Benchmark calculations for the validation of the coupled neutron kinetics/thermohydraulic code complex DYN3D-ATHLET are described. Two benchmark problems concerning hypothetical accident scenarios with leaks in the steam system for a VVER-440 type reactor and the TMI-1 PWR have been solved. The first benchmark task has been defined by FZR in the frame of the international association "Atomic Energy Research" (AER), the second exercise has been organised under the auspices of the OECD. While in the first benchmark the break of the main steam collector in the sub-critical hot zero power state of the reactor was considered, the break of one of the two main steam lines at full reactor power was assumed in the OECD benchmark. Therefore, in this exercise the mixing of the coolant from the intact and the defect loops had to be considered, while in the AER benchmark the steam collector break causes a homogeneous overcooling of the primary circuit. In the AER benchmark, each participant had to use its own macroscopic cross section libraries. In the OECD benchmark, the cross sections were given in the benchmark definition. The main task of both benchmark problems was to analyse the re-criticality of the scrammed reactor due to the overcooling. For both benchmark problems, a good agreement of the DYN3D-ATHLET solution with the results of other codes was achieved. Differences in the time of re-criticality and the height of the power peak between various solutions of the AER benchmark can be explained by the use of different cross section data. Significant differences in the thermohydraulic parameters (coolant temperature, pressure) occurred only at the late stage of the transient during the emergency injection of highly borated water. In the OECD benchmark, a broader scattering of the thermohydraulic results can be observed, while a good agreement between the various 3D reactor core calculations with given thermohydraulic boundary conditions was achieved. Reasons for the differences in the thermohydraulics were assumed in the difficult modelling of the vertical once-through steam generator with steam superheating. Sensitivity analyses which considered the influence of the nodalisation and the impact of the coolant mixing model were performed for the DYN3D-ATHLET solution of the OECD benchmark. The solution of the benchmarks essentially contributed to the qualification of the code complex DYN3D-ATHLET as an advanced tool for the accident analysis for both VVER type reactors and Western PWRs.

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.

Made available in DSpace on 2014-10-09T12:35:26Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:05:40Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:33:05Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:06:06Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:35:26Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:05:40Z (GMT). No. of bitstreams: 0 / Tecnécio-99m (99mTc), o produto de decaimento do molibdênio-99 (99Mo), é um dos radioisótopos mais utilizados na medicina nuclear, abrangendo cerca de 80% de todos os procedimentos de radiodiagnóstico médico pelo mundo. Atualmente o Brasil necessita de uma quantidade de aproximadamente 450 Ci de 99Mo por semana. Devido à crise e à escassez em seu fornecimento que vem sendo observada no cenário mundial desde 2008, o IPEN decidiu desenvolver um projeto próprio para produção de 99Mo através da fissão do urânio-235. O objetivo deste trabalho de dissertação foi desenvolver cálculos neutrônicos e temo-hidráulicos para avaliar a segurança operacional de um dispositivo para produção de 99Mo a ser irradiado no núcleo do reator IEA-R1. Neste dispositivo serão alojados dez alvos do tipo dispersão de UAlx-Al com baixo enriquecimento de urânio (LEU) e densidade de 2,889 gU/cm³. Para o cálculo neutrônico foram utilizados os programas computacionais HAMMER-TECHNION e CITATION e as temperaturas máximas atingidas nos alvos foram calculadas com o código MTRCR-IEAR1. Os cálculos demonstram que a irradiação do dispositivo deverá ocorrer sem consequências adversas à operação do reator. A quantidade total de 99Mo foi calculada com o programa SCALE e considerando que o tempo necessário para o processamento químico e recuperação do 99Mo será de cinco dias após a irradiação, teremos disponível para distribuição uma atividade de 99Mo de 176 Ci para 3 dias de irradiação, 236 Ci para 5 dias de irradiação e 272 Ci para 7 dias de irradiação dos alvos. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Made available in DSpace on 2014-10-09T12:33:05Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:06:06Z (GMT). No. of bitstreams: 0 / Neste trabalho, propõe-se um estudo para verificação do impacto da redução na concentração de urânio nas placas laterais dos elementos combustíveis do reator IEA-R1, nas análises neutrônica e termo-hidráulica. Ao se desenvolver o referido trabalho, reproduziu-se estudo conduzido anteriormente pelo IPEN-CNEN/SP, simulando a queima de elementos combustíveis, cujas placas laterais apresentam densidade de urânio reduzida para 50, 60 e 70% em relação às demais placas do elemento combustível. Tal estudo inicia-se com a análise neutrônica, cujo primeiro passo é o cálculo das seções de choque dos materiais presentes no núcleo a partir de suas concentrações iniciais, com a utilização do código computacional HAMMER; o segundo passo é o cálculo dos fluxos de nêutrons dos grupos rápido e térmico e das densidades de potência nos elementos combustíveis estudados em modelagem do núcleo feita no código computacional CITATION, que utiliza os dados gerados pelo HAMMER. Terminada a análise neutrônica e definidos os elementos combustíveis mais críticos com maior densidade de potência, executa-se a análise termo-hidráulica, que utiliza o modelo termo-hidráulico MCTR-IEA-R1, o qual é baseado no pacote comercial EES. A densidade de potência gerada pelo CITATION é utilizada como dado de entrada da análise termo-hidráulica nas equações de balanço de energia do modelo para o cálculo das temperaturas nos pontos de interesse. Neste trabalho, é feita a comparação da operação do reator com três diferentes densidades de urânio nas placas laterais. Concluiu-se que a redução da densidade de urânio contribui para que a temperatura da superfície do revestimento não ultrapasse o limite estabelecido como condição de operação do reator; não há impacto significativo na queima final dos elementos combustíveis, nem na reatividade do reator IEA-R1. A redução de urânio nas placas laterais dos elementos combustíveis do reator IEA-R1 mostrou ser uma opção viável para evitar problemas de corrosão devido a altas temperaturas. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

A state-of-art research facility has been built at the University of Ottawa, which is suitable for thermalhydraulic experiments in support of the development of the Canadian Supercritical-Water-Cooled Reactor (SCWR). The facility is a recirculating flow loop, using carbon dioxide as a medium and having three different test sections, two tubes with inner diameters of 8 and 22 mm, respectively, and a three-rod bundle. The loop can operate within ranges of pressure, temperature, heat flux and mass flux, which are of interest to the current SCWR design. The present thesis includes a comprehensive description of the facility. It also documents the procedure and results of its commissioning, as well as some preliminary measurements that have been collected so far. It is intended to provide an insight to the design of the facility and its functionality and to serve as a reference for future research activities. A number of tests performed by previous researchers in other facilities were replicated and nearly identical results were obtained. It was demonstrated that the design of the facility is sound and its performance is adequate within the intended ranges of operation conditions. It is expected that the results obtained in this facility will make a significant contribution to the understanding of supercritical heat transfer and pressure losses in the SCWR context.

Das Vorhaben bestand in der Validierung des gekoppelten neutronenkinetisch-thermohydraulischen Programmkomplexes ATHLET/DYN3D für Siedewasserreaktoren durch Teilnahme an dem OECD/NRC Benchmark zum Turbinenschnellschluss. Das von der OECD und der amerikanischen NRC definierte Benchmark basiert auf einem Experiment mit Schließens des Turbinenschnellschlussventils, das 1977 im Rahmen einer Serie von 3 Experimenten im Kernkraftwerk Peach Bottom 2 durchgeführt wurde. Im Experiment erzeugte das Schließen des Ventils eine Druckwelle, die sich unter Abschwächung bis in den Reaktorkern ausbreitete. Die durch den Druckanstieg bewirkte Kondensation von Dampf im Reaktorkern führte zu einem positiven Reaktivitätseintrag. Der folgende Anstieg der Reaktorleistung wurde durch die Rückkopplung und das Einfahren der Regelstäbe begrenzt. Im Rahmen des Benchmarks konnten die Rechenprogramme durch Vergleiche mit den Messergebnissen und den Ergebnissen der anderen Teilnehmer an dem Benchmark validiert werden. Das Benchmark wurde in 3 Phasen oder Exercises eingeteilt. Die Phase I diente der Überprüfung des thermohydraulischen Modells für das System bei vorgegebener Leistungsfreisetzung im Kern. In der Phase II wurden 3-dimensionale Berechnungen des Reaktorkerns für vorgegebene thermohydraulische Randbedingungen durchgeführt. Die gekoppelten Rechnungen für das ausgewählte Experiment und für 4 extreme Szenarien erfolgten in der Phase III. Im Rahmen des Projekts nahm FZR an Phase II und Phase III des Benchmarks teil. Die Rechnungen für Phase II erfolgten mit dem Kernmodell DYN3D unter Berücksichtigung der Heterogenitätsfaktoren und mit 764 thermohydraulischen Kanälen (1 Kanal/Brennelement). Der ATHLET-Eingabedatensatz für die Reaktoranlage wurde von der Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) übernommen und für die Rechnungen zu Phase III, die mit der parallelen Kopplung von ATHLET mit DYN3D erfolgten, geringfügig modifiziert. Für räumlich gemittelte Parameter wurde eine gute Übereinstimmung mit den Messergebnissen und den Resultaten anderer Codes erzielt. Der Einfluss der Modellunterschiede wurde mit Hilfe von Variantenrechnungen zu Phase II untersucht. So können Unterschiede in der Leistungs- und Voidverteilung in einzelnen Brennelementen auf die unterschiedliche neutronenkinetische und thermohydraulische Modellierung des Reaktorkerns zurückgeführt werden. Vergleiche zwischen ATHLET/DYN3D (parallele Kopplung) und ATHLET/QUABOX-CUBBOX (interne Kopplung) zeigen für räumlich gemittelte Parameter nur geringe Unterschiede. Abweichungen in den lokalen Parametern können im wesentlichen mit der unterschiedlichen Modellierung des Reaktorkerns erklärt werden (geringere Anzahl von modellierten Kühlkanälen, keine Berücksichtigung der Heterogenitätsfaktoren und ein anderes Siedemodell in der Rechnung mit ATHLET/QUABOX-CUBBOX). Die Rechnungen für die extremen Szenarien von Phase III zeigen die Anwendbarkeit des gekoppelten Programms ATHLET/DYN3D für die Bedingungen bei Störfällen, die weit über das Experiment hinausgehen.

Benchmark calculations for the validation of the coupled neutron kinetics/thermohydraulic code complex DYN3D-ATHLET are described. Two benchmark problems concerning hypothetical accident scenarios with leaks in the steam system for a VVER-440 type reactor and the TMI-1 PWR have been solved. The first benchmark task has been defined by FZR in the frame of the international association "Atomic Energy Research" (AER), the second exercise has been organised under the auspices of the OECD. While in the first benchmark the break of the main steam collector in the sub-critical hot zero power state of the reactor was considered, the break of one of the two main steam lines at full reactor power was assumed in the OECD benchmark. Therefore, in this exercise the mixing of the coolant from the intact and the defect loops had to be considered, while in the AER benchmark the steam collector break causes a homogeneous overcooling of the primary circuit. In the AER benchmark, each participant had to use its own macroscopic cross section libraries. In the OECD benchmark, the cross sections were given in the benchmark definition. The main task of both benchmark problems was to analyse the re-criticality of the scrammed reactor due to the overcooling. For both benchmark problems, a good agreement of the DYN3D-ATHLET solution with the results of other codes was achieved. Differences in the time of re-criticality and the height of the power peak between various solutions of the AER benchmark can be explained by the use of different cross section data. Significant differences in the thermohydraulic parameters (coolant temperature, pressure) occurred only at the late stage of the transient during the emergency injection of highly borated water. In the OECD benchmark, a broader scattering of the thermohydraulic results can be observed, while a good agreement between the various 3D reactor core calculations with given thermohydraulic boundary conditions was achieved. Reasons for the differences in the thermohydraulics were assumed in the difficult modelling of the vertical once-through steam generator with steam superheating. Sensitivity analyses which considered the influence of the nodalisation and the impact of the coolant mixing model were performed for the DYN3D-ATHLET solution of the OECD benchmark. The solution of the benchmarks essentially contributed to the qualification of the code complex DYN3D-ATHLET as an advanced tool for the accident analysis for both VVER type reactors and Western PWRs.

Dans le contexte de la physique des réacteurs, l’analyse du comportement non stationnaire de la population neutronique avec contre-réactions dans le combustible et dans le modérateur se rend indispensable afin de caractériser les transitoires opérationnels et accidentels dans les systèmes nucléaires et d’en améliorer par conséquent la sûreté. Pour ces configurations non stationnaires, le développement de méthodes Monte-Carlo qui prennent en compte la dépendance en temps du système neutronique, mais aussi le couplage avec les autres physiques, comme la thermohydraulique et la thermomécanique, a pour but de servir de référence aux calculs déterministes.Ce travail de thèse a consisté à mettre en place une chaîne de calcul pour la simulation couplée neutronique Monte-Carlo, avec le code TRIPOLI-4, en conditions non stationnaires et avec prise en compte des contre-réactions thermohydrauliques.Nous avons d'abord considéré les méthodes cinétiques dans TRIPOLI-4, c'est-à-dire avec prise en compte du temps mais sans prise en compte des contre-réactions, en incluant une évaluation des méthodes existantes ainsi que le développement de nouvelles méthodes. Ensuite, nous avons développé un schéma de couplage entre TRIPOLI-4 et le code de thermohydraulique sous-canal SUBCHANFLOW. Enfin, nous avons réalisé une analyse préliminaire de la propagation des incertitudes au sein du calcul couplé sur un modèle simplifié. En effet, les fluctuations statistiques sont inhérentes à notre schéma de par la nature stochastique de TRIPOLI-4. De plus, les équations de la thermohydraulique étant non-linéaires, la propagation des incertitudes au long du calcul doit être étudiée afin de caractériser la convergence du résultat. / One of the main issues for the study of a reactor behaviour is to model the propagation of the neutrons, described by the Boltzmann transport equation, in the presence of multi-physics phenomena, such as the coupling between neutron transport, thermal-hydraulics and thermomecanics. Thanks to the growing computer power, it is now feasible to apply Monte Carlo methods to the solution of non-stationary transport problems in reactor physics, which play an instrumental role in producing reference numerical solutions for the analysis of transients occurring during normal and accidental behaviour.The goal of this Ph. D. thesis is to develop, verify and test a coupling scheme between the Monte Carlo code TRIPOLI-4 and thermal-hydraulics, so as to provide a reference tool for the simulation of reactivity-induced transients in PWRs.We have first tested the kinetic capabilities of TRIPOLI-4 (i.e., time dependent without thermal-hydraulics feedback), evaluating the different existing methods and implementing new techniques. Then, we have developed a multi-physics interface for TRIPOLI-4, and more specifically a coupling scheme between TRIPOLI-4 and the thermal-hydraulics sub-channel code SUBCHANFLOW. Finally, we have performed a preliminary analysis of the stability of the coupling scheme. Indeed, due to the stochastic nature of the outputs produced by TRIPOLI-4, uncertainties are inherent to our coupling scheme and propagate along the coupling iterations. Moreover, thermal-hydraulics equations are non linear, so the prediction of the propagation of the uncertainties is not straightforward.