This study explores the behavior of a Boiling Water Reactor (BWR) during a turbine trip scenario initiated by the abrupt closure of the turbine stop valve. The RELAP5/Mod3.4 code is employed to make calculations using the Laguna Verde Nuclear Power Plant input model provided by Innovative Software Systems Company. The event sequences and initial boundary conditions are sourced from the Boiling Water Reactor Turbine Trip 2 Benchmark created by NEA. Results are subsequently compared against the benchmark values. In order to gauge the risk of a turbine trip event leading to elevated power, which could in turn cause Critical Heat Flux (CHF)-related issues in cladding temperature, a best-estimate case is developed. Our findings indicate that the closure of the turbine stop valve (TSV) resulted in a collapse of the void fraction within the reactor core. Although the core power doubled the initial level, the negative feedback mechanism effectively suppressed the power pulse. Throughout the transient phase, the maximum cladding temperature stayed below the CHF threshold, a fact attributable to the fuel's conductivity and the rapid progression of the transient. We further analyzed three hypothetical scenarios to test the computational boundaries of the plant model. The third scenario, which combines conditions from the first two, produced elevated outcomes (6500MW core power, 598K cladding temperature, and 7900kPa dome pressure) as expected. Notably, while the CHF limit remained unbreached in this scenario, literature reviews suggest potential core meltdown risks in subsequent stages of this calculation. Our sensitivity analyses determined that variations in the gamma heating coefficient or the maximum time step of the calculations have little to no impact on core power or peak cladding temperature. Conversely, we noted a significant reduction, approximately 35\%, in the power peak, underscoring the high sensitivity of the parameters to the initial triggering of the SCRAM mechanism. Our results also recommend rapid and early actuation of the BPV as a measure to dampen the pressure wave, consequently decreasing both the power peak and peak cladding temperatures. / Thesis / Master of Applied Science (MASc) / This research investigates the response of the Laguna Verde Boiling Water Reactor to a turbine trip event using the RELAP5/Mod3.4 thermal-hydraulic analysis code. From reactor safety perspective a best-estimate case is evaluated, as well as three additional hypothetical scenarios. Findings are compared with the Boiling Water Reactor Turbine Trip II Benchmark results. Additionally, sensitivity analyses focusing on plant parameters such as shutdown rod behavior, gamma heating coefficient, turbine stop valve, and steam bypass valve characteristics conducted to determine their impact on the results. Insights from these analyses aim to enhance safety protocols and refine best practices in boiling water reactor management.

Fuel channels are used to keep the components of a nuclear fuel bundle in place. The fuel channel of a new nuclear fuel which is being developed at Westinghouse has a complex geometry which creates challenges in the manufacturing process. FE simulations were developed of a two-stage forming process of the fuel channel. Three different simulations were developed, a simulation of the pre-bending, a simplified simulation of bending, and a combined simulation of the two-stage process of pre-bending and bending. The simulations were done in Ansys Workbench. The simulations of the pre-bending could be validated against experimental results. The simulations of the bending showed significant differences in results compared to experiments. A few different sources of error were investigated due to the difference in results. / Höljerör omsluter kärnbränslet i en kärnreaktor och håller komponenterna i bränslepaketet på plats. Höljeröret av ett nytt kärnbränsle som utvecklas på Westinghouse har en komplex geometri som skapar utmaningar i tillverkningsprocessen. FE simuleringar av en tvåstegsprocess av bockningen av höljeröret utvecklades. Tre typer av simuleringar utvecklades, en simulering av förbockningen, en förenklad simulering av bockningen, och en simulering av den kombinerade förbockningen och bockning. Simuleringarna gjordes i Ansys Workbench. Förbockningssimuleringarna kunde valideras mot experiment. Bockningssimuleringen visade dock signifikanta skillnader i resultat gentemot experiment. Några olika felkällor undersöktes på grund av skillnaderna i resultat.

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.

Thesis (M.S.)--University of Wisconsin -- La Crosse, 1983. / Includes bibliographical references (leaves 47-51).

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluated between the FOREVER-experiments and a prototypical scenario. The thermodynamic and the mechanical model can be coupled recursively to take into account the mutual influence. This approach not only allows to consider the tem-perature dependence of the material parameters and the thermally induced stress in the mechanical model, it also takes into account the response of the temperature field itself upon the changing vessel geometry. New approaches are applied in this work for the simulation of creep and damage. Using a creep data base, the application of single creep laws could be avoided which is especially advantageous if large temperature, stress, and strain ranges have to be covered. Based on experimental investigations, the creep data base has been de-veloped for an RPV-steel and has been validated against creep tests with different scalings and geometries. It can be stated, that the coupled model is able to exactly describe and predict the vessel deformation in the scaled integral FOREVER-tests. There are uncertainties concerning the time to failure which are related to inexactly known material parame-ters and boundary conditions. The main results of this work can be summarised as follows: Due to the thermody-namic behaviour of the large melt pool with internal heat sources, the upper third of the lower RPV head is exposed to the highest thermo-mechanical loads. This region is called hot focus. Contrary to that, the pole part of the lower head has a higher strength and therefore relocates almost vertically downwards under the combined thermal, weight and internal pressure load of the RPV. On the one hand, it will be possible by external flooding to retain the corium within the RPV even at increased pressures and even in reactors with high power (as e.g. KONVOI). On the other hand, there is no chance for melt retention in the considered scenario if neither internal nor external flooding of the RPV can be achieved. Two patents have been derived from the gained insights. Both are related to pas-sively working devices for accident mitigation: The first one is a support of the RPV lower head pole part. It reduces the maximum mechanical load in the highly stressed area of the hot focus. In this way, it can prevent failure or at least extend the time to failure of the vessel. The second device implements a passive accident mitigation measure by making use of the downward movement of the lower head. Through this, a valve or a flap can be opened to flood the reactor pit with water from a storage res-ervoir located at a higher position in the reactor building. With regard to future plant designs it can be stated - differing from former presump-tions - that an In-Vessel-Retention (IVR) of a molten core is possible within the reac-tor pressure vessel even for reactors with higher power.

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.

No description available.

Neutron flux fluctuations are a natural phenomenon of nuclear reactors. Approximately since 2001, pressurized water reactors built by Kraftwerk Union AG have exhibited an unexplained cycle-by-cycle change of the magnitude of these fluctuations. The change has also drawn attention to long-known but also unexplained spatial correlations of the fluctuations in these reactors. The thesis at hand aims to contribute to a better understanding of both the observed change in magnitude and the immanent correlations. Based on the findings of previous research and on the own analysis of measured raw data, a hypothesis was developed, which states that a synchronous coolant-flow driven vibration of major parts of the fuel-assembly ensemble triggers the main contribution to the observed neutron flux fluctuations. The fluctuation correlations are supposed to result from the correlations of the fuel-assembly vibration. This idea was tested using the time-domain reactor dynamics code DYN3D and complementary using the frequency-domain neutronic tool CORE SIM. For this purpose, simplified mechanical models of the synchronous fuel-assembly vibration and models coupling the vibration to the neutron kinetics were developed and implemented. Two effects are distinguished: In case of the “reflector effect”, all fuel assemblies vibrate synchronously, in a way that the main resulting perturbation acts in the radial reflector as a fluctuation of the water layer between the outer fuel assemblies and the core shroud. In case of the “fuel-assembly pitch effect”, the fuel assemblies are unequally involved in the synchronous vibration, in a way that the main perturbations are induced within the reactor core as fluctuations of the fuel-assembly gaps. Both the simulations with DYN3D and the simulations with CORE SIM showed that a synchronized fuel-assembly vibration is a possible main source of the concerned neutron flux fluctuations. In particular, the uniform fluctuation of the gaps between all fuel assemblies, corresponding to high-amplitude fuel-assembly vibrations in the core center and low-amplitude fuel-assembly vibrations in the outer core regions, gave the best approximation to the measured data. A C-like axial vibration-shape provides the best agreement. The simulation results show that the developed hypothesis should be further investigated. In particular, the proposed synchronized vibration of the fuel assemblies suggest correlations of the neutron flux fluctuations with mechanical signals, which have not been taken into account so far. The simulations presented here enable further improvements of the understanding of the neutron flux fluctuations in the concerned reactors by additional measurements involving also specific modes of reactor operation.

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluated between the FOREVER-experiments and a prototypical scenario. The thermodynamic and the mechanical model can be coupled recursively to take into account the mutual influence. This approach not only allows to consider the tem-perature dependence of the material parameters and the thermally induced stress in the mechanical model, it also takes into account the response of the temperature field itself upon the changing vessel geometry. New approaches are applied in this work for the simulation of creep and damage. Using a creep data base, the application of single creep laws could be avoided which is especially advantageous if large temperature, stress, and strain ranges have to be covered. Based on experimental investigations, the creep data base has been de-veloped for an RPV-steel and has been validated against creep tests with different scalings and geometries. It can be stated, that the coupled model is able to exactly describe and predict the vessel deformation in the scaled integral FOREVER-tests. There are uncertainties concerning the time to failure which are related to inexactly known material parame-ters and boundary conditions. The main results of this work can be summarised as follows: Due to the thermody-namic behaviour of the large melt pool with internal heat sources, the upper third of the lower RPV head is exposed to the highest thermo-mechanical loads. This region is called hot focus. Contrary to that, the pole part of the lower head has a higher strength and therefore relocates almost vertically downwards under the combined thermal, weight and internal pressure load of the RPV. On the one hand, it will be possible by external flooding to retain the corium within the RPV even at increased pressures and even in reactors with high power (as e.g. KONVOI). On the other hand, there is no chance for melt retention in the considered scenario if neither internal nor external flooding of the RPV can be achieved. Two patents have been derived from the gained insights. Both are related to pas-sively working devices for accident mitigation: The first one is a support of the RPV lower head pole part. It reduces the maximum mechanical load in the highly stressed area of the hot focus. In this way, it can prevent failure or at least extend the time to failure of the vessel. The second device implements a passive accident mitigation measure by making use of the downward movement of the lower head. Through this, a valve or a flap can be opened to flood the reactor pit with water from a storage res-ervoir located at a higher position in the reactor building. With regard to future plant designs it can be stated - differing from former presump-tions - that an In-Vessel-Retention (IVR) of a molten core is possible within the reac-tor pressure vessel even for reactors with higher power.

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.