CFD. En kompromissmodell identifierades därföridentifierades och studerades i CFD. FloMaster-studien visade att tryckfallet i PTR-nätverket ärcirka 15,5 mCE vid det aktuella flödet. Kavitation i ett huvudtåg är inte ett problem om pumparna ärsänks med minst 1,8 meter från poolens sugpunkt.Dimensioneringsstudien gjorde det därför möjligt för oss att föreslå en värmeväxlardesign som ligger nära specifikationerna,men detta kunde inte studeras exakt i CFD. Tryckfallsstudien gjorde det möjligt att föreslå en pumpaltimetri som förhindrar kavitation.

This diploma thesis deals with storing the spent nuclear fuel and reviewing its safety. The theoretical part analyzes the processes taking place while the fuel is being used, such as fission, isotopic changes, fission gas release, cracking, swelling and densification of fuel pellet. The thesis is also focused on handling the spent fuel and on the way it makes from the reactor, through the spent fuel pool, the transportation, various kinds of storing, till the reprocessing and final deep geological repository. Furthermore, this part of the thesis briefly discusses computing code MCNP, its main characteristics, input files and using. The practical part of the work is focused on creating the model of the spent fuel pool located next to the nuclear reactor WWER 440/V213. This type was chosen, because it is the most used type of nuclear reactor in Czech Republic and Slovakia. With the help of the code MCNP, the multiplication factor of the main configurations of the fuel in the pool was calculated, and then the required safety regulations to ensure sufficient subcriticality, so its safety, were checked. Next, several analysis were performed using this model. These analyses were concerning the temperature of coolant, fuel and the use of various nuclear data libraries. In the future this model can be used to realize new analyses with new kinds of fuels, materials and data libraries.

This dissertation covers the application of heat accumulation systems in nuclear power engineering, namely in nuclear power plants. It is mainly a case of passive emergency systems, whose task is to accumulate the heat produced in the reactor’s active zone and in spent fuel pools during DBA (design-basis accidents) or beyond DBA. A particular example of heat accumulation is steam condensation after LOCA (loss of coolant accident). The primary circuit steam leakage increases containment pressure and has to be decreased by the steam condensation. This thesis deals with a theoretical substitute for ice condensers, which are used as a passive safety measure in some nuclear power plants. The substitute involves a choice of an alternative material, whose melting temperature (for heat accumulation) is closer to nuclear power plant operating temperatures. The other part of the dissertation discusses heat accumulation in spent fuel pools in case of all cooling systems failure.

The diploma thesis aim to the design of equipment for ŠKODA JS a.s., which is part of a new inspection stand (N-SIO). This equipment is an inspection column that provides the possibility to inspect spent fuel assemblies at the operation of the Temelín nuclear power plant. This master thesis is also a summary of the whole design of the new inspection stand and a description of the individual inspection components and devices. One of the parts of the thesis is also a basic strength analysis and a drawing of the main assembly of inspection equipment.

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-of-coolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based control-volume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy. / text