The steam generator plays a primary role in the safety and performance of a pressurized water reactor nuclear power plant. The cost to utilities is in the order of millions of Rands a year as a direct result of damage to steam generators. The damage results in lower efficiency or even plant shutdown. It is necessary for the utility and for academia to have models of nuclear components by which research and analysis may be performed. It must be possible to analyse steam generator performance for both day-to-day operational analysis as well as in the case of extreme accident scenarios. The homogeneous model for two-phase flow is simpler in its implementation than the two-fluid model, and therefore suffers in accuracy. Its advantage lies in its quick turnover time for development of models and subsequent analysis. It is often beneficial for a modeller to be able to quickly set up and analyse a model of a system, and a trade-off between accuracy and time-management is thus required. Searches through available literature failed to provide answers to how the homogeneous model compares with the two-fluid model for operational and safety analysis. It is expected to see variations between the models, from the analysis of the mathematics, but it remains to be shown what these differences are. The purpose of this study was to determine how the homogeneous model for two-phase flow compares with the two-fluid model when applied to a u-tube steam generator of a typical pressurized water reactor. The steam generator was modelled in both RELAP5 and in Flownex. A custom script was written for Flownex in order to implement the Chen correlation for boiling heat transfer. This was significantly less detailed than RELAP5’s solution of a matrix of flow regimes and heat transfer correlations. The geometry of the models were based on technical drawings from Koeberg Nuclear Power Plant, and were simplified to a one-dimensional model. Plant data obtained from Koeberg was used to validate the models at 100%, 80% and 60% power output. It was found that the overall heat transfer rate predicted with the RELAP5 two-fluid model was within 1.5% of the measured data from the Koeberg plant. The results generated by the homogeneous model for the overall heat transfer were within 4.5% of the measured values. However, the differences in the detailed temperature distributions and heat transfer coefficient values were quite significant at the inlet and outlet ends of the tube bundle, at the bottom tube sheet of the steam generator. In this area the water-level was not accurately modelled by the homogeneous model, and therefore there was an under-prediction in heat transfer in that region. Large differences arose between the Flownex and RELAP5 solutions due to difference in the heat transfer correlations used. The Flownex model exclusively implemented the Chen correlation, while RELAP5 implements a flow regime map correlated to a table of heat transfer correlations. It was concluded that the results from the homogeneous model for two-phase flow do not differ significantly when compared with the two-fluid model when applied to the u-tube steam generator at the normal operating conditions. Significant differences do, however, occur in lower regions of the boiler where the quality is lower. We conclude that the homogeneous model offers significant advantage in simplicity over the two-fluid model for normal operational analysis. This may not be the case for detailed accident analysis, which was beyond the scope of this study. / Thesis (MIng (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2013.

The steam generator plays a primary role in the safety and performance of a pressurized water reactor nuclear power plant. The cost to utilities is in the order of millions of Rands a year as a direct result of damage to steam generators. The damage results in lower efficiency or even plant shutdown. It is necessary for the utility and for academia to have models of nuclear components by which research and analysis may be performed. It must be possible to analyse steam generator performance for both day-to-day operational analysis as well as in the case of extreme accident scenarios. The homogeneous model for two-phase flow is simpler in its implementation than the two-fluid model, and therefore suffers in accuracy. Its advantage lies in its quick turnover time for development of models and subsequent analysis. It is often beneficial for a modeller to be able to quickly set up and analyse a model of a system, and a trade-off between accuracy and time-management is thus required. Searches through available literature failed to provide answers to how the homogeneous model compares with the two-fluid model for operational and safety analysis. It is expected to see variations between the models, from the analysis of the mathematics, but it remains to be shown what these differences are. The purpose of this study was to determine how the homogeneous model for two-phase flow compares with the two-fluid model when applied to a u-tube steam generator of a typical pressurized water reactor. The steam generator was modelled in both RELAP5 and in Flownex. A custom script was written for Flownex in order to implement the Chen correlation for boiling heat transfer. This was significantly less detailed than RELAP5’s solution of a matrix of flow regimes and heat transfer correlations. The geometry of the models were based on technical drawings from Koeberg Nuclear Power Plant, and were simplified to a one-dimensional model. Plant data obtained from Koeberg was used to validate the models at 100%, 80% and 60% power output. It was found that the overall heat transfer rate predicted with the RELAP5 two-fluid model was within 1.5% of the measured data from the Koeberg plant. The results generated by the homogeneous model for the overall heat transfer were within 4.5% of the measured values. However, the differences in the detailed temperature distributions and heat transfer coefficient values were quite significant at the inlet and outlet ends of the tube bundle, at the bottom tube sheet of the steam generator. In this area the water-level was not accurately modelled by the homogeneous model, and therefore there was an under-prediction in heat transfer in that region. Large differences arose between the Flownex and RELAP5 solutions due to difference in the heat transfer correlations used. The Flownex model exclusively implemented the Chen correlation, while RELAP5 implements a flow regime map correlated to a table of heat transfer correlations. It was concluded that the results from the homogeneous model for two-phase flow do not differ significantly when compared with the two-fluid model when applied to the u-tube steam generator at the normal operating conditions. Significant differences do, however, occur in lower regions of the boiler where the quality is lower. We conclude that the homogeneous model offers significant advantage in simplicity over the two-fluid model for normal operational analysis. This may not be the case for detailed accident analysis, which was beyond the scope of this study. / Thesis (MIng (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2013.

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.

Innerhalb des Vorhabens wurde ein ATHLET-CD-Eingabedatensatz für einen generischen deutschen DWR vom Typ KONVOI entwickelt. Das ATHLET-CD-Modell wurde für die Simulation schwerer Störfälle aus den Störfallkategorien Station Blackout (SBO) und Kühlmittelverluststörfällen mit kleinen Lecks (SBLOCA) eingesetzt. Dabei ist die vollständige Störfalltransiente für den Zeitbereich zwischen dem einleitenden Ereignis bis zum Versagen des Reaktordruckbehälters (RDB) abgedeckt und alle wesentli-chen Phänomene schwerer Störfällen werden abgebildet: Beginn der Kernaufheizung, Spaltproduktfrei-setzung, Aufschmelzen von Brennstoff- und Absorbermaterialien, Oxidationsprozesse mit Freisetzung von Wasserstoff, Verlagerung von geschmolzenem Material, Verlagerung in das untere Plenum, Schä-digung und Versagen des RDB. Das Modell wurde für die Analyse möglicher präventiver und mitigativer Notfallmaßnahmen für SBO und SBLOCA angewandt. Dafür wurden die Notfallmaßnahmen primärseitige Druckentlastung (PDE), primärseitiges Einspeisen mit mobilen Pumpensystemen sowie für SBLOCA das verzögerte Einspeisen der kaltseitigen Druckspeicher untersucht und die Eigenschaften und Einleitekriterien der Maßnahmen variiert. Es wurden die Zeitverläufe der Unfallszenarien analysiert und die verbleibenden Zeitspannen für die Einleitung zusätzlicher Maßnahmen ermittelt. Für ein SBO-Szenario mit PDE wurde für die Frühphase der Transiente (bis zum Beginn der Kernschmelze) eine Unsicherheits- und Sensititvitätsanalyse durchgeführt. Zusätzlich wurde für ein SBLOCA-Szenario ein Code-zu-Code-Vergleich zwischen ATHLET-CD und dem Störfallcode MELCOR erarbeitet.

Es wurde ein Modell zur realistischen Beschreibung der Kühlmittelvermischung innerhalb des Reaktordruckbehälters von Druckwasserreaktoren in den gekoppelten Programmkomplex DYN3D/ATHLET implementiert. Diese Modell basiert auf dem Prinzip der linearen Superposition Dirac-Impuls-ähnlicher Störungen und kann für die Modellierung der Vermischung von Kühlmittel unterschiedlicher Temperatur und/oder unterschiedlicher Borsäurekonzentration eingesetzt werden. Der um das Vermischungsmodell erweiterte Programmkomplex DYN3D/ATHLET wurde für Analyse von Borverdünnungsstörfällen und Frischdampflecks angewandt. Für den Fall "Start der ersten Hauptkühlmittelpumpe bei Vorhandensein eines minderborierten Pfropfens im kalten Strang" zeigten die Ergebnisse der durchgeführten Parameterstudie, dass es selbst bei Annahme des maximal möglichen Pfropfenvolumens nicht zu einer Schädigung des Brennstoffes kommt. Mit den Analysen zu einem generischen Frischdampfleckszenario wurde die Anwendbarkeit des Programmkomplexes DYN3D/ATHLET auf die zweite Störfallklasse, in der die Kühlmittelvermischung eine wichtige Rollen spielt, demonstriert. Im Rahmen der Arbeiten zum Projekt wurde außerdem gezeigt, dass der Einfluss der turbulenten Schwankungen des Geschwindigkeitsfeldes innerhalb des Reaktordruckbehälters auf neutronenkinetische Parameter im Nominalbetrieb und unter Störfallbedingungen nicht zu vernachlässigen ist. A model for the realistic description of the coolant mixing inside the pressure vessel of pressurized water reactors was implemented into the coupled code complex DYN3D/ATHLET. This model is based on the linear superposition Dirac-pulse-like perturbations. The model can be applied to the mixing of coolant of different temperature and/or boron concentration. The coupled code complex DYN3D/ATHLET with the newly implemented model was applied to the analysis of boron dilution and steam line break accidents. The results of a parameter study for the case "Start-up of the first main coolant pump with a slug of lower borated water in the cold leg" have shown, that even under the conditions of the maximum slug volume there is no fuel damage. The applicability to the second class of accidents, where the coolant mixing has to be considered, was demonstrated by the analysis of a generic main steam line break scenario. Further it was shown, that the influence of turbulent fluctuations of the velocity inside the reactor pressure vessel during nominal and accident conditions on neutron-kinetic parameters cannot be neglected.

Es wurde ein Modell zur realistischen Beschreibung der Kühlmittelvermischung innerhalb des Reaktordruckbehälters von Druckwasserreaktoren in den gekoppelten Programmkomplex DYN3D/ATHLET implementiert. Diese Modell basiert auf dem Prinzip der linearen Superposition Dirac-Impuls-ähnlicher Störungen und kann für die Modellierung der Vermischung von Kühlmittel unterschiedlicher Temperatur und/oder unterschiedlicher Borsäurekonzentration eingesetzt werden. Der um das Vermischungsmodell erweiterte Programmkomplex DYN3D/ATHLET wurde für Analyse von Borverdünnungsstörfällen und Frischdampflecks angewandt. Für den Fall "Start der ersten Hauptkühlmittelpumpe bei Vorhandensein eines minderborierten Pfropfens im kalten Strang" zeigten die Ergebnisse der durchgeführten Parameterstudie, dass es selbst bei Annahme des maximal möglichen Pfropfenvolumens nicht zu einer Schädigung des Brennstoffes kommt. Mit den Analysen zu einem generischen Frischdampfleckszenario wurde die Anwendbarkeit des Programmkomplexes DYN3D/ATHLET auf die zweite Störfallklasse, in der die Kühlmittelvermischung eine wichtige Rollen spielt, demonstriert. Im Rahmen der Arbeiten zum Projekt wurde außerdem gezeigt, dass der Einfluss der turbulenten Schwankungen des Geschwindigkeitsfeldes innerhalb des Reaktordruckbehälters auf neutronenkinetische Parameter im Nominalbetrieb und unter Störfallbedingungen nicht zu vernachlässigen ist. A model for the realistic description of the coolant mixing inside the pressure vessel of pressurized water reactors was implemented into the coupled code complex DYN3D/ATHLET. This model is based on the linear superposition Dirac-pulse-like perturbations. The model can be applied to the mixing of coolant of different temperature and/or boron concentration. The coupled code complex DYN3D/ATHLET with the newly implemented model was applied to the analysis of boron dilution and steam line break accidents. The results of a parameter study for the case "Start-up of the first main coolant pump with a slug of lower borated water in the cold leg" have shown, that even under the conditions of the maximum slug volume there is no fuel damage. The applicability to the second class of accidents, where the coolant mixing has to be considered, was demonstrated by the analysis of a generic main steam line break scenario. Further it was shown, that the influence of turbulent fluctuations of the velocity inside the reactor pressure vessel during nominal and accident conditions on neutron-kinetic parameters cannot be neglected.

Innerhalb des Vorhabens wurde ein ATHLET-CD-Eingabedatensatz für einen generischen deutschen DWR vom Typ KONVOI entwickelt. Das ATHLET-CD-Modell wurde für die Simulation schwerer Störfälle aus den Störfallkategorien Station Blackout (SBO) und Kühlmittelverluststörfällen mit kleinen Lecks (SBLOCA) eingesetzt. Dabei ist die vollständige Störfalltransiente für den Zeitbereich zwischen dem einleitenden Ereignis bis zum Versagen des Reaktordruckbehälters (RDB) abgedeckt und alle wesentli-chen Phänomene schwerer Störfällen werden abgebildet: Beginn der Kernaufheizung, Spaltproduktfrei-setzung, Aufschmelzen von Brennstoff- und Absorbermaterialien, Oxidationsprozesse mit Freisetzung von Wasserstoff, Verlagerung von geschmolzenem Material, Verlagerung in das untere Plenum, Schä-digung und Versagen des RDB. Das Modell wurde für die Analyse möglicher präventiver und mitigativer Notfallmaßnahmen für SBO und SBLOCA angewandt. Dafür wurden die Notfallmaßnahmen primärseitige Druckentlastung (PDE), primärseitiges Einspeisen mit mobilen Pumpensystemen sowie für SBLOCA das verzögerte Einspeisen der kaltseitigen Druckspeicher untersucht und die Eigenschaften und Einleitekriterien der Maßnahmen variiert. Es wurden die Zeitverläufe der Unfallszenarien analysiert und die verbleibenden Zeitspannen für die Einleitung zusätzlicher Maßnahmen ermittelt. Für ein SBO-Szenario mit PDE wurde für die Frühphase der Transiente (bis zum Beginn der Kernschmelze) eine Unsicherheits- und Sensititvitätsanalyse durchgeführt. Zusätzlich wurde für ein SBLOCA-Szenario ein Code-zu-Code-Vergleich zwischen ATHLET-CD und dem Störfallcode MELCOR erarbeitet.

This thesis studied the performance of the thermal hydraulic software TRACE applied on an intermediate sized break (IB) happening on the cold leg in a pressurized water reactor (PWR), causing a loss-of-coolant accident (LOCA). The same accident has previously been simulated in the PKL Test Facility, which is a scaled version of a PWR and is used to simulate transients stemming from different accidents. The thesis was performed as a semi-blind study: firstly, the accident was simulated without any knowledge of the PKL results. When a final blind model was chosen, the PKL results were revealed, and the TRACE model was improved. Before the simulations of the IB-LOCA took place, the new internal parts in the upper parts of the reactor pressure vessel in PKL had to be modelled, and the steady state had to be tuned to attain the correct initial conditions. The simulations were performed by using the software SNAP together with TRACE, providing a graphical interface. TRACE achieved steady state with satisfying results regarding water levels, pressure losses and mass flows. The temperatures in TRACE deviated from the PKL temperatures but an explanation is uncertainties in PKL. To verify TRACE’s core output power, the calculation of the power was done by using mass flow rate and specific entropy and comparing to the heaters’ specified power. This resulted in lower output power meaning that the coolant was not heated enough. This indicated non-physical energy losses in the TRACE model and should be further investigated.The blind transient simulation, modelled with default choked flow and no offtake model, resulted in TRACE overestimating the break mass flow and the peak cladding temperatures, compared to the PKL reference solution. This resulted in the pressure decreasing too quickly and too early activation of the safety system. The modified simulations showed that it is important that the offtake model, which accounts for different flow regimes, is activated. Default choked flow multipliers were the multipliers that performed the best. However, none of the transient simulations could be completed due to fatal errors and memory problems, but some conclusions could be drawn from the observed trends. This concluded in the offtake model being most important due to stratified flow occurring.

Certains éléments des internes de cuve des réacteurs à eau pressurisée (REP) en acier inoxydable austénitique 316L, présentent un endommagement prématuré par corrosion sous contrainte assistée par l’irradiation. Ce phénomène est complexe puisqu’il implique le couplage entre le matériau lui-même, l'état de contrainte, l’irradiation et l’environnement. L’irradiation neutronique couplée à un facteur environnemental, l’hydrogène, pourrait jouer un rôle dans ce phénomène. Ainsi, les travaux présentés dans ce manuscrit porte sur l’étude des effets des défauts induits par l’irradiation sur le piégeage de l’hydrogène dans un acier inoxydable austénitique 316L au cours de son oxydation en milieu primaire des REP. Grâce à des implantations ioniques permettant de reproduire le même type de défauts que ceux induits par les neutrons, les interactions hydrogène - défauts ont tout d’abord été étudié avec une approche modèle par chargement cathodique en deutérium, traceur isotopique de l’hydrogène. Cela a permis de mettre en évidence le piégeage de cet élément au niveau des défauts induits par l’implantation ionique et particulièrement des cavités. Un modèle de résolution numérique des équations de McNabb et Foster permettant de simuler la diffusion et le piégeage de l’hydrogène dans un matériau a confirmé ce résultat. Ensuite des essais d’oxydation en milieu primaire simulé des REP à 320 °C ont été réalisés afin de comparer les couches d’oxyde formées entre matériaux implantés et non implantés, ainsi que leur prise d’hydrogène. Ces essais ont permis de mettre en avant le piégeage de l’hydrogène dans l’alliage sous l’oxyde et, dans une moindre mesure, au niveau des défauts d’implantation plus en profondeur. / Some components of vessel internals in pressurized water reactor (PWR) made of stainless steel, have shown cracks induced by Irradiation-Assisted Stress Corrosion Cracking (IASCC). This complex phenomenon originates from the coupling between the material itself, a tensile stress state, environmental conditions and irradiation. This PhD thesis aims at studying the influence of hydrogen in IASCC and particularly its interactions with the defects created by neutron irradiation in a 316L austenitic stainless steel. Thanks to ion implantation, defects similar to neutron irradiation-induced defects were created. As a first step, hydrogen - defects interactions were studied with a model approach consisting in deuterium cathodic charging. Deuterium was used as an isotopic tracer for hydrogen. This technique allowed to highlight hydrogen trapping by implantation-induced defects (mostly by the cavities) in 316L stainless steel. Simulation of hydrogen diffusion and trapping in the studied materials with a numerical resolution model of McNabb and Foster’s equations confirmed the experimental results. Then, oxidation tests were performed in PWR simulated primary environment at 320 °C in order to study the effects of irradiation-induced defects on the oxidation and the hydrogen uptake of the 316L stainless steel. The major highlight of these experiments was the observation of hydrogen accumulation in the alloy beneath the oxide, due to trapping by vacancies created by oxidation process and by ion implantation. In addition, hydrogen trapping was observed deeper in the alloy and it was attributed to the cavities induced by implantation.

Orientador: João Manoel Losada Moreira / Dissertação (mestrado) - Universidade Federal do ABC, Programa de Pós-Graduação em Energia, 2011