The accident at the Fukushima Daiichi nuclear power plant has shown us that there may be situations where the applied technology will not be able to successfully cool the reactor core. These situations may occur when more elements such as supply of energy to power the pumps and diesel generators are destroyed for example by tsunami or earthquake, or other not expected natural disasters. The inability of the residual heat removal leads to the melting of core, relocation to the bottom of reactor pressure vessel (RPV) and failure of RPV. Result of this accident may be containment failure and leakage of fission products into the environment. One way to prevent this scenario may be a passive system called IVR (In-Vessel Retention) by using external cooling of RPV that retains melted core in. This system counts with flooding of RPV´s shaft by water. After natural circulation of water provides the heat transfers from the wall of RPV. The applicability of IVR for VVER 1000 reactors is still in the course of research. However it´s already clear that the submersion of RPV shaft by water will not sufficient. Other elements as suitable insulation and RPV coating which provides a more intensive heat transfer from the walls of RPV will be needed.

Accident conditions involving significant core degradation are termed severe accidents /IAEA: NS-G-2.15/. Despite the low probability of occurrence of such events, the investigation of severe accident scenarios is an important part of the nuclear safety research. Considering a hypothetical core melt down scenario in a VVER-1000 light water reactor, the early in-vessel phase focusing on the thermal-hydraulic phenomena, and the late in-vessel phase focusing on the melt relocation into the reactor pressure vessel (RPV) lower head, are investigated. The objective of this work is the assessment of severe accident management procedures for VVER-1000 reactors, i.e. the estimation of the maximum period of time available for taking appropriate measures and particular decisions by the plant personnel. During high pressure severe accident sequences it is of prime importance to depressurize the primary circuit in order to allow for effective injection from the emergency core cooling systems and to avoid reactor pressure vessel failure at high pressure that could cause direct containment heating and subsequent challenge to the containment structure. Therefore different accident management measures were investigated for the in-vessel phase of a hypothetical station blackout accident using the severe accident code ASTEC, the mechanistic code ATHLET and the multi-purpose code system ANSYS. The analyses performed on the PHEBUS ISP-46 experiment, as well as simulations of small break loss of coolant accident and station blackout scenarios were used to contribute to the validation and improvement of the integral severe accident code ASTEC. Investigations on the applicability and the effectiveness of accident management procedures in the preventive domain, as well as detailed analyses on the thermal-hydraulic phenomena during the early in-vessel phase of a station blackout accident have been performed with the mechanistic code ATHLET. The results of the simulations show, that the effectiveness of the procedures strongly depends on the ability of the passive safety systems to inject as much water as possible into the reactor coolant system. The results on the early in-vessel phase have shown potentially delayed RPV failure by depressurization of the primary side, as slowing the core damage gives more time and different possibilities for operator interventions to recover systems and to mitigate or terminate the accident. The ANSYS model for the description of the molten pool behaviour in the RPV lower plenum has been extended by a model considering a stratified molten pool configuration. Two different pool configurations were analysed: homogeneous and segregated. The possible failure modes of the RPV and the time to failure were investigated to assess the possible loadings on the containment. The main treated issues are: the temperature field within the corium pool and the RPV and the structure-mechanical behaviour of the vessel wall. The results of the ASTEC calculations of the melt pool configuration were applied as initial conditions for the ANSYS simulations, allowing a more detailed and more accurate modelling of the thermal and mechanical behaviour of the core melt and the RPV wall. Moreover, for the late in-vessel phase, retention of the corium in the RPV was investigated presuming external cooling of the vessel wall as mitigative severe accident management measure. The study was based on the finite element computer code ANSYS. The highest thermomechanical loads are observed in the transition zone between the elliptical and the vertical vessel wall for homogeneous pool and in the vertical part of the vessel wall, which is in contact with the molten metal in case of sub-oxidized pool. Assuming external flooding will retain the corium within the RPV. Without flooding, the vessel wall will fail, as the necessary temperature for a balanced heat release from the external surface via radiation is near to or above the melting point of the steel.

Accident conditions involving significant core degradation are termed severe accidents /IAEA: NS-G-2.15/. Despite the low probability of occurrence of such events, the investigation of severe accident scenarios is an important part of the nuclear safety research. Considering a hypothetical core melt down scenario in a VVER-1000 light water reactor, the early in-vessel phase focusing on the thermal-hydraulic phenomena, and the late in-vessel phase focusing on the melt relocation into the reactor pressure vessel (RPV) lower head, are investigated. The objective of this work is the assessment of severe accident management procedures for VVER-1000 reactors, i.e. the estimation of the maximum period of time available for taking appropriate measures and particular decisions by the plant personnel. During high pressure severe accident sequences it is of prime importance to depressurize the primary circuit in order to allow for effective injection from the emergency core cooling systems and to avoid reactor pressure vessel failure at high pressure that could cause direct containment heating and subsequent challenge to the containment structure. Therefore different accident management measures were investigated for the in-vessel phase of a hypothetical station blackout accident using the severe accident code ASTEC, the mechanistic code ATHLET and the multi-purpose code system ANSYS. The analyses performed on the PHEBUS ISP-46 experiment, as well as simulations of small break loss of coolant accident and station blackout scenarios were used to contribute to the validation and improvement of the integral severe accident code ASTEC. Investigations on the applicability and the effectiveness of accident management procedures in the preventive domain, as well as detailed analyses on the thermal-hydraulic phenomena during the early in-vessel phase of a station blackout accident have been performed with the mechanistic code ATHLET. The results of the simulations show, that the effectiveness of the procedures strongly depends on the ability of the passive safety systems to inject as much water as possible into the reactor coolant system. The results on the early in-vessel phase have shown potentially delayed RPV failure by depressurization of the primary side, as slowing the core damage gives more time and different possibilities for operator interventions to recover systems and to mitigate or terminate the accident. The ANSYS model for the description of the molten pool behaviour in the RPV lower plenum has been extended by a model considering a stratified molten pool configuration. Two different pool configurations were analysed: homogeneous and segregated. The possible failure modes of the RPV and the time to failure were investigated to assess the possible loadings on the containment. The main treated issues are: the temperature field within the corium pool and the RPV and the structure-mechanical behaviour of the vessel wall. The results of the ASTEC calculations of the melt pool configuration were applied as initial conditions for the ANSYS simulations, allowing a more detailed and more accurate modelling of the thermal and mechanical behaviour of the core melt and the RPV wall. Moreover, for the late in-vessel phase, retention of the corium in the RPV was investigated presuming external cooling of the vessel wall as mitigative severe accident management measure. The study was based on the finite element computer code ANSYS. The highest thermomechanical loads are observed in the transition zone between the elliptical and the vertical vessel wall for homogeneous pool and in the vertical part of the vessel wall, which is in contact with the molten metal in case of sub-oxidized pool. Assuming external flooding will retain the corium within the RPV. Without flooding, the vessel wall will fail, as the necessary temperature for a balanced heat release from the external surface via radiation is near to or above the melting point of the steel.

This paper describes the analytical work done by modelling and evaluating a thermal shock in a WWER-440 reactor pressure vessel due to an emergency case. An axial oriented semielliptical underclad/surface crack is assumed to be located in the core weld line. Threedimensional finite element models are used to compute the global transient temperature and stress-strain fields. By using a three-dimensional submodel, which includes the crack, the local crack stress-strain field is obtained. With a subsequent postprocessing using the j-integral technique the stress intensity factors KI along the crack front are obtained. The results for the underclad and surface crack are provided and compared, together with a critical discussion of the VERLIFE code.

This master’s thesis deals with boiling crisis and with departure from nucleate boiling ratio. This thesis explains terms like the boiling crisis in nuclear reactors and the thesis deals with individual parameters of the reactor core, which have an impact on departure from nucleate boiling ratio. After that, the thesis deals with subchannel analysis and describes basic mathematical and physical models of the chosen subchannel program. The thesis then processes, with the ALTHAMC12 subchannel program, the chosen parameters and their impact on departure from nucleate boiling ratio. The conclusion of the diploma thesis deals with the evaluation of the best and worst calculated combination.

This master thesis deals with the possibilities of traffic of nuclear power unit at thermal and power effect at the end of the campaign, focusing on VVER reactors. For a better idea of the reader the design of key components of the unit in terms of performance is analysed. Parameters of relevant components for Dukovany NPP are presented briefly. The possibilities of traffic of nuclear power unit on thermal and power effect at the end of the campaign are particularly demonstrated on the example of the Dukovany NPP. Furthermore the program Moby-Dick is introduced and the basic possibilities for its use to calculate the course of the campaign are described. At the end of the thesis, we conducted sample calculations for the duration of the campaign on the fourth block of the nuclear power plant.

This paper describes the analytical work done by modelling and evaluating a thermal shock in a WWER-440 reactor pressure vessel due to an emergency case. An axial oriented semielliptical underclad/surface crack is assumed to be located in the core weld line. Threedimensional finite element models are used to compute the global transient temperature and stress-strain fields. By using a three-dimensional submodel, which includes the crack, the local crack stress-strain field is obtained. With a subsequent postprocessing using the j-integral technique the stress intensity factors KI along the crack front are obtained. The results for the underclad and surface crack are provided and compared, together with a critical discussion of the VERLIFE code.

The submitted master thesis deals with finding the right combination of models describing the cladding behavior implemented in the FEMAXI-6 computational code and then comparing it with the benchmark Zaporoshye, Novovoronezh and the modified FERMAXI 6 program with implemented models describing the E110 behavior used for VVER-type reactor fuel rods. Initially, there is a description of the FEMAXI-6 nuclear fuel analysis program including its structure, calculation mechanics and input file description. Furthermore, the work presents the benchmarks used to evaluate individual combinations of fuel models. An important part is the description of the PIE measurement, including measurement uncertainty, which can affect the results of the work. The next chapter contains a description of what affects fuel behavior during irradiation, which can be mathematically expressed and modeled. The following chapter describes the material equations defining the E110 alloy used as cladding of the fuel rods and which were subsequently implemented into the FEMAXI-6 computing program. The last chapter is devoted to the description of the results with appropriate comments.

This graduation thesis deals with evaluation of power and coolant flow in reactor core. The first part is a description of nuclear power plant VVER 440. It is focused on parts important for transfer and utilize energy in regular operating of generating block. In the second part, the equations for calculation of power and coolant flow in reactor core are deduced. The last part is about designing the program for calculation of published values. There are specified requirements for the program and on the basis of this the source code is written. The parts of code are described. In conclusion of this part, the user's manual is work out. The program is on CD in the annexe.

Gegenstand des Vorhabens im Rahmen der WTZ mit Russland ist die Versprödung des Reaktordruckbehälters infolge der Strahlenbelastung mit schnellen Neutronen im kernnahen Bereich. Um den Einfluss von bestrahlungsinduzierten Gitterdefekten auf die mechanischen Eigenschaften zu ermitteln, wurden analytische Berechnungen zum Einfluss von Hindernissen auf die Beweglichkeit von Versetzungen und damit auf die Ausbildung einer plastischen Zone an der Rissspitze durchgeführt. Es wird demonstriert, dass sich die an der Rissspitze entstehenden Versetzungen an dem Hindernis (bestrahlungsinduzierte Punktdefekte) aufstauen. In Abhängigkeit der Rissbelastung KI und der Entfernung des Hindernisses von der Rissspitze werden die Versetzungsdichte und das durch den Versetzungsstau verursachte Spannungsfeld berechnet. Mit Hilfe von Experimenten zur Neutronenkleinwinkelstreuung (SANS - small angle neutron scattering) an verschiedenen WWER-Stählen und Modelllegierungen wurden Größenverteilungen und die Volumenanteile der strahleninduzierten Defekte für verschiedene Bestrahlungszustände (Fluenzen, Bestrahlungstemperaturen) ermittelt. Es wurde gezeigt, dass sich die strahleninduzierte Werkstoffschädigung durch Wärmebehandlung weitgehend wieder ausheilen lässt. Nach der thermischen Ausheilung ist der Werkstoff bei erneuter Bestrahlung weniger anfällig für strahleninduzierte Defekte. Die Ergebnisse der SANS-Untersuchungen wurden mit der Änderung der mechanischen Eigenschaften (Härte, Streckgrenze und Sprödbruchübergangstemperatur) korreliert. Mit der kinetischen Gitter-Monte-Carlo-Methode wurden numerische Sensitivitätsstudien zum Einfluss des Cu-Gehalts auf die Stabilität von Defekt-Clustern durchgeführt. Die Berechnungen zeigen, dass die Anwesenheit von Cu-Atomen zur Bildung von langlebigen Defekten führt. Dabei werden Leerstellen in Cu/Leerstellen-Cluster eingefangen. Leerstellen in reinem Eisen sind bei Bestrahlungstemperaturen von 270 °C dagegen nicht stabil, die Lebensdauer liegt zwischen 0.01 s und 1 s. Die kritische Cu-Konzentration, ab welcher stabile Defekte entstehen, beträgt ca. 0.1 Masseprozent.