This study was based on the uncertainty and sensitivity analysis of a generic 10 MW Materials Test Reactor (MTR). In this study an uncertainty and sensitivity analysis methodology called code scaling applicability and uncertainty (CSAU) was implemented. Although this methodology follows 14 steps, only the following were carried out: scenario specification, nuclear power plant (NPP) selection, phenomena identification and ranking table (PIRT), selection of frozen code, provision of code documentation, determination of code applicability, determination of code and experiment accuracy, NPP sensitivity analysis calculations, combination of biases and uncertainties, and total uncertainty to calculate specific scenario in a specific NPP. The thermal hydraulic code Flownex®1 was used to model only the reactor core to investigate the effects of the input parameters on the selected output parameters of the hot channel in the core. These output parameters were mass flow rate, temperature of the coolant, outlet pressure, centreline temperature of the fuel and surface temperature of the cladding. The PIRT process was used in conjunction with the sensitivity analysis results in order to select the relevant input parameters that significantly influenced the selected output parameters. The input parameters that have the largest effect on the selected output parameters were found to be the coolant flow channel width between the plates in the hot channel, the width of the fuel plates itself in the hot channel, the heat generation in the fuel plate of the hot channel, the global mass flow rate, the global coolant inlet temperature, the coolant flow channel width between the plates in the cold channel, and the width of the fuel plates in the cold channel. The uncertainty of input parameters was then propagated in Flownex using the Monte Carlo based uncertainty analysis function. From these results, the corresponding probability density function (PDF) of each selected output parameter was constructed. These functions were found to follow a normal distribution. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2014

With the ever advancing state of computer systems, it is imperative to maintain the most up-to-date and reliable safety evaluation data for nuclear power systems. Commonplace now is the practice of updating old accident simulation results with more advanced models and codes using today's faster computer systems. Though it may be quite an undertaking, the benefits of using a more advanced model and code can be significant especially if the result of the new analysis provides increased safety margin for any plant component or system. A series of parametric and sensitivity studies for the Loss of Normal Feedwater Anticipated Transient without Scram (LONF ATWS) for Southern Company's Vogtle Electric Generating Plant (VEGP) Units 1&2 located near Waynesboro, GA was performed using the best-estimate thermal-hydraulics transient analysis code RETRAN-02w. This thesis includes comparison to the results of a generic plant study published by Westinghouse Electric Corporation in 1974 using an earlier code, LOFTRAN, as well as Vogtle-specific analysis. The comparative analysis exposes and seeks to explain differences between the two codes whereas the Vogtle analysis utilizes data from the Vogtle FSAR to generate plant-specific data. The purpose of this study is to validate and update the previous analysis and gather more information about the plant actions taken in response to a LONF ATWS. As a result, now there is a new and updated evaluation of the LONF ATWS for both a generic 4-loop Westinghouse plant and VEGP using a more advanced code. Beyond the reference case analysis, a series of sensitivity and parametric studies have been performed to show how well each type of plant is designed for handling an ATWS situation. These studies cover a wide range of operating conditions to demonstrate the dependability of the model. It was found that both the generic 4-loop Westinghouse PWR system and VEGP can successfully mitigate a LONF ATWS throughout the core's operating cycle.

This study was based on the uncertainty and sensitivity analysis of a generic 10 MW Materials Test Reactor (MTR). In this study an uncertainty and sensitivity analysis methodology called code scaling applicability and uncertainty (CSAU) was implemented. Although this methodology follows 14 steps, only the following were carried out: scenario specification, nuclear power plant (NPP) selection, phenomena identification and ranking table (PIRT), selection of frozen code, provision of code documentation, determination of code applicability, determination of code and experiment accuracy, NPP sensitivity analysis calculations, combination of biases and uncertainties, and total uncertainty to calculate specific scenario in a specific NPP. The thermal hydraulic code Flownex®1 was used to model only the reactor core to investigate the effects of the input parameters on the selected output parameters of the hot channel in the core. These output parameters were mass flow rate, temperature of the coolant, outlet pressure, centreline temperature of the fuel and surface temperature of the cladding. The PIRT process was used in conjunction with the sensitivity analysis results in order to select the relevant input parameters that significantly influenced the selected output parameters. The input parameters that have the largest effect on the selected output parameters were found to be the coolant flow channel width between the plates in the hot channel, the width of the fuel plates itself in the hot channel, the heat generation in the fuel plate of the hot channel, the global mass flow rate, the global coolant inlet temperature, the coolant flow channel width between the plates in the cold channel, and the width of the fuel plates in the cold channel. The uncertainty of input parameters was then propagated in Flownex using the Monte Carlo based uncertainty analysis function. From these results, the corresponding probability density function (PDF) of each selected output parameter was constructed. These functions were found to follow a normal distribution. / MIng (Nuclear Engineering), North-West University, Potchefstroom Campus, 2014

Made available in DSpace on 2014-10-09T12:55:30Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:16Z (GMT). No. of bitstreams: 0 / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:55:30Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:16Z (GMT). No. of bitstreams: 0 / Foi desenvolvida uma técnica de medida de nível em vasos de pressão usando sondas térmicas resfriadas internamente por um fluido e análise dos dados experimentais com Redes Neurais Artificiais (RNA´s). Esse novo conceito de sondas térmicas foi testado em uma Bancada Experimental para Testes de Sondas de Nível (BETSNI) com duas seções de testes, ST1 e ST2. Dois projetos distintos de sondas foram construídos: Sonda de Tubos Concêntricos e Sonda de Tubo U. Um Sistema de Aquisição de Dados (SAD) foi montado para registrar os dados experimentais. Testes foram realizados tanto para condições de nível nas seções de testes em estado estacionário quanto para transientes. Os dados experimentais de temperatura e de nível obtidos foram usados para compor tabelas de treinamento e de verificação usadas para implementar RNA´s no programa RETRO-05, que simula um Perceptron de Múltiplas Camadas com Retropropagação. As análises mostraram que a técnica pode ser aplicada para medir o nível em vasos de pressão. As análises mostraram ainda que a técnica é aplicável para um número menor de entradas de temperatura que o inicialmente previsto no projeto das sondas e é robusta, aplicando-se mesmo quando ocorre a perda de alguma informação de temperatura. Dados experimentais disponíveis na literatura referentes a uma sonda térmica aquecida eletricamente também foram usados nas análises com RNA´s, produzindo bons resultados. Os resultados das análises indicaram que a técnica é eficaz e robusta, podendo ser aprimorada e aplicada para medidas de nível em vasos de pressão. / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

El Generation IV International Forum (GIF) [1] es un programa internacional dedicado a apoyar, coordinar y dirigir las iniciativas de investigación y desarrollo encaminados a implementar las soluciones tecnológicas que caracterizarán a la siguiente generación de reactores nucleares. Estos reactores se caracterizaran por una gestión más eficiente del combustible nuclear, un incremento en las exigencias de seguridad y una alta competitividad económica. Con tales objetivos, GIF propuso una serie de diseños potencialmente capaces de alcanzarlos. Estos diseños son tecnológicamente muy distintos a las plantas nucleares comerciales actuales al utilizar neutrones de espectro rápido y consecuentemente refrigeración por metales líquidos. Estos nuevos diseños requieren el desarrollo y validación de herramientas computacionales capaces de simular el comportamiento de la planta tanto en fase estacionaria como en transitoria y por tanto sean aplicables en los procesos de diseño y licitación de dichas plantas. El objetivo de esta tesis es el de adaptar los códigos computacionales actuales aplicados a la simulación de reactores refrigerados por agua a reactores rápidos refrigerados por metales líquidos, tales como el sodio o el plomo y el desarrollo de modelos capaces de simular de una manera consistente el comportamientos de los sistemas ante determinados eventos que constituyen la base de diseño de la planta Para ello se adaptaran dichos códigos a la fenomenología específica de estos reactores, se desarrollaran modelos termo-hidráulicos y neutrónicos tanto unidimensionales como tridimensionales de los diseños propuestos y se validarán los resultados para demostrar su aplicabilidad. El trabajo incluye la implementación de correlaciones específicas para habilitar los códigos para el cálculo de la condiciones termo-hidráulicas de los refrigerantes así como la adaptación de los esquemas de acoplamiento termo-hidráulico-neutrónicos existentes a esta nueva tecnología. / Lázaro Chueca, A. (2014). Development, assessment and application of computational tools for design safety analysis of liquid metal cooled fast breeder reactors [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/39353 / TESIS

Der Reaktordynamikcodes DYN3D wurde in der neu entwickelten Mehrgruppen-Version DYN3D-MG für die Anwendung auf wassergekühlte Reaktoren alternativ zu industriellen DWR und SWR ertüch-tigt. Es wurde die Anwendbarkeit für den graphitmoderierten Druckröhrenreaktor EGP-6 (KKW Bilibi-no), eine Konzeptstudie eines fortgeschrittenen Siedewasserreaktors mit schnellem Neutronenspekt-rum (RMWR) und das Reaktorkonzept RUTA-70 zur Wärmeversorgung nachgewiesen. Beim RUTA-Reaktor geht es vor allem um die Modellierung des Naturumlaufs des Kühlmittels bei niedrigen Sys-temdrücken. Zur Validierung wurden Experimente zu flashing-induzierten Naturumlaufinstabilitäten an der Versuchsanlage CIRCUS der TU Delft mit RELAP5 nachgerechnet. Für die Anwendung von DYN3D auf die alternativen Reaktorkonzepte wurden Modellerweiterungen und Anpassungen vorgenommen, u.a. Modifikationen in den Wärmeleitungs- und -übergangsmodellen. Vergleichsrechnungen mit dem stationären russischen Feingitter-Diffusionscode ACADEM ergänzen die Verifikationsdatenbasis von DYN3D-MG. Zur Validierung wurden zwei reak-tordynamische Experimente am Reaktor EGP-6 nachgerechnet. Für Reaktoren EGP-6, RMWR und RUTA wurden verschiedene Transienten mit Ausfahren von Re-gelstäben mit und ohne Reaktorschnellabschaltung gerechnet. Weiterhin wurden Analysen für den ATWS-Störfall \"Abschalten aller Hauptkühlmittelpumpen bei Vollleistung\" für den RUTA-Reaktor mit den gekoppelten Programmkomplexen DYN3D/ATHLET und DYN3D/RELAP5 durchgeführt. Der Reaktor geht in einen sicheren Zustand mit reduzierter Leistung bei Naturumlauf des Kühlmittels über. Die Ergebnisse von Analysen zum unkontrollierten Ausfahren einer Regelgruppe für den RMWR lassen dagegen eine belastbare Schlussfolgerung bezüglich der Beherrschbarkeit des Aus-fahrens einer Regelgruppe nicht zu. Abschließend wurde der Nutzen der Programmertüchtigung von DYN3D für die Anwendung auf GenIV -Konzepte und LWR mit hohem Konversionsfaktor bewertet.

The system code TRACE is the latest development of the U.S. Nuclear Regulatory Commission (US NRC). TRACE, developed for the analysis of operational conditions, transients and accidents of light water reactors (LWR), is a best-estimate code with two fluid, six equation models for mass, energy, and momentum conservation, and related closure models. Since TRACE is mainly applied to LWR specific issues, the validation process related to innovative nuclear systems (liquid metal cooled systems, systems operated with supercritical water, etc.) is very limited, almost not existing. In this work, essential contribution to the validation of TRACE related to lead and lead alloy cooled systems as well as systems operated with supercritical water is provided in a consistent and corporate way. In a first step, model discrepancies of the TRACE source code were removed. This inconsistencies caused the wrong prediction of the thermo physical properties of supercritical water and lead bismuth eutectic, and hence the incorrect prediction of heat transfer relevant characteristic numbers like Reynolds or Prandtl number. In addition to the correction of the models to predict these quantities, models describing the thermo physical properties of lead and Diphyl THT (synthetic heat transfer medium) were implemented. Several experiments and numerical benchmarks were used to validate the modified TRACE version. These experiments, mainly focused on wall-to-fluid heat transfer, revealed that not only the thermo physical properties are afflicted with inconsistencies but also the heat transfer models. The models for the heat transfer to liquid metals were enhanced in a way that the code can now distinguish between pipe and bundle flow by using the right correlation. The heat transfer to supercritical water was not existing in TRACE up to now. Completely new routines were implemented to overcome that issue. The comparison of the calculations to the experiments showed, on one hand, the necessity of these changes and, on the other hand, the success of the new implemented routines and functions. The predictions using the modified TRACE version were close to the experimental data. After validating the modified TRACE version, two design studies related to the Generation IV International Forum (GIF) were investigated. In the first one, a core of a lead-cooled fast reactor (LFR) was analyzed. To include the interaction between the thermal hydraulic and the neutron kinetic due to temperature and density changes, the TRACE code was coupled to the program system ERANOS2.1. The results gained with that coupled system are in accordance with theory and helped to identify sub-assemblies with the highest loads concerning fuel and cladding temperature. The second design which was investigated was the High Performance Light Water Reactor (HPLWR). Since the design of the HPLWR is not finalized, optimization of vital parameters (power, mass flow rate, etc.) are still ongoing. Since most of the parameters are affecting each other, an uncertainty and sensitivity analysis was performed. The uncertainty analysis showed the upper and lower boundaries of selected parameters, which are of importance from the safety point of view (e.g., fuel and cladding temperature, moderator temperature). The sensitivity study identified the most relevant parameters and their influence on the whole system.

In this thesis, an innovative coarse grid CFD approach is developed that aims toexploit the capabilities of sub-channel codes and CFD methods while overcoming theirlimitations. In the approach, a very coarse mesh is implemented in the CFD softwareOpenFOAM and a new wall treatment, based on the traditional concept of the wallfunction, is applied to the wall boundary conditions of the domain to take into accountthe low resolution of the grid which does not allow to effectively capture the effect of thesolid walls on the thermo-hydraulics of the flow. To investigate the performance of thenew approach, the method is implemented first in three simple test cases for whichthe sub-channel codes are the state-of-the-art thermo-hydraulic analysis since theyare single-phase flow problems in which there are no prevailing 3D flow conditions.An additional test case representing a 2x2 fuel bundle with three full-length rods andone half-length rod is investigated to verify the behavior of the new approach in caseswhere secondary flows are present. The results for the pressure fields are comparedwith the analytical pressure profiles for the four test cases that well represent the onesthat would be obtained with sub-channel code analysis, while the results for the wallshear stresses obtained in the four test cases are compared with the ones obtained witha more refined mesh in which the traditional wall function approach is implementedsince they should be the best estimation of the actual wall shear stresses at the walldomain. For the first two cases, the developed approach produces reasonable resultswith a good agreement to the analytical pressure profiles while the other two testcases show that the methodology has a limited applicability and, before proceedingwith the extension of the new approach to single-phase problems with 3D prevailingphenomena and two-phase problems, it is necessary to solve the issues that emerge forsome types of cases. / I denna avhandling utvecklas en innovativ CFD-metod med grovt rutnät som syftar till att utnyttja kapaciteten hos underkanalskoder och CFD-metoder och samtidigt övervinna deras begränsningar. I metoden implementeras ett mycket grovt nät i CFD-programvaran OpenFOAM och en ny väggbehandling, baserad på det traditionella konceptet med vägg väggfunktion, tillämpas på domänens vägggränsvillkor för att ta hänsyn till den låga upplösningen av nätet som inte tillåter att effektivt fånga effekten av de solida väggar på flödets termo-hydraulik. För att undersöka prestandan hos den nya tillvägagångssättet implementeras metoden först i tre enkla testfall för vilka subkanalskoderna är den senaste termo-hydrauliska analysen eftersom de är enfasflödesproblem där det inte finns några rådande 3D-flödesförhållanden.Ett ytterligare testfall som representerar ett 2x2 bränsleknippe med tre fullängdsstavar och en halvlång stav undersöks för att verifiera beteendet hos den nya metoden i fall där sekundära flöden förekommer. Resultaten för tryckfälten jämförs med de analytiska tryckprofilerna för de fyra testfall som väl representerar de som som skulle erhållas med kodanalys av underkanalen, medan resultaten för väggskjuvspänningarna skjuvspänningar som erhållits i de fyra testfallen jämförs med de som erhållits med ett mer förfinat nät i vilket den traditionella väggfunktionsmetoden är implementerad eftersom de bör vara den bästa uppskattningen av de faktiska väggskjuvspänningarna vid väggens domän. För de två första fallen ger den utvecklade metoden rimliga resultat med en god överensstämmelse med de analytiska tryckprofilerna medan de andra två visar att metoden har en begränsad tillämplighet och, innan man går vidare med utvidgningen av den nya metoden till enfasproblem med 3D-fenomen och två fenomen och tvåfasproblem, är det nödvändigt att lösa de problem som uppstår för vissa vissa typer av fall.

DYN3D is an best estimate advanced code for the three-dimensional simulation of steady-states and transients in light water reactor cores with quadratic and hexagonal fuel assemblies. Burnup and poison-dynamic calculations can be performed. For the investigation of wide range transients, DYN3D is coupled with system codes as ATHLET and RELAP5. The neutron kinetic model is based on the solution of the three-dimensional two-group neutron diffusion equation by nodal expansion methods. The thermal-hydraulics comprises a one- or two-phase coolant flow model on the basis of four differential balance equations for mass, energy and momentum of the two-phase mixture and the mass balance for the vapour phase. Various cross section libraries are linked with DYN3D. Systematic code validation is performed by FZR and independent organizations.