O licenciamento de uma instalação nuclear é motivado pela necessidade de proteger os seres humanos e o meio ambiente das radiações ionizantes e, ao mesmo tempo, define as bases para a concepção e a determinação da aceitabilidade da planta. Uma parte importante no processo de licenciamento é a realização de uma análise de acidentes, a qual deve estar documentada no Relatório Final de Análise de Segurança (RFAS). Existem diferentes opções de cálculo na área de acidentes, combinando a utilização de códigos computacionais e dados de entrada, para fins de licenciamento. Uma delas é a Best Estimate Plus Uncertainty (BEPU), que considera dados de entrada realistas e as incertezas associadas. As aplicações de abordagens BEPU em processos de licenciamento iniciaram-se nos anos 2000, primeiro para análise de Acidente de Perda de Refrigerante (Loss of Coolant Accident - LOCA), e depois para a análise de acidentes como um todo, documentados no Capítulo 15 do RFAS. O presente trabalho tem como objetivo principal demonstrar que é possível a aplicação da metodologia BEPU em todas as análises contidas no RFAS, identificando as disciplinas-chave do processo de licenciamento e os códigos computacionais utilizados. Este trabalho foi desenvolvido em conjunto com a Universidade de Pisa, Itália, com a colaboração do Prof. Dr. Francesco D\'Áuria. A principal motivação desse trabalho é o aprimoramento da metodologia BEPU para sua implementação em reatores do tipo PWR (Pressurized Water Reactor) no Brasil e no mundo, especialmente para fins de licenciamento, uma vez que as plantas nucleares brasileiras têm pouca experiência na área de cálculo de incertezas. / The licensing process of a nuclear power plant is motivated by the need to protect humans and the environment from ionizing radiation and, at the same time, sets out the basis for the design and determining the acceptability of the plant. An important part of the licensing process is the realization of accident analysis, which should be documented in the Final Safety Analysis Report (FSAR). There are different options on accidents calculation area by combining the use of computer codes and data entry for licensing purposes. One is the Best Estimate Plus Uncertainty (BEPU), which considers realistic input data and associated uncertainties. Applications of BEPU approaches in licensing procedures were initiated in the 2000s, first to analysis of Loss of Coolant Accident (LOCA), and then to the accident analysis as a whole, documented in Chapter 15 of the FSAR. This work has as main objective demonstrate the implementation of BEPU methodology in all analyses contained in FSAR is possible, identifying the key disciplines of the licensing process and the computer codes. This work was done in conjunction with the University of Pisa, Italy, with the collaboration of Professor Francesco D\'Auria. The main motivation of this work is the improvement of BEPU methodology for its implementation in PWR (Pressurized Water Reactor) reactors in Brazil and the world, especially for licensing purposes, since the Brazilian nuclear plants have little experience in the regulatory area, and specifically in calculation uncertainties.

O licenciamento de uma instalação nuclear é motivado pela necessidade de proteger os seres humanos e o meio ambiente das radiações ionizantes e, ao mesmo tempo, define as bases para a concepção e a determinação da aceitabilidade da planta. Uma parte importante no processo de licenciamento é a realização de uma análise de acidentes, a qual deve estar documentada no Relatório Final de Análise de Segurança (RFAS). Existem diferentes opções de cálculo na área de acidentes, combinando a utilização de códigos computacionais e dados de entrada, para fins de licenciamento. Uma delas é a Best Estimate Plus Uncertainty (BEPU), que considera dados de entrada realistas e as incertezas associadas. As aplicações de abordagens BEPU em processos de licenciamento iniciaram-se nos anos 2000, primeiro para análise de Acidente de Perda de Refrigerante (Loss of Coolant Accident - LOCA), e depois para a análise de acidentes como um todo, documentados no Capítulo 15 do RFAS. O presente trabalho tem como objetivo principal demonstrar que é possível a aplicação da metodologia BEPU em todas as análises contidas no RFAS, identificando as disciplinas-chave do processo de licenciamento e os códigos computacionais utilizados. Este trabalho foi desenvolvido em conjunto com a Universidade de Pisa, Itália, com a colaboração do Prof. Dr. Francesco D\'Áuria. A principal motivação desse trabalho é o aprimoramento da metodologia BEPU para sua implementação em reatores do tipo PWR (Pressurized Water Reactor) no Brasil e no mundo, especialmente para fins de licenciamento, uma vez que as plantas nucleares brasileiras têm pouca experiência na área de cálculo de incertezas. / The licensing process of a nuclear power plant is motivated by the need to protect humans and the environment from ionizing radiation and, at the same time, sets out the basis for the design and determining the acceptability of the plant. An important part of the licensing process is the realization of accident analysis, which should be documented in the Final Safety Analysis Report (FSAR). There are different options on accidents calculation area by combining the use of computer codes and data entry for licensing purposes. One is the Best Estimate Plus Uncertainty (BEPU), which considers realistic input data and associated uncertainties. Applications of BEPU approaches in licensing procedures were initiated in the 2000s, first to analysis of Loss of Coolant Accident (LOCA), and then to the accident analysis as a whole, documented in Chapter 15 of the FSAR. This work has as main objective demonstrate the implementation of BEPU methodology in all analyses contained in FSAR is possible, identifying the key disciplines of the licensing process and the computer codes. This work was done in conjunction with the University of Pisa, Italy, with the collaboration of Professor Francesco D\'Auria. The main motivation of this work is the improvement of BEPU methodology for its implementation in PWR (Pressurized Water Reactor) reactors in Brazil and the world, especially for licensing purposes, since the Brazilian nuclear plants have little experience in the regulatory area, and specifically in calculation uncertainties.

Ce travail de thèse se place dans le contexte des études d’accidents graves sur les réacteurs à neutrons rapides à caloporteur sodium. Dans le cadre de la démarche de conception et de sûreté, tous les types d’accidents doivent être étudiés afin d’assurer l’exhaustivité de l’analyse de sûreté, en traitant la variabilité des scénarios accidentels et en quantifiant les marges de sûreté. Pour cela, des outils physiques sont développés pour être couplés à des techniques avancées de statistique permettant de répondre rapidement et quantitativement aux questions relatives à la conception du réacteur vis-à-vis des conséquences d’un accident grave et de prendre en compte des incertitudes et de la variabilité des scénarios accidentels. La mise au point de l’outil physique OCARINa (Outil de Calcul analytique Rapide pour les Insertions de réactivité dans un RNR-Na) dédié à la phase primaire du transitoire d’insertion de réactivité non protégée UTOP (Unprotected Transient OverPower) a ainsi fait l’objet de cette thèse. Les travaux ont porté sur l’identification des phénomènes physiques prépondérants, leur modélisation (thermique et thermomécanique), et une contribution à la validation expérimentale et numérique. Enfin, une application de l’intérêt de cet outil a été réalisée à partir de deux études BEPU (Best-Estimate Plus Uncertainties). Elle a permis d’identifier les paramètres les plus influents sur la réponse de l’outil, et de quantifier leur impact vis-à-vis des résultats expérimentaux. / Within the framework of the Generation IV Sodium-cooled Fast Reactor R&D French program, a new physico-statistical approach is currently followed by the CEA for accident transient calculations in complement to the reference mechanistic codes. This approach involves a fast-running description of extended accident sequences coupling physical models for the main phenomena to advanced statistical analysis techniques. A large number of simulations may be performed in a reasonable computational time while describing all the possible bifurcations of the accident transient. In this context, this PhD work deals with the development (models and results assessment) of the physical tool dedicated to the primary phase of the Unprotected Transient OverPower accident called OCARINa (Outil de Calcul analytique Rapide pour les Insertions de réactivité dans un RNR-Na). The accident main phenomena, their modelling (thermal and thermomechanical models) and a contribution to the experimental and numerical validation are described. Finally a demonstration of BEPU studies has been done, resulting in the identification and the impact quantification of the more influent uncertain parameters on experimental results.

Durant les dernières décennies, l’évolution de la puissance de calcul a conduit au développement de codes de simulation en physique des réacteurs de plus en plus prédictifs pour la modélisation du comportement d’un réacteur nucléaire en situation de fonctionnement normal et accidentel. Un cadre d’analyse d’incertitudes cohérent avec l’utilisation de modélisations Best Estimate (BE) a été développé. On parle d’approche Best Estimate Plus Uncertain-ties (BEPU) et cette approche donne lieu `a de nombreux travaux de R&D à l’international en simulation numérique. Dans cette thèse, on étudie la quantification d’incertitudes multi-physiques dans le cas d’un transitoire d’ éjection de Grappe de contrôle (REA- Rod Ejection Accident) dans un Réacteur à Eau Pressurisée (REP). La modélisation BE actuellement disponible au CEA est réalisée en couplant les codes APOLLO3 R (netronique) et FLICA4 (thermohydraulique-thermique du combustible) dans l’environnement SALOME/CORPUS. Dans la première partie de la thèse, on examine différents outils statistiques disponibles dans la littérature scientifique dont la réduction de dimension, l’analyse de sensibilité globale, des modèles de substitution et la construction de plans d’expérience. On utilise ces outils pour développer une méthodologie de quantification d’incertitudes. Dans la deuxième partie de la thèse, on améliore la modélisation du comportement du combustible. Un couplage Best Effort pour la simulation d’un transitoire REA est disponible au CEA. Il comprend le code ALCYONE V1.4 qui permet une modélisation fine du comportement thermomécanique du combustible. Cependant, l’utilisation d’une telle modélisation conduit à une augmentation significative du temps de calcul ce qui rend actuellement difficile la réalisation d’une analyse d’incertitudes. Pour cela, une méthodologie de calibrage d’un modèle analytique simplifié pour le transfert de chaleur pastille-gaine basé sur des calculs ALCYONE V1.4 découplés a été développée. Le modèle calibré est finalement intégré dans la modélisation BE pour améliorer sa prédictivité. Ces deux méthodologies sont maquettées initialement sur un cœur de petite échelle représentatif d’un REP puis appliquées sur un cœur REP à l’échelle 1 dans le cadre d’une analyse multi-physique d’un transitoire REA. / The computational advancements of the last decades lead to the development of numerical codes for simulating the reactor physics with increa-sing predictivity allowing the modeling of the beha-vior of a nuclear reactor under both normal and acci-dental conditions. An uncertainty analysis framework consistent with Best Estimate (BE) codes was develo-ped in order to take into account the different sources of uncertainties. This framework is called Best Esti-mate Plus Uncertainties (BEPU) and is currently a field of increasing research internationally. In this the-sis we study the multi-physics uncertainty quantifi-cation for Rod Ejection Accident (REA) in Pressuri-zed Water Reactors (PWR). The BE modeling avai-lable in CEA is used with a coupling of APOLLO3 (neutronics) and FLICA4 (thermal-hydraulics and fuel-thermal) in the framework of SALOME/CORPUS tool. In the first part of the thesis, we explore different statistical tools available in the scientific literature including: dimension reduction, global sensitivity analy-sis, surrogate modeling and design of experiments. We then use them in order to develop an uncer-tainty quantification methodology. In the second part of the thesis, we improve the BE modeling in terms of its uncertainty representation. A Best Effort coupling scheme for REA analysis is available at CEA. This in-cludes ALCYONE V1.4 code for a detailed modeling of fuel-thermomechanics behavior. However, the use of such modeling increases significantly the compu-tational cost for a REA transient rendering the uncer-tainty analysis prohibited. To this purpose, we deve-lop a methodology for calibrating a simplified analytic gap heat transfer model using decoupled ALCYONE V1.4 REA calculations. The calibrated model is finally used to improve the previous BE modeling. Both de-veloped methodologies are tested initially on a small scale core representative of a PWR and then applied on a large scale PWR core.

In nuclear power plant design and, after, when they are under work, in front of any change in the design or periodical safety review, it is necessary to perform safety studies in order to guarantee the safety operation along their useful life. These safety studies, traditionally has been divided between deterministic safety analysis and probabilistic safety analysis, although the last years trending is to integrate the characteristics of both classes of analysis in order to build more complete safety studies. Among the deterministic safety analysis, when the Best Estimate (BE) codes are employed and, in addition, the uncertainty effect are taken into account, we are inside of the methodology called Best Estimate Plus Uncertainty (BEPU). This Thesis provides new tools and procedures in order to perform the deterministic safety analysis of nuclear power plants by means of Best Estimate methodology through of several applications employed. Statistical tools are provided for performing BEPU analysis. Particularly, a procedure is presented for built BEPU studies that can be applied in almost all the transients and facility in a methodical way. This procedure is comprehensive and include from the development of the transient scenario by means of BE thermalhydraulic code and the input parameters selection to the uncertainty propagation over the safety criteria and the verification of their compliance using different uncertainty analysis methods, both parametric and non parametric methods. With the purpose of demonstrating the procedure versatility, this it is applied to the studio of transients and facilities with different phenomenology. Specifically, it have been applied to: PWR nuclear power plant for a Large Break Loss of Coolant Accident (LBLOCA), experimental facility for a Small Break Loss of Coolant Accident (SBLOCA) and in a spent fuel storage of a PWR nuclear power plant for a loss of coolant accident. Last, this Thesis contribute with a methodology accordingly for incorporating assumptions from probabilistic safety analysis about system configuration availability inside the deterministic safety analysis. Therefore, an approach enclosed into the known as Extended BEPU (EBEPU) methodologies is constructed. In order to demonstrate the viability and applicability of this methodology, an application case is provided, which consists in Loss of Feed Water system (LOFW) in a PWR nuclear power plant. The work carried out in this PhD thesis are enclosed into the grant of "Formación de Personal Investigador (FPI)-Subprograma1 de la convocatoria de 2012" supported by the "Programa de Ayudas de Investigación y Desarrollo (PAID)" of the "Universitat Politècnica de València". / En el diseño de las centrales nucleares (CCNN) y, una vez puestas en funcionamiento, ante cualquier cambio de diseño o revisión periódica de seguridad (RPS), es necesario realizar estudios de seguridad para garantizar la operación segura durante su vida útil. Dichos estudios, tradicionalmente se han dividido en análisis deterministas de seguridad (ADS) y análisis probabilistas de seguridad (APS), aunque la tendencia de los últimos años es integrar las características de ambos tipos dando lugar a estudios de seguridad más completos. Dentro de los ADS, cuando se utilizan códigos termohidráulicos Best Estimate (BE) para el análisis de las secuencias accidentales y, además, se tiene en cuenta el efecto de las incertidumbres, estamos dentro de la metodología denominada Best Estimate Plus Uncertainty (BEPU). La presente Tesis aporta nuevas herramientas y procedimientos para realizar el análisis determinista de seguridad de las centrales nucleares mediante metodología Best Estimate a través de varios desarrollos y aplicaciones realizados. Se aportan herramientas estadísticas adecuadas para llevar a cabo análisis BEPU. Particularmente, se presenta un procedimiento para realizar estudios BEPU que puede ser aplicado a casi cualquier tipo de accidente e instalación de manera metódica. Dicho procedimiento es integral y abarca desde el desarrollo del escenario objeto de estudio mediante un código termohidráulico BE y la selección de los parámetros de entrada, hasta la propagación de incertidumbres sobre los criterios de seguridad y la verificación del cumplimiento de los mismos utilizando diferentes métodos de análisis de incertidumbre, tanto paramétricos como no paramétricos. Con el objetivo de demostrar la versatilidad del procedimiento, este se aplica al estudio de transitorios e instalaciones con casuísticas diferentes. En concreto, se ha aplicado en: una CN de tipo PWR para una rotura grande (LBLOCA), en una instalación experimental para una rotura pequeña (SBLOCA) y en una piscina de combustible gastado de una CN de tipo PWR para una pérdida de refrigerante. Por último, en la presente Tesis se propone una metodología para incorporar las suposiciones, propias del APS sobre la configuración de los sistemas de seguridad dentro del ADS. De esta forma, se da lugar a una aproximación enmarcada dentro de las metodologías conocidas como Extended BEPU (EBEPU). Para demostrar la viabilidad y aplicabilidad de dicha metodología, se aporta un caso de aplicación que consiste en la pérdida de agua de alimentación (LOFW) en una CN de tipo PWR. El trabajo realizado en esta tesis doctoral se enmarca dentro de la beca de Formación de Personal Investigador (FPI)-Subprograma1 de la convocatoria de 2012 auspiciada por el Programa de Ayudas de Investigación y Desarrollo (PAID) de la Universitat Politècnica de València. / Al disseny de les centrals nuclears (CCNN) i, una vegada que aquestes estan en funcionament, davant qualsevol canvi de disseny o revisió periòdica de seguretat (RPS), és necessari realitzar estudis de seguretat per garantir l'operació segura durant la seua vida útil. Aquests estudis, tradicionalment s'han dividit en anàlisis deterministes de seguretat (ADS) y anàlisis probabilistes de seguretat (APS), encara que la tendència del últims anys és integrar les característiques d'ambos tipus resultant en estudis de seguretat més complets. Dins del ADS, quan s'utilitzen codis termohidràulics Best Estimate (BE) per a l'anàlisi de les seqüències accidentals i, a més, es te en compte l'efecte de les incerteses, estem dins de la metodologia anomenada Best Estimate Plus Uncertainty (BEPU). La present Tesi aporta noves ferramentes i procediments per realitzar l'anàlisi determinista de seguretat de les centrals nuclear mitjançant metodologia Best Estimat emplenant diferents desenvolupaments y aplicacions realitzades. S'aporten ferramentes estadístiques adequades per dur a terme anàlisis BEPU. Particularment, es presenta un procediment per realitzar estudis BEPU que pot ser aplicat a quasi qualsevol tipus d'accident i instal·lació de manera metòdica. Aquest procediment és integral i assoleix des del desenvolupament de l'escenari objecte d'estudi mitjançant un codi termohidràulic BE i la selecció del paràmetres d'entrada fins a la propagació d'incerteses al voltant dels criteris de seguretat i la verificació del acompliment dels mateixos fent ús de diversos mètodes d'anàlisi d'incertesa, tant paramètrics com no paramètrics. Amb l'objectiu de demostrar la versatilitat del procediment, aquest s'aplica a l'estudi de transitoris i instal·lacions amb casuístiques diferents. En concret, s'ha aplicat en: una CN de tipus PWR per a un trencament gran (LBLOCA), en una instal·lació experimental per a un trencament xicotet (SBLOCA) i en una piscina de combustible gastat de una CN de tipus PWR per a una pèrdua de refrigerant. Per últim, a la present Tesi es proposa una metodologia per a incorporar les suposicions, pròpies del APS en quant a la configuració dels sistemes de seguretat dins del ADS. D'aquesta forma, s'obté una aproximació emmarcada dins de les metodologies conegudes com Extended BEPU (EBEPU). Per demostrar la viabilitat i aplicabilitat d'aquesta metodologia, s'aporta un cas d'aplicació que consisteix en la pèrdua d'aigua d'alimentació (LOFW) en una CN de tipus PWR. El treball realitzat en aquesta Tesi s'emmarca dins de la beca de "Formación de Personal Investigador (FPI)-Subprograma1 de la convocatoria de 2012" finançada pel "Programa de Ayudas de Investigación y Desarrollo (PAID)" de la Universitat Politècnica de València. / Sánchez Sáez, F. (2017). APORTACIONES AL ANÁLISIS DETERMINISTA DE SEGURIDAD DEL LAS CENTRALES NUCLEARES MEDIANTE METODOLOGÍA BEST ESTIMATE [Tesis doctoral]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/86131

McMaster University DOCTOR OF PHILOSOPHY (2022) Hamilton, Ontario (Engineering Physics) TITLE: Development and Evaluation of Polaris CANDU Geometry Modelling and of TRACE_Mac/PARCS_Mac Coupling with RRS for CANDU Analysis AUTHOR: Simon Younan, M.A.Sc. (McMaster University), B.Eng. (McMaster University) SUPERVISOR: Dr. David Novog NUMBER OF PAGES: xiv, 163 / In the field of nuclear safety analysis, as computers have become more powerful, there has been a trend away from low-fidelity models using conservative assumptions, to high-fidelity best-estimate models combined with uncertainty analysis. A number of these tools have been developed in the United States, due to the popularity of light water reactors. These include the SCALE analysis suite developed by ORNL, as well as the PARCS and TRACE tools backed by the USNRC. This work explores adapting the capabilities of these tools to the analysis of CANDU reactors. The Polaris sequence, introduced in SCALE 6.2, was extended in this work to support CANDU geometries and compared to existing SCALE sequences such as TRITON. Emphasis was placed on the Embedded Self-Shielding Method (ESSM), introduced with Polaris. Both Polaris and ESSM were evaluated and found to perform adequately for CANDU geometries. The accuracy of ESSM was found to improve when the precomputed selfshielding factors were updated using a CANDU representation. The PARCS diffusion code and the TRACE system thermalhydraulics code were coupled, using the built-in coupling capability between the two codes. In addition, the Exterior Communications Interface (ECI), used for coupling with TRACE, was utilized. A Python interface to the ECI library was developed in this work and used to couple an RRS model written in Python to the coupled PARCS/TRACE model. A number of code modifications were made to accommodate the required coupling and correct code deficiencies, with the modified versions named PARCS_Mac and TRACE_Mac. The coupled codes were able to simulate multiple transients based on prior studies as well as operational events. The code updates performed in this work may be used for many future studies, particularly for uncertainty propagation through a full set of calculations, from the lattice model to a full coupled system model. / Thesis / Doctor of Philosophy (PhD) / Modern nuclear safety analysis tools offer more accurate predictions for the safety and operation of nuclear reactors, including CANDU reactors. These codes take advantage of modern computer hardware, and also a shift in philosophy from conservative analysis to best estimate plus uncertainty analysis. The goal of this thesis was to adapt a number of modern tools to support CANDU analysis and uncertainty propagation, with a particular emphasis on coupling of multiple interacting models. These tools were then demonstrated, and results analyzed. The simulations performed in this work were successful in producing results comparable to prior studies along with experimental and operational data. This included the simulation of four weeks of reactor operation including “shim mode” operation. Sensitivity and uncertainty analyses were performed over the course of the work to quantify the precision and significance of the results as well as to identify areas of interest for future research.

Transient analyses at Forsmark nuclear power plant are currently performed using realistic computer models in conjunction with conservative estimates of initial- and boundary conditions. This is known as the combined methodology. Due to the conservative estimates, the methodology runs the risk of sometimes over-estimating certain safety criteria which will negatively affect the optimization of reactor operation. The Best Estimate Plus Uncertainty (BEPU) methodology can provide higher safety margins by using probabilities instead of conservatisms when estimating initial and boundary conditions. The BEPU methodology applies a Monte Carlo method to assess the distribution of one or several key outputs. This study focuses on the lowest dryout margin achieved during each Monte Carlo simulation. The tolerance limits of the output are set with the help of Wilks formula using a one-sided 95% tolerance limit with 95% confidence. A total of 36 unique parameters describing initial and boundary conditions have been sampled for each Monte Carlo simulation. The parameters have been sampled using either Gaussian or Uniform distribution functions. The random nature of the Monte Carlo simulations has uncovered alternative event sequences and end states that are not seen in the combined methodology. Assessing the choice of order statistic in Wilks formula also concludes that there are diminishing returns the higher the order statistic is. When choosing the order statistic, one should consider the trade-off between an increased accuracy in the estimated outputs and the increased computational time required. The conservative methodology uses a mix of conservative and nominal estimations of key parameters. The difference in dryout margin between the conservative and the Monte Carlo results should therefore not be used to draw a conclusion about which methodology out-performs the other. While the Monte Carlo simulations do not result in an improved core optimization, they can act as a complement to the combined methodology by providing a more detailed analysis of possible event pathways for a postulated transient.

In 1988, the US Nuclear Regulatory Commission approved an amendment that allowed the use of best-estimate methods. This led to an increased development, and application of Best Estimate Plus Uncertainty (BEPU) safety analyses. However, a greater burden was placed on the licensee to justify all uncertainty estimates. A review of the current state of the BEPU methods indicate that there exists a number of significant criticisms, which limits the BEPU methods from reaching its full potential as a comprehensive licensing basis. The most significant criticism relates to the lack of a formal framework for distinguishing between aleatory and epistemic uncertainties. This has led to a prevalent belief that such separation of uncertainties is for convenience, rather than one out of necessity. In this thesis, we address the above concerns by developing a statistically rigorous framework to characterize the different uncertainty types. This framework is grounded on the philosophical concepts of knowledge. Considering the Plato problem, we explore the use of probability as a means to gain knowledge, which allows us to relate the inherent distinctness in knowledge with the different uncertaintytypesforanycomplexphysicalsystem. Thisframeworkis demonstrated using nuclear analysis problems, and we show through the use of structural models that the separation of these uncertainties leads to more accurate tolerance limits relative to existing BEPU methods. In existing BEPU methods, where such a distinction is not applied, the total uncertainty is essentially treated as the aleatory uncertainty. Thus, the resulting estimated percentile is much larger than the actual (true) percentile of the system's response. Our results support the premise that the separation of these two distinct uncertainty types is necessary and leads to more accurate estimates of the reactor safety margins. / Thesis / Doctor of Philosophy (PhD)

The nuclear industry has long relied upon bounding parametric analyses in predicting the safety margins of reactor designs undergoing design-basis accidents. These methods have been known to return highly-conservative results, limiting the operating conditions of the reactor. The Best-Estimate Plus Uncertainty (BEPU) method using a modernized version of the Code-Scaling, Applicability, and Uncertainty (CSAU) methodology has been applied to more accurately predict the safety margins of the Oregon State University Advanced Plant Experiment (APEX) facility experiencing a Loss-of-Feedwater Accident (LOFA). The statistical advantages of the Bayesian paradigm of probability was utilized to incorporate prior knowledge when determining the analysis required to justify the safety margins. RELAP5 Mod 3.3 was used to accurately predict the thermal-hydraulics of a primary Feed-and-Bleed response to the accident using assumptions to accompany the lumped-parameter calculation approach. A novel coupling of thermal-hydraulic and statistical software was accomplished using the Symbolic Nuclear Analysis Package (SNAP). Uncertainty in Peak Cladding Temperature (PCT) was calculated at the 95/95 probability/confidence levels under a series of four separate sensitivity studies. / Graduation date: 2013