Dissertations / Theses on the topic 'Nuclear reactor monitoring'

A significant proportion of the UK energy needs are currently serviced by a fleet of ageing nuclear reactors. Ensuring that these reactors are operated safely is the highest priority and the structural health of their cores, that provide channels for control rods and coolant gas, is a key aspect. This thesis focuses on the application of structuralhealth monitoring to the graphite reactor cores used in the UK and presents a specification for the use of structural health monitoring (SHM) techniques already es- tablished in bridge and aircraft monitoring, with data obtained through existing reactor monitoring processes. This approach utilises statistical and clustering techniques on monitoring data that can be acquired during online operation of the plant. The use of existing monitoring processes to complement the established inspection regime for nuclear reactors is a novel contribution from this work. As part of proving the SHM approach, this thesis reports on work undertaken to identify suitable data and numerical limits for the cluster analysis. This analysis considers the data with respect to the stated aim of detectin~ core distortion and demonstrates that the chosen data and values are acceptable and conservative in the context of reactor condition monitoring. An assessment of the SHM solution is presented describing the im- plementation of the SHM approach using a multi-agent system (MAS), IMAPS. This implementation required consideration of using MAS tech- nology for condition monitoring, and the novel contribution of a technique for storing and retrieving historical data in a manner concomitant with both MAS and relational database theory is presented.ij The thesis concludes that condition monitoring is feasible on the graphite cores, and that multi-variate analysis through SHM implemented within a MAS offers a storage and analysis platform that can both handle the data volumes and accommodate further extensions as required.

Nuclear power stations worldwide are exceeding their originally specified design lives and with only limited construction of new generation underway, there is a desire to continue the operation of existing stations to ensure electricity supply. Continued operation of nuclear power stations with degrading and life-limiting components necessitates increased monitoring and inspection, particularly of the reactor cores, to ensure they are safe to operate. The monitoring of a large number of components and their related data sources is a distributed and time consuming process for the engineer given the lack of infrastructure available for collecting, managing and analysing monitoring data. This thesis describes the issues associated with nuclear Condition Monitoring (CM) and investigates the suitability of a distributed framework utilising intelligent software agents to collect, manage and analyse data autonomously. The application of data fusion techniques is examined to estimate unre corded parameters, provide contextualisation for anomalies in order to quickly identify true faults from explainable anomalies and to extract more detail from existing CM data. A generalised framework is described for nuclear CM of any type of reactor, specifying the required components and capabilites based on the design of a suitable Multi Agent System, including the interaction of the framework with existing CM systems and human users. A high level ontology for nuclear CM is proposed and is emphasised as a crucial aspect of the data management and extendability of the framework to incorporate further data sources and analyses. A prototype system, based on the generalised framework is developed for the case of the Advanced Gas-cooled Reactor, with new and existing CM analyses formalised within intelligent agents. Using real station data and simulated fault data, the prototype system was shown to be capable of performing the existing monitoring tasks considerably faster than a human user while retaining all data and analyses for justification and traceability of decisions based on the analyses.

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

The main objective of this research is to develop a fault detection and isolation (FDI) methodologies for Cylla-Haase polymerization reactor, and implement the developed methods to the nonlinear simulation model of the proposed reactor to evaluate the effectiveness of FDI methods. The first part of this research focus of this chapter is to understand the nonlinear dynamic behaviour of the Chylla-Haase polymerization reactor. In this part, the mathematical model of the proposed reactor is described. The Simulink model of the proposed reactor is set up using Simulink/MATLAB. The design of Simulink model is developed based on a set of ordinary differential equations that describe the dynamic behaviour of the proposed polymerization reactor. An independent radial basis function neural networks (RBFNN) are developed and employed here for an on-line diagnosis of actuator and sensor faults. In this research, a robust fault detection and isolation (FDI) scheme is developed for open-loop exothermic semi-batch polymerization reactor described by Chylla-Haase. The independent (RBFNN) is employed here when the system is subjected to system uncertainties and disturbances. Two different techniques to employ RBF neural networks are investigated. Firstly, an independent neural network is used to model the reactor dynamics and generate residuals. Secondly, an additional RBF neural network is developed as a classifier to isolate faults from the generated residuals. In the third part of this research, a robust fault detection and isolation (FDI) scheme is developed to monitor the Chylla-Haase polymerization reactor, when it is under the cascade PI control. This part is really challenging task as the controller output cannot be designed when the reactor is under closed-loop control, and the control action will correct small changes of the states caused by faults. The proposed FDI strategy employed a radial basis function neural network (RBFNN) in an independent mode to model the process dynamics, and using the weighted sum-squared prediction error as the residual. The Recursive Orthogonal Least Squares algorithm (ROLS) is employed to train the model to overcome the training difficulty of the independent mode of the network. Then, another RBFNN is used as a fault classifier to isolate faults from different features involved in the residual vector. In this research, an independent MLP neural network is implemented here to generate residuals for detection task. And another RBF is applied for isolation task performing as a classifier. The fault diagnosis scheme is developed for a Chylla-Haase reactor under open-loop and closed-loop control system. The comparison between these two neural network architectures (MPL and RBF) are shown that RBF configuration trained by (RLS) algorithm have several advantages. The first one is greater efficiency in finding optimal weights for field strength prediction in complex dynamic systems. The RBF configuration is less complex network that results in faster convergence. The training algorithms (RLs and ROLS) that used for training RBFNN in chapter (4) and (5) have proven to be efficient, which results in significant faster computer time in comparison to back-propagation one. Another fault diagnosis (FD) scheme is developed in this research for an exothermic semi-batch polymerization reactor. The scheme includes two parts: the first part is to generate residual using an extended Kalman filter (EKF), and the second part is the decision making to report fault using a standardized hypothesis of statistical tests. The FD simulation results are presented to demonstrate the effectiveness of the proposed method. In the lase section of this research, a robust fault diagnosis scheme for abrupt and incipient faults in nonlinear dynamic system. A general framework is developed for model-based fault detection and diagnosis using on-line approximators and adaptation/learning schemes. In this framework, neural network models constitute an important class of on-line approximators. The changes in the system dynamics due to fault are modelled as nonlinear functions of the state, while the time profile of the fault is assumed to be exponentially developing. The changes in the system dynamics are monitored by an on-line approximation model, which is used for detecting the failures. A systematic procedure for constructing nonlinear estimation algorithm is developed, and a stable learning scheme is derived using Lyapunov theory. Simulation studies are used to illustrate the results and to show the effectiveness of the fault diagnosis methodology. Finally, the success of the proposed fault diagnosis methods illustrates the potential of the application of an independent RBFNN, an independent MLP, an Extended kalman filter and an adaptive nonlinear observer based FD, to chemical reactors.

As the Advanced Gas-cooled Reactor (AGR) nuclear power stations are ageing, the nuclear core composed by graphite bricks can distort. The direct measurement of the core condition is costly and time-consuming, hence, alternative methods have been developed to provide the necessary information about the core condition. This thesis presents a model-based technique for condition monitoring of AGRs cores using measurements obtained during routine core refuelling process. It has been demonstrated that Fuel Grab Load Trace (FGLT) data gathered during refuelling operations provides, through the magnitude of its friction component, information relating to the condition of the graphite bricks. Therefore, the condition monitoring of an AGR leads to the estimation of the friction force resulting from the interaction of the fuel assembly and the core channel. To this end the main objective of this work is to investigate estimation techniques that are needed in industrial applications and in particular can be used in the refuelling filtering problem. As a result of this study, a novel LPV estimator and robust estimator have been designed and implemented. A model for the refuelling system was initially developed from the first principles of the process. Then its fuel assembly dynamics subsystem was identified to be used in a model based filtering application. Finally a H robust estimator was employed to estimate the friction force to be used for the core condition analysis.

Since the first observation of antineutrinos from beta decay of the fission products inside a nuclear reactor in 1956, the design and operating experience of antineutrino detectors near reactors has increased to the point where monitoring the reactor's power level and progression through its burnup cycle has become possible. With the expected increase in world nuclear energy capacity, including the dissemination of reactor technologies to non-nuclear states, the need for safeguards measures which are able to provide continuous, near-real-time information about the state of the core, including its isotopic composition, in a tamper- and spoof-resistant manner is evident. Near-field (~20 m from the core) antineutrino detectors are able to fulfill this demand without perturbing normal reactor operation, without requiring instrumentation which penetrates the reactor vessel, and without displacing other plant structures. Two sodium-cooled long-life fast reactors that are characteristic of next-generation reactors which are attractive for installation in non-nuclear states, one large and one small power rating, have been modeled throughout their reference burnup cycles using MCC-3 and DIF3D/REBUS. Various diversions of fissile material from the core designed to obtain weapons-usable material for the purpose of nuclear proliferation were studied as perturbed core states. The difference in detector event rates between the reference and perturbed states was used to determine the probability that a particular diversionary activity would be apparent before the material could be converted into a weapon. These data indicate which types of diversion antineutrino safeguards are particularly strong against and how the technology might be implemented in current and future international policies concerning nuclear proliferation.

Accelerator-driven systems have been proposed for incineration of transuranic elements from spent nuclear fuel. For safe operation of such facilities, a robust method for reactivity monitoring is required. Experience has shown that the performance of reactivity measurement methods in terms of accuracy and applicability is highly system dependent. Further investigations are needed to increase the knowledge data bank before applying the methods to an industrial facility and to achieve license to operate such a facility. In this thesis, two systems have been subject to investigation of various reactivity measurement methods. Conditions for successful utilization of the methods are presented, based on the experimental experience. In contrast to previous studies in this field, the reactivity has not only been determined, but also monitored based on the so called beam trip methodology which is applicable also to non-zero power systems. The results of this work constitute a part of the knowledge base for the definition of a validated online reactivity monitoring methodology for facilities currently being under development in Europe (XT-ADS and EFIT).
QC 20100621

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Thesis (PhD)--Stellenbosch University, 2012.
ENGLISH ABSTRACT: The objective of this research was to develop a neutron detection system that incorporates a mass spectrometer to measure high neutron flux in a nuclear reactor environment. This system consists of slow and fast neutron detector elements for measuring fluxes in those energy regions respectively. The detector should further be capable of withstanding the harsh conditions associated with a high temperature reactor. This novel detector which was initially intended for use in the PBMR reactor has possible applications as an in-core neutron and indirect temperature-monitoring device in any of the HTGR. Simulations of a generic HTGR core model were performed in order to obtain the neutron energy spectrum with emphasis on the behavior of three energy regions, slow, intermediate and fast neutrons within the core at different temperatures. The slow neutron flux which has the characteristic of a Maxwell- Boltzmann distribution were found to shift to larger values of neutron flux at higher energies as the fuel temperature increased, while fast neutron flux spectra remained relatively constant. In addition, the results of the fit of the slow neutron flux with a modified Maxwell-Boltzmann equation confirmed that in the presence of the neutron source, leakage and absorption, the effective neutron temperatures is above the medium temperatures. From these results, it was clear that the detection system will need to monitor both slow and fast neutron flux. Placing neutron detectors inside the reactor core, that are sensitive to a particular energy range of slow and fast neutrons, would thus provide information about the change of temperature in the fuel and hence act as an in-core temperature monitor. A detection mechanism was developed that employs the neutron-induced break-up reaction of 6Li and 12C into α-particles. These materials make excellent neutron converters without interference due to γ-rays, as the contributions from 6Li(γ,np)4He and 12C(γ,3α) reactions are negligible. The mass spectrometer measures the 4He partial pressure as a function of time under high vacuum with the help of pressure gradient provided by a high-vacuum turbomolecular pump and a positive-displacement fore-vacuum pump connected in series. A cryogenic trap, which contains a molecular sieve made of pellets 1.6 mm in diameter, was also designed and manufactured to remove impurities which cause a background in the lighter mass region of the spectrum. The development and testing of the high flux neutron detection system were performed at the iThemba Laboratory for Accelerator Based Sciences (LABS), South Africa. These tests were carried out with a high energy proton beam at the D-line neutron facility, and with a fast neutron beam at the neutron radiation therapy facility. To test the principle and capability of the detection system in measuring high fluxes, a high intensity 66 MeV proton beam was used to produce a large yield of α-particles. This was done because the proton inelastic scattering cross-section with 12C nuclei is similar to that of neutrons, with a threshold energy of about 8 MeV for both reactions. Secondly, the secondary fast neutrons produced from the 9Be(p,n)9B reaction were also measured with the fast neutron detector. The response of this detection system during irradiation was found to be relatively fast, with a rise time of a few seconds. This is seen as a sharp increase in the partial pressure of 4He gas as the proton or neutron beam bombards the 12C material. It was found that the production of 4He with the proton beam was directly proportional to the beam intensity. The number of 4He atoms produced per second was deduced from the partial pressure observed during the irradiation period. With a neutron beam of 1010 s−1 irradiating the detector, the deduced number of 4He atoms was 109 s−1. When irradiation stops, the partial pressure drops exponentially. This response is attributed to a small quantity of 4He trapped in the present design. Overall, the measurements of 4He partial pressure produced during the tests with proton and fast neutron beams were successful and demonstrated proof of principle of the new detection technique. It was also found that this system has no upper neutron flux detection limit; it can be even higher than 1014 n·cm−2·s−1. The lifetime of this detection system in nuclear reactor environment is practically unlimited, as determined by the known ability of stainless steel to keeps its integrity under the high radiation levels. Hence, it is concluded that this high flux neutron detection system is excellent for neutron detection in the presence of high γ-radiation level and provides real-time flux measurements.
AFRIKAANSE OPSOMMING: Die doel van hierdie navorsing was om ’n neutrondetektorstelsel te ontwikkel wat hoë neutronvloed binne in ’n kernreaktor kan meet. Die stelsel bevat twee aparte detektorelemente sodat die termiese sowel as snelneutronvloed gemeet kan word. Die detektor moet verder in staat wees om die strawwe toestande, kenmerkend aan ’n hoë temperatuur reaktor, te kan weerstaan. Die innoverende detektorstelsel, oorspronklik geoormerk vir gebruik in die PBMR reaktor, het toepassingsmoontlikhede as in-kern neutron- sowel as indirekte temperatuurmonitor. Simulasies van ’n generiese model van ’n HTGR reaktorkern is uitgevoer ten einde die neutronenergiespektrum in die kern by verskillende temperature te bekom met klem op die gedrag van neutrone in drie energiegroepe: stadig (termies), intermediêr en snel (vinnig). Daar is bevind dat die stadige neutrone, wat ’n Maxwell-Boltzman verdeling toon, in intensiteit toeneem en dat die piek na hoër energie verskuif met toename in temperatuur, terwyl die vinnige neutronspektrum relatief onveranderd bly. ’n Passing van die stadige spektrum op ’n gemodifiseerde Maxwell-Boltzmann verdeling het bevestig dat die effektiewe neutrontemperatuur weens die teenwoordigheid van bronterme, verliese en absorpsie, hoër as die temperatuur van die medium is. Hierdie resultate maak dit duidelik dat die detektorstelsel beide die stadige sowel as die vinnige neutronvloed moet kan waarneem. Deur detektorelemente wat sensitief is vir die onderskeie spekrale gebiede in die reaktorhart te plaas, kan informasie bekom word wat tot in-kern temperatuur herleibaar is sodat die stelsel inderdaad as indirekte temperatuurmonitor kan dien. Die feit dat alfa-deeltjies geproduseer word in neutron-geïnduseerde opbreekreaksies van 6Li en 12C is as die basis van die nuwe opsporingsmeganisme aangewend. Hierdie materiale funksioneer uitstekend as neutron-selektiewe omsetters in die teenwoordigheid van gamma-strale aangesien laasgenoemde se bydraes tot helium produksie via die 6Li(γ,np)4He en 12C(γ,3α) reaksies, weglaatbaar is. Die massaspektrometer meet die tydgedrag van die 4He parsiële druk binne ’n hoogvakuum wat met behulp van ’n seriegeskakelde kombinasie van ’n turbomolekulêre en positiewe-verplasingsvoorpomp verkry word. ’n Koueval met ’n molekulêre sif, bestaande uit 1.6 mm diameter korrels, is ontwerp en vervaardig om onsuiwerhede te verwyder wat andersins as agtergrond by die ligter gedeelte van die massaspektrum sou wys. Die ontwikkeling en toetsing van die hoëvloed detektorstelsel is te iThembaLABS (iThemba Laboratories for Accelerator Based Sciences) gedoen. Dit is uitgevoer deur gebruik te maak van die hoë energie protonbundel van die D-lyn neutronfasiliteit asook van die bundel vinnige neutrone by die neutronterapiefasiliteit. Om die beginsel en vermoë te toets om by ’n hoë neutronvloed te kan meet, is van die intense 66 MeV protonbudel gebruik gemaak om ’n hoë opbrengs alfa-deeltjies te verkry. Dit is gedoen omdat die reaksiedeursnit vir onelastiese verstrooiing van protone vanaf 12C kerne soortgelyk is aan die van neutrone, met ’n drumpelenergie van 8 MeV vir beide reaksies. Tweedens is die sekondêre vinnige neutrone afkomstig van die 9Be(p,n)9B reaksie ook met die neutrondetektor gemeet. Daar is bevind dat die reaksietyd van die deteksiestelsel tydens bestraling relatief vinnig is, soos gekenmerk deur ’n stygtyd van etlike sekondes. Laasgenoemde manifesteer as ’n toename in die parsiële druk van die 4He sodra die proton- of neutronbundel op die 12C teiken inval. Daar is verder bevind dat die 4He produksie direk eweredig aan die bundelintensiteit is. Vir ’n neutronbundel van nagenoeg 1010 s−1, invallend op die neutrondetektor, is vanaf die gemete parsiële druk afgelei dat die produksie van 4He atome sowat 109 s−1 beloop. In die geheel beoordeel, was die meting van die 4He parsiële druk tydens die toetse met vinnige protone en neutrone suksesvol en het dit die nuwe meetbeginsel bevestig. Dit is verder bevind dat die meetstelsel nie ’n beperking op die boonste neutronvloed plaas nie, maar dat dit vloede van selfs hoër as 1014 s−1 kan hanteer. Die leeftyd van die detektorstelsel in die reaktor is prakties onbeperk en onderhewig aan die bevestigde integriteit van vlekvrystaal onder hoë bestraling. Die gevolgtrekking is dus dat die nuwe detektorstelsel uitstekend geskik is vir die in-tyd meting van ’n baie hoë vloed van neutrone ook in die teenwoordigheid van intense gammabestraling.

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Tese (Doutoramento)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

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Um desafio recorrente em processos produtivos é o desenvolvimento de sistemas de monitoração e diagnóstico. Esses sistemas ajudam na detecção de mudanças inesperadas e interrupções, prevenindo perdas e mitigando riscos. Redes Neurais Artificiais (RNA) têm sido largamente utilizadas na criação de sistemas de monitoração. Normalmente as RNA utilizadas para resolver este tipo de problema são criadas levando-se em conta apenas parâmetros como o número de entradas, saídas e quantidade de neurônios nas camadas escondidas. Assim, as redes resultantes geralmente possuem uma configuração onde há uma total conexão entre os neurônios de uma camada e os da camada seguinte, sem que haja melhorias em sua topologia. Este trabalho utiliza o algoritmo de Otimização por Colônia de Formigas (OCF) para criar redes neurais otimizadas. O algoritmo de busca OCF utiliza a técnica de retropropagação de erros para otimizar a topologia da rede neural sugerindo as melhores conexões entre os neurônios. A RNA resultante foi aplicada para monitorar variáveis do reator de pesquisas IEA-R1 do IPEN. Os resultados obtidos mostram que o algoritmo desenvolvido é capaz de melhorar o desempenho do modelo que estima o valor de variáveis do reator. Em testes com diferentes números de neurônios na camada escondida, utilizando como comparativos o erro quadrático médio, o erro absoluto médio e o coeficiente de correlação, o desempenho da RNA otimizada foi igual ou superior ao da tradicional.
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

Radiation damage occurs in reactor pressure vessel (RPV) steel, causing microstructural changes such as point defect clusters, interstitial loops, vacancy-solute clusters, and precipitates, that cause material embrittlement. Radiation damage is a crucial concern in the nuclear industry since many nuclear plants throughout the US are entering the first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. The result of extended operation is that the RPV and other components will be exposed to higher levels of neutron radiation than they were originally designed to withstand. There is currently no nondestructive evaluation technique that can unambiguously assess the amount of radiation damage in RPV steels. Nonlinear ultrasound (NLU) is a nondestructive evaluation technique that is sensitive to microstructural features such as dislocations, precipitates, and their interactions in metallic materials. The physical effect monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features. This effect is quantified with the measurable acoustic nonlinearity parameter, beta. In this work, nonlinear ultrasound is used to characterize radiation damage in reactor pressure vessel steels over a range of fluence levels, irradiation temperatures, and material composition. Experimental results are presented and interpreted with newly developed analytical models that combine different irradiation-induced microstructural contributions to the acoustic nonlinearity parameter.

Cette thèse de doctorat expose l'étude la luminescence induite dans des capteurs à gaz de type chambre à fission, afin d'évaluer son utilisation pour le suivi de puissance des réacteurs nucléaires de quatrième génération comme ASTRID.Le dépôt d'énergie dans un gaz par des ions lourds produits lors de fissions nucléaires dans une chambre à fission est d'ordinaire exploité pour sa capacité à produire des charges libres, qui peuvent être détectées avec des électrodes. Cette méthode, utilisée depuis le début de l'ère atomique, n'est pas sans inconvénients, et des axes d'amélioration ont été clairement identifiés.En estimant la production d'états excités du gaz d'une chambre à fission classique, on déduit que de la lumière peut être émise en son sein, et être détectée par un photo-détecteur. L'utilisation du signal optique plutôt qu'électrique apporte une robustesse accrue aux systèmes de mesures neutroniques, insensibles au bruit électromagnétique.L'étude spectroscopique d'un plasma d'origine nucléaire dans un dispositif analytique dimensionné et conçu par nos soins permet d'enregistrer des spectres de raies d'émission caractéristiques de la scintillation du gaz. L'analyse de ses raies permet d'affirmer la supériorité de la détection optique de neutrons, car la pression et la composition du gaz de remplissage peuvent alors être estimées en ligne.Parallèlement à nos activités expérimentales, un code de simulation de la dynamique de peuplement des niveaux des plasmas froids excités par ions-lourds est développé afin d'optimiser de futurs détecteurs basés sur ce principe. Les temps caractéristiques nécessaires à la génération d'états radiatifs et la position des excitations autour d'une trace d'ionisation ont été estimés.Plusieurs prototype de capteurs de neutrons fonctionnant uniquement sur un principe optique sont exploités dans des faisceaux de neutrons froids du réacteur Orphée afin de vérifier la preuve de concept d'un tel système de mesure neutronique. Des essais dans le réacteur Cabri permettent quant à eux de quantifier la limite de détection, la linéarité et la dynamique du système.Les très bons résultats apportés par ces trois volets valident la preuve de concept de la détection passive et en ligne de neutrons, capable d'auto-diagnostic, pour des applications exigeantes en milieu difficile
This PhD thesis details the study of luminescence produced in gaseous detectors as fission chambers, to evaluate its use in 4th generation nuclear reactor power monitoring, as ASTRID.Energy deposition in the gas, from heavy-ions produced during nuclear fission in a fission chamber is usually used for its charge generation ability, that can be detected with electrodes. Such method has been used since the beginning of atomic era present disadvantages, and necessary improvements have been assessed.By estimating fission chamber gas excited states production, one can deduce intrinsic light emission, and its detection by appropriate photo-detectors. Use of an optical signal against an electric one improves dependability of a neutron flux monitoring system, immune to electromagnetic noise.Spectroscopic studies of a nuclear plasma in a self-designed analytical device allows observation of typical gas scintillation emission lines. Analysis of emission lines reinforce optical detection attraction, as pressure and gas composition can be monitored online.Alongside our experiments, a simulation code dedicated to heavy-ions induced cold-plasma excited levels population has been produced to optimise future detectors relying on optical emission. Typical times encountered for excited states generation and their position around an ionization track have been estimated.Several neutron detector prototypes based on the sole gas scintillation principle have been assessed in cold neutron beam lines of the Orphée reactor, to validate a proof of concept for neutron monitoring. Experiments on the Cabri reactor allowed quantification of detection limit, linearity and dynamic range of our system.The positive outcomes provided by these 3 lines of work validate the proof of concept of passive, on-line, and self-diagnosed neutron detection with high dependability in harsh environments

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O progresso de projetos anteriores salientou a necessidade de tratar o problema dos softwares para detecção, a partir de bases de dados de publicações científicas, de tendências emergentes de pesquisa e desenvolvimento. Evidenciou-se a carência de aplicações computacionais eficientes dedicadas a este propósito, que são artigos de grande utilidade para um melhor planejamento de programas de pesquisa e desenvolvimento em instituições. Foi realizada, então, uma revisão dos softwares atualmente disponíveis, para poder-se delinear claramente a oportunidade de desenvolver novas ferramentas. Como resultado, implementou-se um aplicativo chamado Citesnake, projetado especialmente para auxiliar a detecção e o estudo de tendências emergentes a partir da análise de redes de vários tipos, extraídas das bases de dados científicas. Através desta ferramenta computacional robusta e eficaz, foram conduzidas análises de frentes emergentes de pesquisa e desenvolvimento na área de Sistemas Geradores de Energia Nuclear de Geração IV, de forma que se pudesse evidenciar, dentre os tipos de reatores selecionados como os mais promissores pelo GIF - Generation IV International Forum, aqueles que mais se desenvolveram nos últimos dez anos e que se apresentam, atualmente, como os mais capazes de cumprir as promessas realizadas sobre os seus conceitos inovadores.
Dissertação (Mestrado em Tecnologia Nuclear)
IPEN/D
Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Thesis (Ph. D.)--University of Missouri-Columbia, 2006.
The entire dissertation/thesis text is included in the research.pdf file; the official abstract appears in the short.pdf file (which also appears in the research.pdf); a non-technical general description, or public abstract, appears in the public.pdf file. Title from title screen of research.pdf file (viewed on August 13, 2007) Vita. Includes bibliographical references.

In the following work, the MAVRIC sequence of the Scale6.1 code package was tested for its efficacy in calculating a wide range of shielding parameters with respect to HTGRs. One of the NGNP designs that has gained large support internationally is the VHTR. The development of the Scale6.1 code package at ORNL has been primarily directed towards supporting the current United States' reactor fleet of LWR technology. Since plans have been made to build a prototype VHTR, it is important to verify that the MAVRIC sequence can adequately meet the simulation needs of a different reactor technology. This was accomplished by creating a detailed model of the VHTR power plant; identifying important, relevant radiation indicators; and implementing methods using MAVRIC to simulate those indicators in the VHTR model. The graphite moderator used in the design shapes a different flux spectrum than water-moderated reactors. The different flux spectrum could lead to new considerations when quantifying shielding characteristics and possibly a different gamma-ray spectrum escaping the core and surrounding components. One key portion of this study was obtaining personnel dose rates in accessible areas within the power plant from both neutron and gamma sources. Additionally, building from professional and regulatory standards a surveillance capsule monitoring program was designed to mimic those used in the nuclear industry. The high temperatures were designed to supply heat for industrial purposes and not just for power production. Since tritium, a heavier radioactive isotope of hydrogen, is produced in the reactor it is important to know the distribution of tritium production and the subsequent diffusion from the core to secondary systems to prevent contamination outside of the nuclear island. Accurately modeling indicators using MAVRIC is the main goal. However, it is almost equally as important for simulations to be carried out in a timely manner. MAVRIC uses the discrete ordinates method to solve the fixed-source transport equation for both neutron and gamma rays on a crude geometric representation of the detailed model. This deterministic forward solution is used to solve an adjoint equation with the adjoint source specified by the user. The adjoint solution is then used to create an importance map that can weight particles in a stochastic Monte Carlo simulation. The goal of using this hybrid methodology is to provide complete accuracy with high precision while decreasing overall simulation times by orders of magnitude. The MAVRIC sequence provides a platform to quickly alter inputs so that vastly different shielding studies can be simulated using one model with minimal effort by the user. Each separate shielding study required unique strategies while looking at different regions in the VHTR plant. MAVRIC proved to be effective for each case.

碩士
國立清華大學
工程與系統科學系
99
Power monitoring system is one of the most important parts of nuclear planet. Due to the fact that the neutron flux in the reactor core is proportional to the reactor power, the wide range neutronmonitor system (WRNM) plays an important role in controlling and monitoring the situation of the reactor core. It can provide two main functions. First, it offers the information about the power variation; second, it provides the trip signals for the reactor protecting system in order to adjust the control rod. The classic WRNM systems were built by analog circuit system for the most parts. It is difficult to tune the parameters among different reactor system. Comparing to the analog cir-cuit system, the digital circuit system is flexible to redesign the parameters, providing a wide adaptability to different reactor systems; Moreover, analog circuit system is sensitive to the noise more than digital circuit system does, hence the readout could have errors for analog circuit system. Thanks to the rapid growing of IC technology, the efficiency of ADCs, the memory spaces of system, and the capability of realizing DSP algorithms becomes better and better these years. Using digital system to achieve the real-time processing as what can be done by classic analog system is possible for today’s WRNM. Therefore, the goal of our research is to evaluate the feasibility of digitizing the classic WRNM system, especially for the Campbell technique realization.

International Atomic Energy Agency (IAEA) safeguards are one method of proliferation resistance which is applied at most nuclear facilities worldwide. IAEA safeguards act to prevent the diversion of nuclear materials from a facility through the deterrence of detection. However, even with IAEA safeguards present at a facility, the country where the facility is located may still attempt to proliferate nuclear material by exploiting weaknesses in the safeguards system. The IAEA's mission is to detect the diversion of nuclear materials as soon as possible and ideally before it can be weaponized. Modern IAEA safeguards utilize unattended monitoring systems (UMS) to perform nuclear material accountancy and maintain the continuity of knowledge with regards to the position of nuclear material at a facility. This research focuses on evaluating the reliability of unattended monitoring systems and integrating the probabilistic failure of these systems into the comprehensive probabilistic proliferation resistance model of a facility. To accomplish this, this research applies reliability engineering analysis methods to probabilistic proliferation resistance modeling. This approach is demonstrated through the analysis of a safeguards design for the Example Sodium Fast Reactor Fuel Cycle Facility (ESFR FCF). The ESFR FCF UMS were analyzed to demonstrate the analysis and design processes that an analyst or designer would go through when evaluating/designing the proliferation resistance component of a safeguards system. When comparing the mean time to failure (MTTF) for the system without redundancies versus one with redundancies, it is apparent that redundancies are necessary to achieve a design without routine failures. A reliability engineering approach to probabilistic safeguards system analysis and design can be used to reach meaningful conclusions regarding the proliferation resistance of a UMS. The methods developed in this research provide analysts and designers alike a process to follow to evaluate the reliability of a UMS.

In the pharmaceutical industry, both in research and development and industrial process, continuous reaction monitoring became early on an important part of drug development and control. The generally used chromatography techniques have some disadvantages such as being destructive, time-consuming and expensive. Nuclear Magnetic Resonance spectroscopy has been used as an alternative to these techniques however the acquisition and maintenance cost of these instruments have been an obstacle towards a generalized use in an industrial setting. With the breakthrough in the development of benchtop NMR instruments, this technique has become a competitive alternative to traditional analysis methods, offering a rapid, non-destructive, more economic, and environmentally friendly analysis, applicable to both simple and more complex systems. In this thesis, we aimed to assess the viability of using benchtop NMR to study chemical reactions, in continuous flow conditions, in a research and development pharmaceutical laboratory. We performed an in-depth study on the sensitivity and capacity of the instrument, like reproducibility and validation parameters (limit of detection (LOD) and limit of quantification (LOQ)) to evaluate its response to challenging problems in the production of drugs in the pharmaceutics industry, and in a research laboratory environment. Later, we used reaction models as proof of concept for its application in online reaction monitoring. These studies led us to conclude that benchtop NMR can be an asset in the study of reactions in flow mode, obtaining real-time quantitative information, allowing the concentration of the rea-gent/product to be determined throughout the reaction with rapid analysis time. This led us to make a general guideline that allows a user to decide if benchtop NMR can be used for a specific reaction and if so, what are the conditions that must be addressed during the monitoring experimental conditions. Benchtop NMR spectroscopy offers great possibilities to become, in the future, a common, affordable tool for reaction monitoring both in industry and even in an academic environment.
Na indústria farmacêutica, a monitorização de reações tornou-se uma parte importante no desenvolvimento e controlo de fármacos, quer em ambiente laboratorial de investigação e desenvolvimento (&D) quer a nível industrial. Este controlo é, usualmente, feito utilizando técnicas cromatográficas, contudo estas têm a desvantagens de serem técnicas destrutivas, demoradas e de custo monetário elevado. A espectroscopia de ressonância magnética nuclear (RMN) tem vindo a ser considerada uma possível alternativa a estas técnicas, contudo o custo de aquisição e de manutenção dos equipamentos tem sido o maior obstáculo para a sua utilização generalizada num ambiente industrial. Com o avanço no desenvolvimento de aparelhos de RMN de bancada, esta técnica tornou-se uma alternativa competitiva aos métodos analíticos tradicionais, oferecendo uma análise rápida, não destrutiva, mais económica e sustentável, aplicável tanto a sistemas simples como a sistemas moleculares mais complexos. O objetivo desta tese foi avaliar a viabilidade da utilização de aparelhos de RMN de bancada para monitorizar reações químicas em fluxo contínuo, num laboratório farmacêutico de I&D. Assim, para avaliar a resposta deste tipo de equipamentos para os problemas presentes na produção de fármacos e num ambiente laboratorial, foi feito um estudo da sensibilidade e da capacidade do instrumento, analisando parâmetros de reprodutibilidade e de validação (limite de deteção (LOD) e de quantificação (LOQ)). Posteriormente, usámos modelos de reações como prova de conceito para verificar a sua aplicação na monitorização online de reações. Estes estudos permitiram-nos concluir que os aparelhos de RMN de bancada podem ser uma mais-valia no estudo de reações em fluxo, obtendo informação quantitativa em tempo real, permitindo determinar a concentração de reagentes/produtos ao longo do tempo da reação de um modo rápido e com mínimo de manipulação. Os resultados permitiram construir um protocolo geral para o utilizador decidir se perante uma reação específica se este tipo de equipamento pode ser usado uma reação especifica e, em caso afirmativo, quais são as condições experimentais que devem ser abordadas durante a monitorização. A técnica de espectroscopia de RMN de bancada apresenta um enorme potencial de se tornar, no futuro, uma ferramenta comum e acessível para a monitorização de reações, tanto em ambiente industrial como em ambiente académico.