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The supercritical carbon dioxide (sCO21) heat removal system, which is based on multiple closed Brayton cycles with sCO2 as the working fluid, is an innovative, self-propelling and modular heat removal system for existing and future nuclear power plants. Previous studies analysed its design, layout, control and operation. In addition, this novel study considers different sudden failures during the accident progress, e.g. failure of single sCO2 cycles, control systems and valves. These abnormal conditions were investigated with the thermalhydraulic system code ATHLET for a generic Konvoi pressurized water reactor. In most cases, the failure of a single sCO2 cycle can be compensated. On the one hand, failure of the fans of the gas cooler leads to a pressure increase which may be mitigated by an inventory control system or cycle shutdown. On the other hand, unintended fan speed-up may cause compressor surge without adequate countermeasures. Furthermore, the system can operate under the cyclic blow-off from the steam generator safety valves when the relief valves are not available. Finally, the unintended closure of the valve which controls the steam flow through the compact heat exchanger triggers a fast cycle shutdown but a subsequent restart might be possible., Conference Proceedings of the European sCO2 Conference5th European sCO2 Conference for Energy Systems: March 14-16, 2023, Prague, Czech Republic, p. 11

The supercritical carbon dioxide (sCO21) heat removal system, which is based on a closed Brayton cycle with sCO2 as a working fluid, is an innovative heat removal system for existing and future nuclear power plants. This paper provides the design, layout and control of the system based on assumptions developed in the project sCO2-4-NPP. A self-propelling operational readiness state enables a fast start-up and consumes only 12% of the design thermal power input. The system is analysed over a wide range of ambient and steam-side conditions in ATHLET, using performance maps for the turbomachinery, which were designed recently. The performance analysis suggests that it is a good option to operate the system at the design compressor inlet temperature of 55 °C at any boundary condition. With decreasing thermal power input, the rotational speed of the turbomachinery must be decreased to keep the system self-propelling. Moreover, the turbomachinery design with a higher surge margin is preferred. By controlling the compressor inlet temperature via the air mass flow rate and turbine inlet temperature via the turbomachinery speed, the heat removal system is successfully operated in interaction with a pressurized water reactor.

Innovative heat removal systems are currently investigated for use in existing and future nuclear power plants. One of them is the supercritical carbon dioxide (sCO21) heat removal system, which is based on a closed Brayton cycle with sCO2 as a working fluid. This paper provides the design and layout of the sCO2 cycle based on assumptions developed in the project sCO2-4-NPP. The system is analysed over a wide range of ambient and steam-side conditions in ATHLET, using performance maps for the turbomachinery, which were designed recently. Bypasses are considered in the layout of the cycle to cope with special operation conditions, e.g. start-up. Different operational readiness states for the system are shown, which enable a fast start-up of the system. Air mass flow rate control is implemented to keep the compressor inlet temperature constant with controller parameters depending on the ambient temperature. The performance analysis of the system suggests that it is a good option to operate the system at the design compressor inlet temperature of 55 °C at any ambient or steam-side boundary condition. With decreasing thermal power input, the rotational speed of the turbomachinery must be decreased to keep the system self-propelling. Turbomachinery design with a higher surge margin is preferred and different operation strategies are feasible and need to be tested in interaction with the nuclear power plant. 1sCO2 is defined as carbon dioxide at supercritical conditions with p > 73.8 bar and T > 31 °C, Conference Proceedings of the European sCO2 Conference4th European sCO2 Conference for Energy Systems: March 23-24, 2021, Online Conference, p. 14

In the frame of EU-project sCO2-flex the design of a 25 MWe supercritical CO2 (sCO2) Brayton cycle will be designed. The system will be optimized to meet flexibility requirements, while reducing environmental impact and focusing on cost efficiency. In the context of a sCO2 Brayton cycle, the gas cooler is a key component to achieve a high overall efficiency. Close to the critical point, due to varying properties, heat transfer and pressure drop of carbon dioxide (CO2) are difficult to predict. In case of vertical flow, acceleration and buoyancy effects induced by strong density gradients can cause a significant deterioration of the heat transfer. In this publication, the cooling heat transfer coefficient (htc) is investigated in a 3 mm diameter tube with vertical flow orientation. Commonly used calculation methods of the heat transfer coefficient are presented. Although developed for heating of sCO2, the mixed convection criterion of Jackson and Hall [7] is used to evaluate the heat transfer deterioration. The effects of the CO2 mass flux of 141 − 354 kg/m2s and bulk fluid temperatures of 20 − 50 °C with a constant pressure of 80 bar on the heat transfer were examined. The transition between forced and mixed convection can be explained by the htc -values. The upwards flow shows a steady decrease in the htc with the reduction of the mass flux. However, the downwards flow shows significant effects of buoyancy. At low mass flux the distinct peak in the htc at the pseudocritical temperature (Tpc) disappears., Conference Proceedings of the European sCO2 Conference4th European sCO2 Conference for Energy Systems: March 23-24, 2021, Online Conference, p. 146

This study proposes preliminary guidelines for the design and operation of a supercritical carbon-dioxide (sCO21) heat removal system for nuclear power plants. Based on a thermodynamic optimization the design point is calculated incorporating an existing small-scale compressor map. The behavior of the cycle is tested under varying boundary conditions on the steam side of the compact heat exchanger. The simulations are carried out with the thermal-hydraulic system code ATHLET, which has been extended for the simulation of sCO2 power cycles. The extensions include the thermodynamic properties, heat transfer and pressure drop correlations as well as performance map based turbomachinery models, which take the real gas behavior of sCO2 into account. During the decay heat transient, compressor surge occurs in some of the simulated cases. In order to avoid compressor surge and to follow the decay heat curve, the compressor speed is reduced together with the steam temperature. This enables to operate one single system down to a thermal load of less than 50 % even under the design restriction caused by the application of the existing compressor performance map. 1 sCO2 is defined as carbon dioxide at supercritical conditions with p > 73.8 bar and T > 31 °C, Conference Proceedings of the European sCO2 Conference4th European sCO2 Conference for Energy Systems: March 23-24, 2021, Online Conference, p. 400

In the frame of the sCO2-QA (qualification of analysis tools for the evaluation of a residual heat-driven, self-sufficient system for decay heat removal) project, a residual heat-driven self-sufficient sCO2-operated decay heat removal system based on a Brayton cycle is simulated with the German thermal-hydraulics system code ATHLET (analysis of thermal-hydraulics of leaks and transients). The heat removal system consists of a compact heat exchanger in the containment, a turbo-compressor system located in the reactor building and a gas cooler in the outdoor area. The validation of ATHLET and other numerical codes as well as understanding the heat transfer characteristics of sCO2 near the critical point requires experimental data. At IKE (Institute of Nuclear Technology and Energy Systems), the SCARLETT (supercritical carbon dioxide loop at IKE Stuttgart) test facility is available for various experiments with sCO2. This publication includes an experimental investigation of the thermal stratification in heated horizontal sCO2 pipe flows. For this investigation, eight test series with overall 48 experiments were carried out in two pipes with inner diameters of 4 mm and 8 mm. The experiments were carried out at a pressure of approximately 7.75 MPa. The target values of mass flux were set at 400 kg/m2s and 800 kg/m2s and those of heat flux at 50, 90 and 130 kW/m2, resulting in a heat to mass flux ratio of 62.5-225 J/kg. The inlet Reynolds numbers are between 16000 and 120000. The measured parameters are the flow rate, the pressure, the inlet and outlet fluid temperature as well as the outer surface temperature along the test pipe in three different radial angles. The results show the influence of the pipe diameter, Reynolds number, mass and heat flux on the temperature stratification. Also, inflow lengths were determined for a fully developed temperature stratification. This data set can be used for the validation of computer codes., Conference Proceedings of the European sCO2 Conference4th European sCO2 Conference for Energy Systems: March 23-24, 2021, Online Conference, p. 38

The sCO2 sCO2-4-NPP European project aims to develop an innovative technology based on supercritical CO2 (sCO2) for heat removal to improve the safety of current and future nuclear power plants. The heat removal from the reactor core will be achieved with multiple highly compact self self-prope llant, self self-launching, and self self-sustaining cooling system modules, powered by a sCO2 Brayton cycle. Heat exchangers are one of the key components required for advanced Brayton cycles using supercritical CO2. To this end, two compact and highly effective brazed plates and fins heat exchangers are designed: a main heat recovery heat exchanger (CHX), wh which allows the heat transfer directly from the steam generator to the sCO2 sCO2-4-NPP cycle, a heat sink exchanger (DUHS), which evacuate the remaining heat to the atmosphere. An important work has been achieved in the frame of this project to conceive the pre liminary design of these components, in close collaboration between Fives Cryo, a French brazed plates and fins heat exchangers manufacturer, the Institut für Kernenergetik und Energiesysteme (IKE) of University of Stuttgart and KSG/GFS Institute, a simula tor center, both in Germany. In fact, several constraints needed to be taken into account. For instance, for the DUHS, the low fans power available for the cycle and the necessary air flow for effective heat exchange implies considering almost inexisting pressure drops on the air side. To this end, very specific design ideas has been adressed to meet the desired thermal duty. Also, this project benefits from the recent results achieved among the european project sCO2-flex, related to the mechanical resistance of heat exchanger components, the assembly process and their thermal and hydraulic performances, along with Fives Cryo expertise and background. A second challenge of the sCO2-4-NPP project is to qualify the designed plates and fins heat exchangers, at cycle operating conditions, in order to meet with pressure vessels codes and regulations according to nuclear requirements. This paper presents the work achieved on the design of DUHS heat exchanger components and a preliminary part of the qualification of this equipment according to nuclear power plants regulations., Conference Proceedings of the European sCO2 Conference4th European sCO2 Conference for Energy Systems: March 23-24, 2021, Online Conference, p. 274

Innovative heat removal systems are currently investigated for use in existing and future nuclear power plants. One of them is the supercritical carbon dioxide (sCO21) heat removal system, which is based on a closed Brayton cycle with sCO2 as a working fluid. This paper provides the design and layout of the sCO2 cycle based on assumptions developed in the project sCO2-4-NPP. The system is analysed over a wide range of ambient and steam-side conditions in ATHLET, using performance maps for the turbomachinery, which were designed recently. Bypasses are considered in the layout of the cycle to cope with special operation conditions, e.g. start-up. Different operational readiness states for the system are shown, which enable a fast start-up of the system. Air mass flow rate control is implemented to keep the compressor inlet temperature constant with controller parameters depending on the ambient temperature. The performance analysis of the system suggests that it is a good option to operate the system at the design compressor inlet temperature of 55 °C at any ambient or steam-side boundary condition. With decreasing thermal power input, the rotational speed of the turbomachinery must be decreased to keep the system self-propelling. Turbomachinery design with a higher surge margin is preferred and different operation strategies are feasible and need to be tested in interaction with the nuclear power plant. 1sCO2 is defined as carbon dioxide at supercritical conditions with p > 73.8 bar and T > 31 °C OA platinum

In the frame of EU-project sCO2-flex the design of a 25 MWe supercritical CO2 (sCO2) Brayton cycle will be developed. The system will be optimized to meet flexibility requirements, while reducing environmental impact and focusing on cost efficiency. In the context of a sCO2 Brayton cycle, the gas cooler is a key component to achieve a high overall efficiency. Close to the critical point, due to varying properties, heat transfer and pressure drop of carbon dioxide (CO2) are difficult to predict. By performing experiments with the “SCARLETT” facility of the University of Stuttgart, expertise will be gained using CO2 as working fluid in the pseudocritical region. The results of the fundamental experiments will be used for validation and improvement of correlations to develop heat exchangers working with high efficiency. The heat transfer and pressure drop of carbon dioxide near the critical point cooled in a 2 mm diameter was investigated. The outer wall surface temperature is determined by soldering calibrated T-type thermocouples into the copper tube wall. Thermocouples are evenly distributed along the 1.2 m cooled length. The effects of the CO2 mass flux of 400-1400 kg/m²s, inlet pressures of 7.7-8.5 MPa, bulk fluid temperatures of 10-85 °C and the flow orientation (upward, downward and horizontal) on the heat transfer and pressure drop were examined., Conference Proceedings of the European sCO2 Conference3rd European Conference on Supercritical CO2 (sCO2) Power Systems 2019: 19th-20th September 2019, p. 255

Conference Proceedings of the European sCO2 Conference3rd European Conference on Supercritical CO2 (sCO2) Power Systems 2019: 19th-20th September 2019, p. 39

Conference Proceedings of the European sCO2 Conference3rd European Conference on Supercritical CO2 (sCO2) Power Systems 2019: 19th-20th September 2019, p. 30

The supercritical CO2 heat removal system is a very innovative reactor heat removal concept as it improves the safety of both currently operating and future BWRs and PWRs through a self-propellant, self-sustaining and self-launching, highly compact cooling system powered by an integrated Brayton-cycle using supercritical carbon dioxide (sCO2) as its working fluid. Since this system is powered by the decay heat itself, it provides new ways to deal with beyond design accidents. The turbine of a Brayton-cycle provides more energy than necessary to drive the compressor, which means that the sCO2-HeRo system provides electricity in addition. Therefore, this system can be an excellent backup cooling system for the reactor core in case of a Fukushima-like scenario, with a combined station blackout (SBO), loss of ultimate heat sink (LUHS) and loss of emergency cooling. In addition, this system might also be used as a heat removal system for the reactor in hot stand-by condition, removing the decay heat by keeping the reactor pressure vessel at operation temperature and pressure. The system is developed within a EU funded project called "Supercritical CO2 Heat Removal System, sCO2-HeRo". The objective of this project is to build a small-size demonstrator and install it at the PWR glass model at GfS in Essen, Germany. By means of this down-scaled demonstration unit, important operational data will be gained to demonstrate the feasibility of this heat removal system. OA platinum

Conference Proceedings of the European sCO2 Conference3rd European Conference on Supercritical CO2 (sCO2) Power Systems 2019: 19th-20th September 2019, p. 125

The High Performance Light Water Reactor is a nuclear reactor concept of the 4th generation which is cooled and moderated with supercritical water. The concept has been worked out by a consortium of European partners, co-funded by the European Commission. It features a once through steam cycle, a pressure vessel type reactor, and a compact containment with pressure suppression pool. The conceptual design enables to assess its feasibility, its safety features and its economic potential.

The aim of this work is to investigate important aspects of ex-vessel molten corium cooling by bottom flooding through a porous concrete core catcher. These are the transport of cooling water through the porous concrete to the melt and the fragmenta-tion and porosity formation in the melt due to the interaction with cooling water/steam. Firstly, this work investigates the hydraulic parameters of the porous concrete core catcher. Cooling water flow simulations are performed by means of the code COCOMO3D for the passive distribution of the cooling water into the melt layer from below with sufficient flow rates over the large reactor cavity. These investigations show that the permeability values of two different layers of the porous concrete and their relation to one another have a significant effect on the rising superficial velocity of the cooling water and on the required pressure head for the coolant supply. A methodology is developed in this work in order to optimize the core catcher for hydraulic properties, which can be applied to various boundary conditions. Based on this methodology, for the inlet configuration that provides the cooling water around to the core catcher over the whole perimeter, various concrete pairings can be chosen to provide sufficient cooling water into the molten corium uniformly, with a feasible pressure head. Due to the restrictions on the design for back fitting, in these cases the water supply can be provided to the core catcher only from a very small inlet connection. Thus, very high velocities of cooling water are expected around the inlet region, and the linear friction laws are not adequate anymore. To gather the needed data for improving the modelling a dedicated experiment set-up is built within the framework of this work, and the relation between pressure and superficial velocity of water for porous concrete samples from CometPC core catcher, which are provided by KIT, is measured. The falling head method is used which enables the measurements for a wide range of pressure values. The non-linear friction law with the values for permeability and passability obtained from the measurements is then implemented into COCOMO3D. The simula-tions are performed for restricted water inlet case with the quadratic friction law for porous concretes. These simulations show that back fitting of the porous concrete core catcher device with limited water inlet configuration can raise many challenges. Increasing the area of the water inlet and providing the water uniformly from the perimeter of the porous core catcher device is the more feasible approach for the reactor application. Finally, the fragmentation and porosity formation phenomenon caused by bottom flooding is modelled in this work. The initial molten corium-steam interaction is assumed as the decisive phenomenon having a lasting effect on the fragmentation. Therefore, the fragmentation is modelled as a void fraction modeling of the two-phase flow of molten corium and steam in thermal equilibrium. In order to simulate this two-phase flow a new model for the interfacial friction force for molten corium-steam two-phase flow has been necessary. A new Bubbly-Channel two-phase flow interfacial friction force model is developed in this work in order to model the porosity formation during corium cooling by bottom flooding. With a stand-alone simulation program, this new model is validated against the COMET and CometPC experiments with varying boundary conditions and material properties. The results show very good agreement with the final porosity range achieved in the experiments. The new interfacial friction model and the material properties of liquid corium are implemented into COCOMO3D code. The COCOMO3D simulations of twophase flow of molten corium and steam are performed for CometPC plus experiment data. The flow pattern of steam and melt during these simulations shows good replication of the post-test morphology of solidified porous thermite from the experiments as well as the dispersion and ejection that happened during the experiment. Two further cases are simulated for the reactor case, which show that the partial ablation of the sacrificial concrete between core catcher and molten corium has an effect on the coolant distribution in the compact melt layer. These investigations show that being able to simulate molten corium as a moving liquid, due to the new model, provides more realistic modelling of coolant ingression into the compact melt layer via bottom flooding and the space that coolant actually occupies, hence the fragmentation phenomenon. The two-phase flow modelling of molten corium and steam presented in this work can be extended further in the future to three-phase flow of molten corium-steam-water with evaporation phenomenon in order to model the entire cooling and solidification process by bottom cooling.

Kraftwerkskreisläufe mit überkritischem Kohlendioxid (sCO2) als Arbeitsmittel haben hohes Potenzial in Hinsicht von Effizienz- und Flexibilitätssteigerungen im Vergleich zu konventionell wasserdampfbetriebenen Kraftwerken. Das kompaktere Anlagenlayout reduziert den Materialeinsatz, was wiederum zu niedrigeren Investitionskosten führt. Bei der Wärmeabfuhr nahe dem kritischen Druck durchlaufen die thermophysikalischen Stoffeigenschaften abrupte Änderungen. Daher variieren die Wärmeübertragung und die hydraulischen Eigenschaften innerhalb eines engen Temperaturbereichs stark. Die vorliegende Arbeit untersucht die Wärmeübertragung und den Druckverlust von sCO2 in Rohrströmungen mit kleinem Durchmesser, um eine Designempfehlungen für neuartige Konzepte kompakter Wärmeübertrager zu liefern. Der Einsatz kompakter Wärmeübertrager hat das Potential die Effizienz, Lastwechselgeschwindigkeit und Zuverlässigkeit eines Kraftwerkskreislauf unter hohem CO2-Druck und kleiner Differenztemperatur zu verbessern. Es wurden zwei Teststrecken in Form eines Doppelrohrwärmeübertrager konzipiert und aufgebaut. Die sCO2-Rohrströmung, mit jeweils 2 mm und 3 mm Innendurchmesser, wird durch das Kühlmedium im Ringspalt abgekühlt. Temperaturmessungen in der Wandung des Innenrohres ermöglichen die quantitative Bewertung des Wärme-übertragungskoeffizienten der sCO2 Strömung. Die Arbeit wurde thematisch in vier Kaptitel unterteilt. In jedem Abschnitt wird eine Messkampagne vorgestellt und die Ergebnisse mit entsprechenden Literaturdaten verglichen. Mit dem 2 mm-Versuchsrohr wurde die turbulente Wärmeübertragung untersucht. Unter weitreichender Variation der Betriebsbedingungen von sCO2 (Druck, Temperatur, Massenfluss) und Kühlmedium (Volumenstrom, Temperatur) wurde der Einfluss auf den Wärmeübertragungskoeffizient untersucht. Die signifikanten Einflüsse aller Parametern wurden aufgezeigt und eine neue Nusselt-Wärmeübertragungs-gleichung wurde entwickelt. Die Wärmeübertragung in vertikaler Strömungsausrichtung wurde mit beiden Versuchsrohren untersucht. Eine deutliche Verschlechterung der Wärmeübertragung in der Abwärtsströmung wurde festgestellt, welche durch Beschleunigungs- und Auftriebseffekte verursacht wird. In den horizontalen Versuchsreihen mit dem 3 mm Rohr wurden Messungen doppelt durchgeführt mit jeweils Temperaturmessungen oben und unten in der Rohrwandung. Bei Variation von CO2-Massenfluss und CO2-Temperatur wurde eine Temperaturschichtung festgestellt. Mit beiden Versuchrohren wurde der Druckverlust untersucht. Mit jeweils einer isothermen Versuchskampagne wurden eine Gleichung für den Reibungs-koeffizienten validiert. Basierend darauf wurde die Vorhersagbarkeit des Druckverlustes beim Kühlen untersucht. Die Berechnung der Druckverluste mit den Stoffeigenschaften bei Kernströmungstemperatur zeigten die beste Übereinstimmung mit den experimentellen Messungen.

In mehreren Versuchsaufbauten wurde schrittweise die passive Kühlung eines Nasslagerbeckens durch geschlossene Zwei-Phasen-Thermosiphons experimentell untersucht. Durch die Verdampfung und Kondensation von Wasser in einem geschlossenen Rohr wird latente Wärme von einem Wasserbecken an die Umgebungsluft abgegeben. Die Temperatur des Nasslagerbeckens darf im Normalbetrieb 45°C nicht überschreiten. Untersucht wurden verschiedene Aufbau und Rohrkonfigurationen und -modifikationen, sowie mehrere Arbeitsmedien und thermisch variable Randbedingungen.

After the Fukushima accident, the interest of the scientific community in severe accident (SA) research has been renewed. Great efforts are being made internationally to reassess and strengthen the safety of nuclear power plants. The recriticality potential in debris beds formed after the core meltdown is one of the SA research issues that needs further attention, and it is also the focus of this work. An inadvertent criticality event may cause the release of nuclear radiation and have severe consequences. Thus, the criticality in debris beds must be evaluated to predict possible risks and establish the appropriate control measures if necessary. The available criticality data for debris beds are still very scarce. Thus, the Japan Atomic Energy Agency has begun the ambitious task of building a criticality map for debris beds. That is an arduous enterprise, which requires the investigation of appropriate debris bed models and numerous computations under a broad range of possible conditions. A global effort and international cooperation are essential. The present work aims to contribute to this common endeavor by improving debris bed models, extending the criticality database, and facilitating future analyses. Alternatives for modeling the debris bed characteristics with a potential influence on the criticality are discussed in this thesis, from the most conservative assumptions to more realistic approaches. Among other things, it was found that debris beds can be modeled with high accuracy as spheres regularly arranged in a water matrix if an adequate equivalent diameter is chosen. Besides, coupled neutronic-thermohydraulic calculations were proven to be not necessary for assessing the criticality of Fukushima debris beds. This work also investigates the criticality characteristics of UO2-concrete systems. The calculation results prove the good moderation capacities of concrete, which has a significant positive reactivity effect at very low porosities. Not only the bound water is capable of thermalizing neutrons but also the SiO2, a major component of concrete. Consequently, MCCI products should be treated carefully in the criticality analyses. A preliminary conservative criticality assessment of Fukushima debris beds has revealed safety parameter ranges, i.e., conditions for which recriticality can be excluded. On the one hand, dry debris beds cannot become critical under any conditions due to the lack of sufficient moderator. On the other hand, debris beds submerged in water will remain subcritical if the porosity is sufficiently low (< 0.24 for debris beds without concrete, < 0.1 if concrete is mixed with fuel), the mass is sufficiently small (< 124 kg), or the cooling water is sufficiently borated (> 2600 ppm B). Finally, a statistical method is proposed as an alternative and more realistic way to evaluate the criticality in debris beds. A first exploratory analysis of the debris bed at Fukushima Unit 1 reveals that the probability of recriticality is extremely low. Additionally, the sensitivity analysis has concluded that the amount of control rod material (B4C) mixed with fuel is by far the most relevant parameter. Other parameters with a strong correlation with keff are the percentage of fuel in the corium, the amount of debris in particulate form, and the debris bed spreading. Based on them, future areas of research and improvement are proposed.

In a loss-of-coolant accident, deteriorated cooling conditions facilitate the high-temperature oxidation of zirconium in steam. This exothermic reaction entails the acceleration of core degradation and causes the generation of hydrogen. Due to the significance of oxidation reaction modeling for accident progression, it is an essential part of severe accident simulation codes. However, due to high computation cost, integral codes commonly rely on simple parabolic rate laws as their oxidation models. These models are subject to assumptions such as isothermal conditions, semi-infinite media, and unlimited steam supply, which are infrequently met under accident conditions. Integral models, as considered here, relax these restrictive assumptions and have the potential to provide significant improvement, and they share the advantage of being relatively computationally inexpensive. Nonetheless, a satisfactory solution is yet to be found for their extension to cover zirconium oxidation with its various phase transitions and accompanying heat effects. The obstacles inhibiting further advances in oxidation reaction modeling using the integral approach are twofold. The first challenge arises from the high complexity of a thermodynamic system that can undergo phase transitions due to changes in both composition and temperature and that is complicated by chemical reactions, which cause a variety of heat effects. Therefore, a suitable numerical model must be both physically precise regarding the heat and mass transport mechanisms and must incorporate detailed thermodynamic data. The second challenge consists in the lack of experimental data providing insight into the oxidation reaction, especially under transient conditions, which, however, is much needed for the validation of new modeling approaches. In this light, this thesis presents the development of a fast-running advanced integral model and its novel coupling to a lookup table that comprises physically self-consistent thermodynamic data. It addresses the challenge of missing experimental data, which is suitable for its validation, using model tests in comparison to a spatially discretized model, which has also been developed in this thesis for this very purpose. Furthermore, this work draws comparisons with models proposed in the literature and models that are implemented in severe accident analysis codes today. It demonstrates the capabilities of the advanced integral model to describe complex heat effects and phase changes, which surpass the abilities of state-of-the-art modeling approaches. In terms of structure, this thesis can be divided into three parts. The first part is dedicated to a detailed physical description of the oxidation reaction and an investigation of the state of the art regarding its mathematical modeling in today's severe accident analysis codes. On that basis, limitations of existing modeling approaches are identified and requirements are defined that an advanced model must satisfy. In the second part, a description is given of the development of the thermodynamic lookup table, a material and transport property library, the advanced integral model, and the spatially discretized model. Following this, the third part presents a series of model tests, starting with verification measures, continuing with comparisons to models from the literature and the model used in ATHLET-CD, and finally addressing a test case, whose complexity regarding the occurring phase changes exceeds the range of applicability of today's modeling approaches. The result of this systematic development is an advanced integral model that captures both chemically- and thermally-induced phase transitions, and which describes heat sinks and sources with precise thermodynamic data. Hence, it overcomes the limitations of the commonly used parabolic rate approaches and surpasses the capabilities of existing integral models. Thus, in a single model, it achieves the urgently required capability of following the oxidation reaction through different phases of core degradation, starting from the classic oxidation of cladding tube surfaces, through the phases of steam starvation, over the melting and chemical dissolution processes, and ending with the solidification and oxidation of crusts. Consequently, the advanced integral model's improved prediction of heat and hydrogen generation has the potential to reduce the uncertainty associated with today's severe accident analysis codes.

The accident of Fukushima in Japan in March 2011 highlighted the need of further research on severe accidents and on the prediction capabilities of current integral system codes. The present work deals with simulation of quenching of hot particles bed, which may form in the core from the melting and fragmentation of core components during a severe accident. The coolability of hot debris bed, still generating decay heat and threatening to re-melt, is a key-issue in terms of severe accident management. Codes that are supposed to be used as decision-making tools have to be able to calculate accident sequences quickly and accurately enough. For this reason, the computation domain is typically coarsely discretized, yielding large mesh cells (mesh size > 20 cm). The capabilities of COCOMO-3D regarding simulations of quenching of hot debris bed is first assessed against experiments such as DEBRIS and PEARL, or by simulating the quenching of a reactor-scale debris bed, on which larger mesh cells can be generated like for integral codes. The latter simulation with large cells yields strong computation instabilities, due to the fact that two-phase cells, represented as a homogeneous water-steam mixture, are no longer representative of the real topology. Therefore, a new method is developed in order to track and reconstruct the quench front in an unstructured meshing evolving with time. Additionally, the mass, momentum and energy conservation equations for water and steam have to be locally adapted in order to take into account the moving quench front. Moreover, the present work proposes a method to reproduce the geometry of the debris bed domain, since a coarse meshing cannot reproduce pre-defined bed geometries properly, i.e. without smearing of the boundaries. The new modelling is verified against base cases that are analytically solvable. Finally, the capability of simulating with coarse meshing (and large cells), quickly and without any instabilities, was assessed by repeating the reactor-scale simulations.

In the frame of the sCO2-HeRo project, a self-launching, self-propelling and self-sustaining decay heat removal system with supercritical CO2 as working fluid is developed. This system can be attached to existing nuclear power plants and should reliably transfer the decay heat to an ultimate heat sink, in case of a combined station black-out and loss-of-the-ultimate-heat-sink accident scenario. Thereby the nuclear core is sufficiently cooled, which leads to safe conditions. To demonstrate the feasibility of such a system and to gain experimental experience, a small-scale sCO2-HeRo system is designed, built and installed into the pressurized water reactor glass model at GfS, Essen. The obtained experimental results are used to validate correlations and models for pressure drop and heat transfer, which are implemented in the German thermal-hydraulic code ATHLET. In consideration of the validated models and correlations, new ATHLET simulations of the sCO2-HeRo system attached to a NPP are performed and the results are analyzed. After the motivation, the state of the art is summarized, including an outline of the simulation work with the ATHLET code, a summary of sCO2 test facilities and a description of currently performed experimental heat transfer investigations in heat exchangers with sCO2 as working fluid. The main objectives of the work are derived from there. The chapter "sCO2-HeRo" starts with a description of the pressurized water reactor glass model. Afterwards, the basic sCO2-HeRo system is explained before a detailed description of the sCO2-HeRo system for the PWR glass model and for the reactor application is presented. Afterwards, cycle calculations are performed for both systems to determine the design point parameters in consideration of boundary conditions, restrictions and assumptions with respect to maximum generator excess electricity. In the following chapter, the test facility for the investigations on the heat transfer capability between condensing steam and sCO2 is described. It consists of the sCO2 SCARLETT loop, a high-pressure steam cycle and a low-pressure steam cycle.The installed measurement devices, measurement uncertainties and calculated error propagations are explained as well. After a fundamental classification of heat exchangers, the 7 heat exchanger test configurations are summarized before the diffusion bonding technique, the used plate material, the mechanical design, the plate design and the manufacturing steps of the heat exchanger test plates are described. In the chapter "Results", the measurement points are described and an overview of the performed measurement configurations is given before a summary of all measurement results is presented. After that, experimental results of the sCO2 pressure drop for unheated flows as well as for heated flows are depicted and explained, followed by the analysis of the experimental heat transfer results. The chapter "CHX for the PWR glass model" starts with a summary of the boundary conditions and measurement results with regard to the design of the heat exchanger for the sCO2-HeRo system of the glass model. Subsequently, the plate design is presented and manufacturing steps of the heat exchanger are described by means of pictures. The chapter "ATHLET simulations" starts with an introduction before the development of performance maps, models and the validation of correlations based on experimental results as well as CFD simulation results are described. In the following, these models and performance maps are transferred to a sCO2-HeRo system that can be attached to a nuclear power plant. Finally, further cycle simulations are carried out and the simulation results are analyzed.

Für das weitere Fortschreiten oder Beenden eines schweren Störfalls ist die Kühlbarkeit der Partikelschüttung von größter Bedeutung. Ist die Kühlbarkeit des Schüttbetts durch Wassereinspeisung in den Reaktordruckbehälter (oder später in den Sicherheitsbehälter) erreicht, kann die Kernschmelze gestoppt oder zumindest ihre Folgen abgeschwächt werden. Die Beurteilung der Vor- und Nachteile solcher Maßnahmen erfordert ausreichende Kenntnisse der dreidimensionalen thermohydraulischen Effekte. Dies stellt eine Herausforderung für die Weiterentwicklung schwerer Unfallcodes dar. Die Unfälle in TMI-2 und Fukushima Dai-ichi haben gezeigt, wie wichtig die Entwicklung geeigneter Computercodes und -modelle ist, um eine realistischere Darstellung der Phänomene bei einem schweren Störfall zu erhalten. Diese Simulationen könnten Antworten geben, wie solche Ereignisse bewältigt werden können und ein stabil kühlbarer Zustand erreicht werden kann. Unter Berücksichtigung der Ergebnisse der Untersuchungen des TMI-2-Unfalls ist zu erwarten, dass das Schüttbett, welches sich während des Verlaufs eines schweren Störfalls bilden kann, inhomogen und dreidimensional ist. Für realistische Simulationen ist es wichtig, reale Konfigurationen zu berücksichtigen: z. B. Regionen innerhalb des Schüttbetts mit niedrigerer oder höherer Porosität, Regionen mit größeren oder kleineren Partikeln, ähnlich der Endzustandskonfiguration des TMI-2-Unfalls. Der Schwerpunkt der vorliegenden Arbeit ist die Entwicklung der dreidimensionalen Version des MEWA-Codes für die Untersuchung der Kühlbarkeit von Schüttbetten und deren mehrdimensionaler Effekte. Der Code wurde entwickelt für die Beschreibung des Verhaltens von Kernmaterial während der späten Phase schwerer Störfälle in Leichtwasserreaktoren. Die Erweiterung des zweidimensionalen MEWA-Codes (MEWA 2D) auf 3D (MEWA 3D) umfasst die Erweiterung des physikalischen Modells (Einführung der dritten Dimension in den Erhaltungsgleichungen) sowie die Einführung eines unstrukturierten, nicht orthogonalen Gitters. Ebenfalls wurden die Datenstrukturen von zweidimensionalen Feldern (strukturiert) auf Listen (unstrukturiert) umgestellt. MEWA 3D wurde verifiziert und anhand mehrerer Experimente mit unterschiedlichen Konfigurationen und Randbedingungen validiert. Im Rahmen der Validierung konnte zum ersten Mal die Kühlbarkeit einer dreidimensionalen Konfiguration erfolgreich durchgeführt werden. Die Ergebnisse zeigen, dass MEWA 3D die Trends hinsichtlich des Einflusses multidimensionaler Effekte reproduzieren kann. Die gute Übereinstimmung mit den Experimenten bestätigt, dass MEWA 3D auch für Analysen unter schweren Reaktorunfallbedingungen geeignet ist. Abschließend wurde das dreidimensionale inhomogene Schüttbett von TMI-2 mit MEWA 3D zum ersten Mal simuliert. Ein Vergleich der Simulationen mit den Analysen des TMI-2-Schüttbetts zeigt eine durchweg gute Übereinstimmung.

Die vorliegende Arbeit befasst sich zum einen mit der Entwicklung und dem Aufbau einer Mehrzweck-Versuchsanlage für überkritisches CO2 und zum anderen, mit einer daran anknüpfenden theoretischen und experimentellen Untersuchung von diffusionsgeschweißten Kompaktwärmeübertragern. Zu diesem Zweck wird nach einer anfänglichen Beschreibung des Hintergrunds und des Ziels der Arbeiten im Rahmen dieser Dissertation eine Übersicht über weltweit bestehende Versuchsanlagen und Anlagenkonzepte gegeben, sowie die Grundlagen und der Stand des Wissens zum Thema Handhabung und Wärmeübertragung mittels überkritischem CO2 vermittelt. Im Weiteren werden Konzeption, Auslegung, Design und Aufbau der Versuchsanlage, sowie verwendete Mess-, Steuer- und Regelungstechnik erläutert. Der weitere Teil der Arbeit legt die Auslegung und den Aufbau des Versuchsaufbaus dar. Der abschließende Teil der Arbeit widmet sich den Ergebnissen aus den Experimenten und zieht einen Vergleich mit numerischen Analysen anhand eines vereinfachten Modells für die Wärmeübertragung in einem Kompaktwärmeübertrager. Es wird gezeigt, dass diffusionsgeschweißte Kompaktwärmeübertrager in Verbindung mit überkritischem CO2 großes Potential hinsichtlich Wärmeübertragung und Druckverlust zeigen, was wiederum eine Steigerung der Effizienz und eine Verringerung der Investitionskosten von CO2-Kreisläufen bedingt. Das höchste Potential weisen speziell Einströmbedingungen nahe dem kritischen Punkt von CO2 auf, welches aber durch eine Metastabilität in diesem Betriebszustand erkauft wird. Des Weiteren wird die Strömungsführung in den Verteilräumen des Kompaktwärmeübertragers als ausschlaggebender Faktor bezüglich der maximal erreichbaren Strömungsgleichverteilung und in der Konsequenz als beschränkende Größe der erzielbaren Wärmeübertragung und des geringstmöglichen Druckverlusts identifiziert.

In the aftermath of the Fukushima Dai-Ichi nuclear accident, the issue of corium coolability has received considerable attention in the severe accident research. One of the accident mitigation strategies for the ex-vessel debris cooling is the employment of deep pool water in the cavity below the Reactor Pressure Vessel (RPV). During a hypothetical Severe Accident (SA) in Light Water Reactors (LWR), degraded core materials released from the RPV after its failure will be fragmented and quenched by contact with water. The solidified particles will settle on the bottom forming a porous bed. However, this strategy succeeds only if the residual decay heat is sufficiently removed and the bed is thermally stabilized and will not re-melt again damaging the containment integrity. One of the main factors determining the ability of decay heat removal and long-term coolability of debris bed is its geometrical configuration. A flatter and broader bed can be easier cooled than a higher bed with the same mass of debris. For this purpose, the present work focuses on the development of a two-dimensional continuum model describing the formation process of the debris bed resulting from the deposition of the settling particles and their relocation along the surface of the heap. The mathematical model is based on a hyperbolic system of partial differential equations determining the overall bed height, the distribution of the flowing particles layer depth and the depth-averaged velocity component tangential to the sliding layer. Because of the hyperbolicity of the system, a successful implementation of a solver is challenging, notably when large gradients of the physical variables appear, e.g., for a moving front in the flowing layer or possibly formed shock waves during the deposition. In this thesis, several numerical methods are applied to solve the system and compared. The implemented Roe-solver has provided promising results, which are verified with analytical solutions in the steady state. The spatial convergence is also reported and quantified with the use of the Grid Convergence Index (GCI). A sensitivity analysis is subsequently performed to study the influence of the uncertainties in the input parameters on the bed geometry. A dedicated test facility, named BeForE, is designed and built in the framework of this study, with the aim of providing the necessary experimental data for the model validation. A series of tests were conducted using different shaped and multi-size mixtures of particles. It could be evinced, that in addition to the modeled particles sliding the smaller particles (< 3 mm) are subject to the influence of a suspension and convection flows, which are affecting the final bed shape. The comparison between the numerical and the experimental results has shown a very good agreement, notably for the cases where the two last-mentioned phenomena are less present. Moreover, the test facility could also be used to gain an insight into the influence of steam production on the particulate bed spreading. The decay-heat-induced coolant boiling and the resulting two-phase flow serve as a source of mechanical disturbance, which might lead ultimately to leveling of the debris bed (a.k.a. self-leveling). A series of experiments were then conducted by discharging solid particles in the two-dimensional viewing vessel of the facility, while air bubbles simulating the steam production are injected simultaneously from the bottom. Depending on the quantity of the settled particles on the top of each section of the vessel, air flow rate is so monitored and adjusted in time to simulate the corresponding amount of steam produced by the similar quantity of hot debris. This study shows that, in most of the cases, the two-phase flow inside the vessel alters the sedimentation process resulting in a broader and flatter bed than under quiescent conditions. However, it was observed that in the case of customarily formed concave beds in the quiescent conditions, the presence of the gas flow can change the mound shape to a convex type with a higher bed height, at least in the beginning. It was also shown that, for high gas flow rates, the convective flows induced by the bubble plumes inside the bed would contribute to the diminishment of self-leveling effect and a slower particles redistribution. Lastly, it was mathematically described how the steam production could reduce the characteristic angles of repose of a debris bed, putting forth a physical explanation of the self-leveling phenomenon. With the coupling of the developed continuum model with a model simulating the two-phase flow within the bed, a full numerical simulation of the avalanche-like particles motion during the self-leveling process could also be successfully provided. This allows a more accurate simulation of the bed formation process under the influence of steam production, which is of particular importance for the bed coolability and a decisive requirement for the nuclear accident progression and termination.

Der Rückbau von kerntechnischen Anlagen stellt eine große Herausforderung dar; zum einen wegen ihrer Komplexität, zum anderen wegen des Umfangs der zu planenden Ausführungsschritte. Bei der Vielzahl an Anlagen ist ein Hauptkriterium des sicheren und nachhaltigen Rückbaus der Schutz von Mensch und Umwelt. Dazu gehört auch die Reduzierung von radioaktivem Abfall, um eine sichere und kostengünstige Entsorgung zu gewährleisten. Bei der Zerlegung von hochaktiven Stahlbauteilen mittels Wasser-Abrasiv-Suspension-Schneidtechnik entstehen signifikante Sekundärabfälle. Mit der hier vorgestellten Entwicklung einer Separationsanlage mittels magnetischer Sortierung der entstehenden Korngemische wird ein wesentlicher Beitrag zur Reduzierung des radioaktiven Abfalls geleistet. Die hierzu entwickelte Separationsanlage trennt dabei die Stahlspäne (aktivierter Anteil) aus dem Korngemisch, wobei ein Stabmagnetfilter zur Anwendung kommt. Diese Art von Filter wird konventionell nicht für die Aufgabenstellung der magnetischen Sortierung von Korngemischen verwendet, verfügt jedoch über maßgebliche Vorteile für die Anwendung mit radioaktiven Korngemischen. Versuche mit dem entwickelten Separationsverfahren zeigen, dass je nach Parametereinstellung signifikante Unterschiede bei der Separation der Fraktionen aus dem Korngemisch entstehen. Fehlende Kenntnisse über diese Art von Magnetfilter machen daher eine genauere Untersuchung des Stabmagnetfilters zur Sortierung von Korngemischen unabdingbar, was Thema dieser Arbeit ist. Um eine Bewertung der Separationsmöglichkeiten sowie deren Versuchsergebnisse vorzunehmen, wurde eine detaillierte Charakterisierung des Korngemisches aus der Wasser-Abrasiv-Suspension-Schneidtechnik durchgeführt. Neben der Gewinnung von wertvollen Erkenntnissen über die Zusammensetzung des Korngemisches werden auch Partikelabrasion, Partikelformen und Eigenschaften der beinhalteten Elemente im Abrasiv betrachtet. Mithilfe eines Extinktionszählers konnte eine Messeinrichtung zur Partikelgrößenanalyse realisiert werden. Diese erlaubt eine schnelle und unkomplizierte qualitative Aussage über die Zusammensetzung des magnetisch abgeschiedenen Materials, womit eine Bewertung der Separation ermöglicht wird. Ergänzt werden die experimentellen Untersuchungen durch eine numerische Mehrphasenströmungssimulation des Stabmagnetfilters, um Rückschlüsse auf die Strömungsverhältnisse und ihre Auswirkung auf die Separation im Filter zu ermöglichen. Das hierbei entwickelte Separationsverfahren ermöglicht eine Separation des überwiegenden Stahlanteils im Korngemisch, wobei eine Effizienzrate von ca. 90 Massen-% erreicht wird. Bezogen auf die Problematik der radioaktiven Abfallreduzierung wird somit eine maßgebliche Reduktion des radiologischen Inventars im Korngemisch erreicht. Dies erleichtert signifikant die nachfolgende Handhabung, Konditionierung und Entsorgung. Gleichzeitig liefert die Arbeit die Grundlagen zu einer vertiefenden Verbesserung des Separationsverfahrens zur Fest/Fest-Sortierung von Korngemischen durch einen Stabmagnetfilter.

Ziel dieser Arbeit besteht in der Entwicklung eines Code-Systems zur Analyse transienter Störfälle in schnellen Natrium gekühlten Reaktoren (SFRs). Bei SFRs handelt es sich um ein Konzept, das gegenüber Leichtwasserreaktoren (LWRs) wesentliche Vorteile bietet. Aufgrund der exzellenten Brennstoffausnutzung, der Möglichkeit langlebigen, radioaktiven Abfall zu reduzieren und der Notwendigkeit schneller Reaktoren zur Umsetzung eines geschlossenen Brennstoffkreislaufs, erweist sich die Technik als besonders geeignet, den von Experten definierten Anforderung an künftige Reaktoren bezüglich Sicherheit, Nachhaltigkeit und Wirtschaftlichkeit gerecht zu werden. Die Umsetzung erfordert ein Kerndesign, bei dem der Kern während des Betriebs nicht in der Konfiguration mit der höchsten Reaktivität vorliegt. Während bei LWRs ein Kühlmittelverlust zum nuklearen Abschalten des Reaktors führt, führt ein solcher Störfall in einem SFR aufgrund fehlender parasitärer Neutronenabsorption und verminderter Neutronenbremsung zu einem Anstieg der Reaktivität. Diese neutronenphysikalischen Besonderheiten erfordern bei Analysen zum transienten Verhalten von SFRs daher eine kollektive Betrachtung von thermohydraulischem und neutronischem Verhalten zur korrekten Beschreibung der Entwicklung des Gesamtsystems während der Transienten. Zur Entwicklung eines Codesystems, das die Abhängigkeiten von Neutronik und Thermohydraulik erfassen kann, wird auf zwei bestehende Codes zurückgegriffen. Als Thermohydraulikmodul dient der Code ASTEC-Na (GRS,IRSN), der in einem internationalen Projekt speziell als SFR-Analysetool entwickelt wurde. Mit Validierungsrechnungen zu den CABRI-Versuchen wird die Eignung des Tools zur Beschreibung von Natrium gekühlten Systemen bestätigt. Zwar enthält ASTEC-Na ein Neutronik-Modul, das jedoch auf einem Punktkinetikmodell basiert. Damit können lokale Flussänderungen sowie Verschiebungen des Neutronenspektrums nicht erfasst werden. Um eine detaillierte Beschreibung der Neutronik und der Rückwirkungseffekte zu ermöglichen, wird der Diffusionscode DYN3D (HZDR) mit ASTEC-Na gekoppelt. Die Verwendung eines nodalen Diffusionscodes bedarf der Erstellung einer problemspezifischen Querschnittsdatenbasis. Neben einer geeigneten Gruppenstruktur muss die Datenbasis alle möglichen Systemzustände während der Transienten erfassen können. Der Monte Carlo Code SERPENT2 (VTT) hat standardmäßig Routinen implementiert, womit Querschnittsdaten in beliebiger Gruppenstruktur erstellt werden können. Das gekoppelte Code-System wird auf einen generischen SFR-Kern (OECD/NEA Benchmark) angewendet. Zu diesem Zweck wurde mit SERPENT2 eine Querschnittsdatenbasis in 35 Gruppen mit Variationen der Brennstofftemperatur, der Brennstoffdichte sowie der Kühlmitteldichte erzeugt. Diese Daten werden DYN3D als Input bereitgestellt. Durch Vergleich der Ergebnisse stationärer Rechnungen mit DYN3D und Monte Carlo Simulationen mit SERPENT2 und MCNP6 (LANL) wird die Eignung der Querschnitte und des erstellten DYN3D Modells verifiziert. Die Absteuerung der gekoppelten Rechnungen erfolgt über DYN3D und ist über Modifikationen am Quellcode eingebaut. ASTEC-Na wird in jedem Zeitschritt mittels „Restart“ aufgerufen. Restriktionen bezüglich des Quellcodes von ASTEC-Na machen eine interne Kopplung nicht möglich. Zusätzlich wurden Module entwickelt, die den Datenaustausch zwischen den beiden Codes organisieren. Zur Demonstration der Berechnung gekoppelter, transienter Analysen für SFRs werden zwei transiente Störfälle berechnet (OECD/NEA Benchmark): ein unkontrollierter Kühlmitteldurchsatz-Störfall, bei dem der Durchsatz exponentiell auf 10% des nominellen Durchflusses zurückgeht, sowie eine unkontrollierte Reaktivitätszufuhr mit einem externen Reaktivitätseintrag von 1,8$. Eine detaillierte Analyse über die Entwicklung der einzelnen Parameter zeigt, dass die gekoppelte Code-Version plausible Ergebnisse liefert und eine Vorhersage des Störfallablaufs ermöglicht.

This work investigates a mid-sized break loss of coolant accident (MBLOCA) with a 200 cm2 leak in the hot leg next to the pressurizer surge line in the most common reactor type which is the 1; 300MWel KONVOI class pressurized water reactor (PWR). The scope was to analyze different reflooding scenarios, where the reactor core has already started melting when the reflooding is initiated. For this scenario a basic simulation called „BaseCase“ is conducted to get an overview of the accident progression and all the phenomena that occur. It also serves as a basis for all the following studies. In the MBLOCA scenario all active emergency core cooling systems are supposed to work correctly until the flooding pool is empty. Regularly, a switch to sump circulation would take place at that time, but in the simulation it is assumed to fail. So from then on no more water injection into the primary system takes place, which will eventually lead to complete core destruction. In order to investigate whether stable and coolable core conditions can be reached by feeding water into the primary loops during core melting, a diverse set of reflooding scenarios were selected to be simulated based the findings of the BaseCase simulation The feeding can be realized by turning on the high or low pressure injection pumps (HPI / LPI) in sump circulation mode for example, and it can take place at different locations. This leads to different reflooding scenarios, which are analyzed for differences and cooling potential. From study A „Variation of the beginning of reflooding“, it could be conducted that at least up to 40 t of melt can be cooled in the core, if the reflooding system has a sufficient injection rate. The results of study B „Variation of the reflooding system“ show that the reflooding via the LPI creates a self-blocking effect, which compromises the injection, thus preventing the core quenching for a time. This effect does not occur while using the HPI, leading to the question whether a middle pressure injection (MPI) system might be the optimum for the reflooding of a partially degraded PWR core, since it combines a higher mass flow rate than the HPI with a higher pressure head than the LPI. The injection location varied in study C „Variation of the reflooding location“ does affect the simulations, but to a much lower extend than the used reflooding system. Nevertheless, a combined reflooding using both the hot and the cold legs seems to be the optimum. Study D „Variation of the number of available pumps“ revealed that at least two LPI pumps are necessary to performa successful quenching of the core area. The simulation with only one pump did end with a core status that seems to be stable as well, but even after a simulation time twice as high as usual, the core could not be flooded with water and has to be cooled by steam.

The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO2) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO2 Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO2 Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO2 Brayton cycles and to evaluate the introduced heat removal system. A first analysis of the system revealed the ability to remove the decay heat over more than 72 hours, even for combined station blackout and loss of ultimate heat sink scenarios. In addition, the simulations exposed an interaction between the retrofitted and already existing systems. Parameters, which influence the operation of the self-propelling heat removal system, have been identified and summarized in set of prerequisites. The simulations indicate the potential of the system to serve as a diverse heat removal system for existing boiling water reactors., Das Ziel der vorliegenden Arbeit war die Entwicklung eines Modells zur Simulation von Joule-Kreisläufen mit dem Arbeitsmedium überkritisches Kohlenstoffdioxid (sCO2), für die Anwendung in Siedewasserreaktoren. Dieses Modell wurde in das Systemrechenprogramm ATHLET integriert. Mit dieser modifizierten Version können neuartige Nachwärmeabfuhrsysteme, die auf einem geschlossenen Joule-Kreisprozess aufbauen und eine potentielle Nachrüstmaßnahme für bestehende Leichtwasserreaktoren darstellen, modelliert und thermohydraulisch untersucht werden. Besonders wichtig sind hierbei Erkenntnisse über das instationäre Verhalten des Nachwärmeabfuhrsystems und dessen Reaktion auf die sich im Laufe eines Störfallszenarios verändernden Randbedingungen. Zur Simulation des neuen Arbeitsmediums wurden zunächst die Stoffdaten von überkritischem Kohlenstoffdioxid in den Systemcode eingebaut. Des Weiteren wurde eine Korrelation zur Bestimmung des Wärmeübergangskoeffizienten ausgewählt, mit der sich das Aufheizen und Abkühlen von sCO2 berechnen lässt. Druckverluste durch Wandreibung werden mit einem bereits in ATHLET vorhandenen Ansatz für einphasige, turbulente Rohrströmungen bestimmt. Zusätzlich wurde ein Komponentenmodel zur Darstellung radialer Kompressoren implementiert, sowie das in ATHLET vorhandene Modell für axiale Turbinen für die Simulation von Radialturbinen angepasst. Alle durchgeführten Erweiterungen wurden anhand experimenteller Daten validiert. Das neuartige System wurde exemplarisch als potentielle Nachrüstmaßnahme für einen generischen Siedewasserreaktor untersucht, wofür zunächst eine erste Dimensionierung der Hauptkomponenten erfolgte. Der Datensatz eines generischen Siedewasserreaktors wurde anschließend mit dem Nachwärmeabfuhrsystem erweitert und erste Simulationen wurden durchgeführt. Die Ergebnisse haben gezeigt, dass das System die Nachzerfallswärme auch in bestimmten auslegungsüberschreitenden Störfällen wie dem Verlust der Hauptwärmesenke bei gleichzeitigem Station Blackout, für mehr als 72 Stunden abführen kann. Die Auswertung hat weiter ergeben, dass es zu Interaktionen mit bereits existierenden Systemen kommt. Parameter, die den Betrieb des Nachwärmeabfuhrsystems beeinflussen, wurden herausgearbeitet und in einem Anforderungsprofil zusammengefasst. Die Simulationen deuten auf das Potential des Systems hin, in bestehenden Siedewasserreaktoren als diversitäres Nachwärmeabfuhrsystem zu fungieren.