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Modular remote radiation monitor
Get PDFThe Modular Remote Radiation Monitor (MRRM) is a novel radiation monitor suitable for monitoring environmental exposure to ionizing radiation. It is a portable compact-size low-power microprocessor-based electronic device which provides its monitoring data to other electronic systems, physically distant from it, by means of an electronic communication channel, which can be wired or wireless according to the requirements of each application. Besides its low-power highly-integrated circuit design, the Modular Remote Radiation Monitor is presented in a modular architecture, which promotes full compliance to the technical requirements of different applications while minimizing cost, size and power consumption. Its communication capability also supports the implementation of a network of multiple radiation monitors connected to a supervisory system, capable of remotely controlling each monitor independently as well as visualizing the radiation levels from all monitors. A prototype of the MRRM, functionally equivalent to the MRA-7027 radiation monitor, was implemented and connected to a wired MODBUS network of MRA-7027 monitors, responsible for monitoring ionizing radiation inside Argonauta reactor room at Instituto de Engenharia Nuclear. Based on the highly positive experimental results obtained, further design is currently underway in order to produce a consumer version of the MRRM
Modeling the critical safety functions status tree of NPP using FPGA
Get PDFField programmable gate arrays (FPGAs) based systems and equipment are beginning to appear in new plants I&C applications, as well as in retrofits for operating plants, in particular for safety applications due to their capability to face the systems obsolescence since they are circuit independent. The circuits implemented can be portable to different FPGAs architectures. Moreover, they reduce complexity for regulatory approval as compared to conventional microprocessor-based systems. Critical safety function (CSF) is the most significant design concept for prioritize operator actions for NPP based on the potential threat to the three barriers (fuel cladding, primary coolant system boundary, and containment) and allows the operator to respond to these threats prior to event diagnosis. CSF has a hierarchical information structure that organizes the system variables affecting the plant safety in terms of goal-means relations. This paper describes the application of FPGA in the implementation of the CSFs status tree logic for a Westinghouse 3-loops NPP simulator
Basic CFD investigation of decay heat removal in a pool type research reactor
Get PDFSafety is one of the most important and desirable characteristic in a nuclear plant. Natural circulation cooling systems are noted for providing passive safety. These systems can be used as mechanism for removing the residual heat from the reactor, or even as the main cooling system for heated sections, such as the core. In this work, a computational fluid-dynamics (CFD) code is used to simulate the process of natural circulation in an open pool research reactor after its shutdown. The physical model studied is similar to the Open Pool Australian Light water reactor (OPAL), and contains the core, cooling pool, reflecting tank, circulation pipes and chimney. For best computing performance, the core region was modeled as a porous media, where the parameters were obtained from a separately detailed CFD analysis
Hardware Architecture For Radionuclide Identification.
No full textA radioactividade é a emissão espontânea de energia a partir de átomos instáveis. As fontes radioactivas têm radionuclídeos. Radionuclide sofre decaimento radioativo e emite raios gama e partículas subatômicas, constituindo a radiação ionizante. A energia do raio gama de um radionuclídeo é utilizado para determinar a identidade de emissores gama presentes na fonte. Este artigo descreve uma arquitetura de hardware para executar a identificação de radionuclídeos. Também mostramos uma comparação entre o desempenho da arquitectura proposta e outras soluções.Radioactivity is the spontaneous emission of energy from unstable atoms. Radioactive sources have radionuclides. Radionuclide undergoes radioactive decay and emits gamma rays and subatomic particles, constituting the ionizing radiation. The gamma ray energy of a radionuclide is used to determine the identity of gamma emitters present in the source. This paper describes a hardware architecture to perform radionuclide identification. We also show a comparison between the performance of the proposed architecture and other solutions
Estudo de confiabilidade em reatores de pesquisa
No full textÁrvores de falhas e árvores de eventos são amplamente utilizadas nas indústrias para modelar e calcular a confiabilidade dos sistemas de segurança. Análises detalhadas nas instalações nucleares requerem a combinação dessas duas técnicas. O presente trabalho utiliza as metodologias de AF (Árvore de Falhas) e AE (Árvore de Eventos) para fazer um estudo sobre APS (Avaliação Probabilística de Segurança) em reatores de pesquisa. A APS segundo a AIEA (Agência Internacional de Energia Atômica) é dividida em Nível 1, Nível 2, e Nível 3. No Nível 1, conceitualmente os sistemas de segurança atuam para evitar a ocorrência de acidentes, no Nível 2, ocorrido o acidente, procura-se minimizar as consequências deste, sendo conhecido como fase de gerenciamento do acidente, e no Nível 3 são mensurados os impactos e desdobramentos do acidente. Esta dissertação concentra os estudos no Nível 1, e busca através da aquisição do conhecimento a consolidação das metodologias para futuros estudos de confiabilidade. O reator de pesquisa grego, GRR-1, foi utilizado como caso exemplo. O LOCA (Loss of Coolant Accident) foi escolhido como evento iniciador e a partir daí foram desenvolvidos os possíveis desdobramentos do acidente (sequência de eventos), utilizando AE, que poderiam causar danos ao núcleo. Além disso, para cada um dos sistemas afetados ou envolvidos no acidente, foram construídas AF e calculadas as probabilidades de cada evento topo das AF. Estimativas das medidas de importância dos eventos básicos estão presentes. Os estudos desta pesquisa foram conduzidos utilizando a ferramenta computacional comercial SAPHIRE. Os resultados assim obtidos, para atuação ou a falha dos sistemas analisados, foram considerados satisfatórios.Fault trees and event trees are widely used in industry to model and to evaluate the reliability of safety systems. Detailed analyzes in nuclear installations require the combination of these two techniques. This study uses the methods of FT (Fault Tree) and ET (Event Tree) to accomplish the PSA (Probabilistic Safety Assessment) in research reactors. According to IAEA (International Atomic Energy Agency), the PSA is divided into Level 1, Level 2 and Level 3. At the Level 1, conceptually, the security systems perform to prevent the occurrence of accidents, At the Level 2, once accidents happened, this Level seeks to minimize consequences, known as stage management of accident, and at Level 3 accident impacts are determined. This study focuses on analyzing the Level 1, and searching through the acquisition of knowledge, the consolidation of methodologies for future reliability studies. The Greek Research Reactor, GRR-1, is a case example. The LOCA (Loss of Coolant Accident) was chosen as the initiating event and from it, using ET, possible accidental sequences were developed, which could lead damage to the core. Moreover, for each of affected systems, probabilities of each event top of FT were developed and evaluated in possible accidental sequences. Also, the estimates of importance measures for basic events are presented in this work. The studies of this research were conducted using a commercial computational tool SAPHIRE. Additionally, achieved results thus were considered satisfactory for the performance or the failure of analyzed systems. . Keywords
