An in-core experiment is proposed to measure all delayed-neutron parameters (λᵢ and βᵢ for each delayed group) and the neutron generation time applicable for a given reactor system. The method uses a least-squares fitting technique to simultaneously fit a series of transient responses produced by step changes (or rapid ramps) in reactivity of arbitrary size. The function which is least-squares fit is the exact analytic solution describing a reactor response following a step change in reactivity as given by the standard multi-group point-reactor model. The method does not require any knowledge of the absolute reactivity of the system. The results are based solely upon the measurable quantities of relative power, time, and asymptotic inverse period. Unlike out-of-core delayed neutron experiments, the results are independent of the total number of fissions that have occurred in the fissioning isotope that produced the delayed neutrons, and are independent of the efficiency of the detector used to monitor the transients. Modal contamination is presumed not to be a serious problem in small or intermediate size cores as long as small reactivity changes are used (i.e. +$.10 to -$.25), which is consistent with the best method of performing this experiment. The method is equally applicable to both homogeneous and heterogeneous reactors and may be used in either thermal or fast systems. The number of delayed-neutron groups is not a constraint.

No description available.

Thesis (M.S.)--University of Michigan, 1983. / Project completed in 1981. Degree awarded in 1983.

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L'énergie nucléaire est l'une des plus propres en matière d'émission de gaz à effet de serre et, malgré ses atouts, elle n'est développée que dans quelques pays du monde. La sûreté reste une question ouverte pour l'avenir de cette énergie après l'accident de Fukushima. En France, la loi de 2006 sur la gestion des déchets soutient le développement d'une nouvelle génération de réacteurs nucléaires et du prototype de Réacteur Technologiquement Avancé au Sodium pour la Démonstration Industrielle (projet ASTRID) qui vise à apporter une réponse industrielle et technologique à de nombreux enjeux de ce siècle. L'une des préoccupations de la technologie du Réacteur à Neutrons Rapide et caloporteur sodium (RNR Na) est la perte de ce dernier car elle pourrait entraîner un emballement de la réaction en chaîne si l'effet en réactivité de vidange sodium (SVRE) est positif. Lorsque le sodium est retiré du cœur, deux effets antagonistes se produisent qui affectent l'équilibre neutronique: l'un augmente la réactivité du cœur et est appelé la composante centrale (CC) et l'autre est la composante de fuite (LC) avec un effet négatif sur la réactivité. Maximiser la dernière composante est l'une des réponses pour augmenter la sûreté inhérente aux RNR-Na. C'est pourquoi le CEA a développé un concept de cœur innovant: le «Cœur à Faible Vidange» (CFV) qui donne une SVRE négatif. Cependant, de telles innovations doivent être validées expérimentalement et l'incertitude sur cet effet en réactivité doit être maîtrisée. En soutien au développement des RNR Na : la base de données expérimentale existante est assez importante (PRE-RACINE, CIRANO, BFS). / The nuclear energy is one of the cleanest energy in regard of greenhouse gas emission and despite its assets is only developed in few countries in the world. Safety remains an open issue for the future of this energy after the Fukushima accident . In France the 2006 law on the waste management ensures the development of a new generation of nuclear reactor and has lead to the Advanced Sodium Technology Reactor for Industrial Demonstration (ASTRID) which aims to bring an industrial and technological advanced answer to many issues of this century. One of the concerns in the sodium cooled fast reactor (SFR) technology is the loss of sodium coolant accident because it might lead to a snowball effect in the chain reaction if the sodium void reactivity effect (SVRE) is positive. When the sodium is removed from the core, two antagonistic effects arise that affect the neutron balance: one increases the reactivity of the core and is called the central component (CC) and the other is the leakage component (LC) with a negative feedback on the reactivity. Maximizing the last component is one of the answer to increase the inherent safety of the SFRs. That is why the CEA has developed an innovative core design: the “Cœur à Faible Vidange” (CFV : Core with low void effect) which exhibits a negative SVRE. However, such innovations have to be experimentally validated and the uncertainty on this reactivity effect has to be mastered. In support of SFRs the existing experimental data base is quite large (PRE-RACINE, CIRANO, BFS).

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Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2016-06-22T12:55:19Z No. of bitstreams: 0 / Made available in DSpace on 2016-06-22T12:55:19Z (GMT). No. of bitstreams: 0 / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2016-06-22T12:55:19Z No. of bitstreams: 0 / Made available in DSpace on 2016-06-22T12:55:19Z (GMT). No. of bitstreams: 0 / O objetivo do presente trabalho foi o de estabelecer uma nova característica ao programa de Monte Carlo ESQUEMA, destinado a simular todo o processo de detecção em um sistema de coincidência 4π(β,EC)-γ e possibilitou o conhecimento antecipado do comportamento da atividade aparente da fonte radioativa como função da eficiência do detector 4π. Esta metodologia foi aplicada anteriormente para os casos onde a eficiência é variada pela colocação de absorvedores sobre a fonte radioativa ou variando-se o nível de discriminação eletrônica, com a exclusão completa das contagens abaixo deste nível. No presente trabalho, esta metodologia foi modificada com a inclusão dos eventos não coincidentes abaixo do nível de discriminação eletrônica, adotado para a via beta. Desta forma, a curva de extrapolação passa a ter um comportamento rigorosamente linear e o valor extrapolado corresponde a atividade da fonte radioativa, esperando-se obter uma melhor acurácia no seu valor. Esta alteração é inovadora e particularmente apropriada para ser aplicada ao caso de medidas com o Sistema de Coincidências por Software (SCS), desenvolvido recentemente pelo LMN. O trabalho consiste na modificação de um programa desenvolvido no LMN, denominado ESQUEMA, que leva em conta as características gerais de um esquema de desintegração complexo, incluindo-se os diversos tipos de transição, além das características experimentais do sistema de medidas por coincidência 4π(β,EC)-γ. Estas modificações permitem a obtenção das curvas de extrapolação, aplicando-se o método proposto de Discriminação com Reposição. O software MCNP5 é usado para simular o transporte de radiação. A simulação de Monte Carlo é comparada com valores experimentais para o radionuclídeo padronizados no LMN, o 111In . / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP