Durch die Kopplung von Szintillationskristallen an Lichtwellenleiter könnten robuste Messgeräte für die spektroskopitsche Messung von Strahlungsfeldgrößen in Bohrlöchern in Betonstrukturen beim Rückbau von Kernkraftwerken realisiert werden. In dieser Arbeit wurden zwei Prototypen solcher Messgeräte entwickelt, wobei ein auf einem Gadolinium-Aluminium-Gallium-Granat-Szintillationskristall basierendes Messsystem mit einem Kunststofflichtwellenleiter mit hohem Durchmesser aufgrund seiner hohen Zählrate bevorzugt wurde. Mit diesem Messystem konnten spektroskopische Messungen von Gammastrahlung durchgeführt werden. Weiterhin konnte gezeigt werden, dass Kontaminationen einer bestimmten Aktivität bis in einen Abstand von mehreren Zentimetern von der Messsonde im Beton nachweisbar sind. Ein Schätzwert der minimale Messzeit zur Erkennung einer nachweisbaren Kontamination konnte bestimmt werden. Eine hinreichend große, gemessene Ereigniszahl ermöglicht zusätzlich die Bestimmung des Abstands einer radioaktiven, punktförmigen Kontamination eines bekannten Nuklids. Für ein bekanntes Nuklid konnte weiterhin aus der Zählrate ein Schätzwert für die Energiedosis am Ort der Messsonde mithilfe von Referenzmessungen der Dosis mittels Berylliumoxid-Detektoren bestimmt werden.:Inhaltsverzeichnis Abbildungsverzeichnis vii Tabellenverzeichnis ix 1 Einleitung 1 2 Physikalische Grundlagen 3 2.1 Radioaktiver Zerfall 3 2.2 Gröÿen des Strahlungsfelds 3 2.3 Wechselwirkungen von Photonen mit Materie 5 2.4 Lichtentstehung in anorganischen Szintillatoren 8 2.5 Lichtmessung 10 2.6 Pulshöhenverteilungen von Szintillationsdetektoren 12 2.7 Ein ussfaktoren der Lichtausbeute 14 3 Material und Methoden 16 3.1 Aufbau LSO-Sonde und Messungen mit radioaktiven Quellen 16 3.2 Aufbau Lichtkopplung mittels Linse 18 3.3 Aufbau und Versuchsablauf Quarzglaskopplungen 20 3.4 Aufbau GAGG-Sonde 21 3.5 Aufbau des Betonphantoms 24 3.5.1 Dichtebestimmung des Betonphantoms 24 3.6 Messablauf Tiefenkurven in Betonphantom 27 3.7 Datenanalyse mittels Kolmogorv-Smirnov-Test 27 4 Ergebnisse 29 4.1 Spektroskopische Eigenschaften LSO-Sonde 29 4.2 Winkelverteilung austretender Photonen aus Szintillationskristall 33 4.3 Quarzglaskopplung 36 4.4 Spektroskopische Eigenschaften GAGG-Sonde 43 4.5 Vergleich GAGG- und LSO-Sonde 47 4.6 Abstandsabhängigkeit der Zählrate 50 4.7 Abschätzung der maximalen Abschirmdicke von Beton für Kontaminationserkennung 53 4.8 Abschätzung der minimalen Messzeit zur Kontaminationserkennung 56 4.9 Abschirmungsdickenbestimmung mittels Abschirmungsparameter 57 4.10 Bestimmung der Dosis 62 5 Diskussion 64 6 Zusammenfassung 67 / A robust measuring system for spectroscopic measurementes of gamma-ray radiation in boreholes in concrete structures can be built by coupling of a scintillation crystal to a light guide. Two prototypes of such measuring systems are developed one based on a Gadolinium-Aluminium-Gallium-Garnet scintillation crystal with a plastic optical fiber with a high diameter is preferred due to the higher count rate. Spectroscopic measurements of gamma-ray radiation with this measuring system can be performed. It can be shown that contamination of a specific activity can be detected even if they are located in concrete in a distance of several centimeters from the radiation sensor. for the minimal measurement time of 38 s An estimated value to detect a traceable contamination could be determined. If a high number of events can be detected, even the distance between a point like radioaktive source of a known nuclide and the radiation sensor can be determined. An estimated value for the applied dose at the place of the radiation sensor could be determined for a known nuclide with the help of the count rate. Therefore, reference measurements of dose with berylliumoxide detectors were performed.:Inhaltsverzeichnis Abbildungsverzeichnis vii Tabellenverzeichnis ix 1 Einleitung 1 2 Physikalische Grundlagen 3 2.1 Radioaktiver Zerfall 3 2.2 Gröÿen des Strahlungsfelds 3 2.3 Wechselwirkungen von Photonen mit Materie 5 2.4 Lichtentstehung in anorganischen Szintillatoren 8 2.5 Lichtmessung 10 2.6 Pulshöhenverteilungen von Szintillationsdetektoren 12 2.7 Ein ussfaktoren der Lichtausbeute 14 3 Material und Methoden 16 3.1 Aufbau LSO-Sonde und Messungen mit radioaktiven Quellen 16 3.2 Aufbau Lichtkopplung mittels Linse 18 3.3 Aufbau und Versuchsablauf Quarzglaskopplungen 20 3.4 Aufbau GAGG-Sonde 21 3.5 Aufbau des Betonphantoms 24 3.5.1 Dichtebestimmung des Betonphantoms 24 3.6 Messablauf Tiefenkurven in Betonphantom 27 3.7 Datenanalyse mittels Kolmogorv-Smirnov-Test 27 4 Ergebnisse 29 4.1 Spektroskopische Eigenschaften LSO-Sonde 29 4.2 Winkelverteilung austretender Photonen aus Szintillationskristall 33 4.3 Quarzglaskopplung 36 4.4 Spektroskopische Eigenschaften GAGG-Sonde 43 4.5 Vergleich GAGG- und LSO-Sonde 47 4.6 Abstandsabhängigkeit der Zählrate 50 4.7 Abschätzung der maximalen Abschirmdicke von Beton für Kontaminationserkennung 53 4.8 Abschätzung der minimalen Messzeit zur Kontaminationserkennung 56 4.9 Abschirmungsdickenbestimmung mittels Abschirmungsparameter 57 4.10 Bestimmung der Dosis 62 5 Diskussion 64 6 Zusammenfassung 67

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The electric utility companies that own and/or operate the nation’s 112 licensed commercial nuclear power reactors are required by federal law to provide emergency information to residents living around those plants in advance of a nuclear plant emergency. This requires the owner/operators to acknowledge certain risks (e.g., radiation, nuclear plant accidents, evacuation, etc.) that face people living near nuclear generating plants. This dissertation critiques the effectiveness of pre-emergency risk communication strategies by nuclear utilities. Specifically, the dissertation demonstrates that certain historic message themes about nuclear power- termed the "nuclear ethic" -have become embedded in the rhetoric of current nuclear risk communication programs and downplay or mask the seriousness of nuclear plant emergencies, thereby contributing to the apparent ineffectiveness of these communication programs. For example, a survey of residents living around four nuclear plants who receive utility risk communication materials found that nearly two-thirds said they would not follow official instructions in a nuclear plant emergency. Such promotional rhetoric and images remain a part of nuclear risk communication programs because agencies which regulate nuclear power delegate their responsibility for pre-emergency risk communication to the utilities operating the plants. Moreover, there is little involvement in pre-emergency nuclear risk communication by state and local governments. This suggests that risk communication serves a latent symbolic role rather than a functional role for both the regulatory agencies and the utilities by making both groups appear to be isomorphic with societal goals of safety and security for a risky technology. The dissertation concludes by suggesting federal regulatory agencies, and specifically the Federal Emergency Management Agency, intensify their vigilance of risk communication planning and take steps to create authentic two-way communication between the nuclear utilities and the public living near the plants. One way this could be done is by establishing local citizens advisory committees to assess utility risk communication programs and suggest improvements that would help bridge the gap between the nuclear industry’s view of nuclear plant risk and that of the public. / Ph. D.

Multispectral video (MSV) images were used to measure phenological changes in cultivated and uncultivated vegetation communities surrounding the Palo Verde Nuclear Generating Station (PVNGS). Multispectral video imagery was acquired from aircraft on seven dates between the middle of June and the end of September, 1986. Images representing three sites near the PVNGS were selected to calculate Ratio Vegetation Index (RVI) values for seven surface cover types. Mean RVI values were tested sequentially for change, plotted as a function of time, and then compared to a moisture index and the crop calendar. MSV detected changes in cultivated vegetation corresponding to the crop calendar. Changes in natural vegetation and the non-vegetated cover types were also detected, but did not correlate to the moisture index. There is insufficient evidence to determine if detected changes in uncultivated vegetation were the result of phenological changes or electronic noise.

As the United States expands its quantity of nuclear reactors in the near future, the amount of spent nuclear fuel (SNF) will also increase. Closing the nuclear fuel cycle has become the next major technical challenge for the nuclear energy industry. By separating the transuranics (TRU) from the SNF discharged by Light Water Reactors, it is possible to fuel Advanced Burner Reactors to minimize the amount of SNF that must be stored in High Level Waste Repositories. One such ABR concept is the Subcritical Advanced Burner Reactor (SABR) being developed at the Georgia Institute of Technology. SABR is a subcritical, sodium-cooled fast reactor with a fusion neutron source capable of burning up to 25% of the TRU fuel over an 8.2 year residence time. In the SABR concept an annular core with a thickness of 0.6 m and an active height of 3.2 m surrounds the toroidal fusion neutron source. Neutron multiplication varies during the lifetime of the reactor from keff = 0.95 at the beginning of reactor life to 0.83 at the end of an equilibrium fuel cycle. Sixteen control rods worth 9$ are symmetrically positioned around the reactor. This thesis describes the dynamic safety analysis of the coupled neutron source, reactor core and reactor heat removal systems. A special purpose simulation model was written to predict steady-state conditions and accident scenarios in SABR by calculating the coupled evolution of the power output from the fusion and fission cores and the axial and radial temperature distributions of a fuel pin in the reactor. Reactivity Feedback was modeled for Doppler and sodium coolant voiding. SABR has a positive temperature reactivity feedback coefficient. A series of accident scenarios were simulated to determine how much time exists to implement corrective measures during an accident before damage to the reactor occurs.

abstract: This study explores the potential risks associated with the 65 U.S.-based commercial nuclear power plants (NPPs) and the distribution of those risks among the populations of both their respective host communities and of the communities located in outlying areas. First, I examine the relevant environmental justice issues. I start by examining the racial/ethnic composition of the host community populations, as well as the disparities in socio-economic status that exist, if any, between the host communities and communities located in outlying areas. Second, I estimate the statistical associations that exist, if any, between a population's distance from a NPP and several independent variables. I conduct multivariate ordinary least square (OLS) regression analyses and spatial autocorrelation regression (SAR) analyses at the national, regional and individual-NPP levels. Third, I construct a NPP potential risk index (NPP PRI) that defines four discrete risk categories--namely, very high risk, high risk, moderate risk, and low risk. The NPP PRI allows me then to estimate the demographic characteristics of the populations exposed to each so-defined level of risk. Fourth, using the Palo Verde NPP as the subject, I simulate a scenario in which a NPP experiences a core-damage accident. I use the RASCAL 4.3 software to simulate the path of dispersion of the resultant radioactive plume, and to investigate the statistical associations that exist, if any, between the dispersed radioactive plume and the demographic characteristics of the populations located within the plume's footprint. This study utilizes distributive justice theories to understand the distribution of the potential risks associated with NPPs, many of which are unpredictable, irreversible and inescapable. I employ an approach that takes into account multiple stakeholders in order to provide avenues for all parties to express concerns, and to ensure the relevance and actionability of any resulting policy recommendations. / Dissertation/Thesis / Ph.D. Environmental Social Science 2014

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2016-06-22T14:12:07Z No. of bitstreams: 0 / Made available in DSpace on 2016-06-22T14:12:07Z (GMT). No. of bitstreams: 0 / Após os acidentes nucleares ocorridos no mundo, critérios e requisitos extremamente rígidos para a operação das instalações nucleares foram determinados pelos órgãos internacionais que regulam essas instalações. A partir da ocorrência destes eventos, as operadoras de plantas nucleares necessitam simular alguns acidentes e transientes, por meio de programas computacionais específicos, para obter a licença de operação de uma planta nuclear. Com base neste cenário, algumas ferramentas computacionais sofisticadas têm sido utilizadas como o Reactor Excursion and Leak Analysis Program (RELAP5), que é o código mais utilizado para a análise de acidentes e transientes termo-hidráulicos em reatores nucleares no Brasil e no mundo. Uma das maiores dificuldades na simulação usando o código RELAP5 é a quantidade de informações geométricas da planta necessárias para a análise de acidentes e transientes termo-hidráulicos. Para a preparação de seus dados de entrada é necessário um grande número de operações matemáticas para calcular a geometria dos componentes. Assim, a fim de realizar estes cálculos e preparar dados de entrada para o RELAP5, um pré-processador matemático amigável foi desenvolvido, neste trabalho. O Visual Basic for Applications (VBA), combinado com o Microsoft Excel, foi utilizado e demonstrou ser um instrumento eficiente para executar uma série de tarefas no desenvolvimento desse pré-processador. A fim de atender as necessidades dos usuários do RELAP5, foi desenvolvido o Programa de Cálculo do RELAP5 PCRELAP5 onde foram codificados todos os componentes que constituem o código, neste caso, todos os cartões de entrada inclusive os opcionais de cada um deles foram programados. Adicionalmente, uma versão em inglês foi criada para PCRELAP5. Também um design amigável do PCRELAP5 foi desenvolvido com a finalidade de minimizar o tempo de preparação dos dados de entrada e diminuir os erros cometidos pelos usuários do código RELAP5. Nesse trabalho, a versão final desse pré-processador foi aplicada com sucesso para o Sistema de Injeção de Emergência (SIE) da usina Angra 2. / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Submitted by Pedro Silva Filho (pfsilva@ipen.br) on 2017-11-21T11:45:58Z No. of bitstreams: 0 / Made available in DSpace on 2017-11-21T11:45:58Z (GMT). No. of bitstreams: 0 / O gerenciamento de riscos em uma instalação nuclear é necessário para a segurança de trabalhadores e de populações vizinhas. Parte desse processo é a comunicação dos riscos que propicia o diálogo entre gestores da empresa e moradores das áreas de risco. A população que conhece os riscos a que está exposta, como esses riscos são gerenciados e o que deve ser feito em uma situação de emergência tende a se sentir mais segura e a confiar nas instituições responsáveis pelo plano de emergência. Sem diálogo entre empresa e público, o conhecimento dos procedimentos a serem seguidos em caso de acidente não chega à população, ou quando chega, não há confiança dessas pessoas na sua eficácia. Em Angra dos Reis, no litoral sul do Estado do Rio de Janeiro, está a Central Nuclear Almirante Álvaro Alberto. No entorno dessa Central Nuclear existe uma população que, de acordo com o Plano de Emergência Externo (PEE/RJ), deverá ser evacuada ou ficar abrigada, caso ocorra um acidente na instalação. Um trabalho de comunicação de riscos entre esses moradores é necessário para que eles conheçam o plano de emergência e os procedimentos corretos para uma situação de emergência, além de buscar esclarecer dúvidas e mitos. Esse trabalho apresenta uma análise da comunicação dos riscos feita para a população local, a percepção que ela tem dos riscos e o grau de conhecimento do plano de emergência externo por parte dessas pessoas. / Tese (Doutorado em Tecnologia Nuclear) / IPEN/T / Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP

The thesis deals with the analysis of possibilities for use of III. and IV. generation of nuclear reactors for the Czech Republic. Introduction deals with basic principles of nuclear energetics and fuel cycle which is closely associated with the issue and its development and application in practice will have a significant effect both for choice of future types of nuclear reactors of IV. generation and the matter of settlement with spent nuclear fuel. In other parts the thesis deals with the present state of nuclear energetics in the Czech Republic, compares concrete chosen types of nuclear power plants of III. generation and generally describes nuclear reactors and power plants of IV. generation. Final part of the work is focused on questions concerning the possibility of construction of new blocks of nuclear power plants in the Czech Republic and presents concrete recommendation from author´s point of view.