Canada has participated in the Generation IV International Forum (GIF) collaboration in the area of Super Critical Water-cooled Reactors (SCWR). Similar to the current CANDU technologies, in the Canadian SCWR design the low pressure heavy water moderator system is separated from the supercritical coolant system (25MPa). The High Efficiency Re-entrant Channel (HERC) design in the Canadian SCWR has multiple coolant regions (i.e. coolant in the downward center flow tube and coolant in the upward outer fuel region) and provides thermal isolation between the moderator and heat transport system fluid. Although the overall reactivity feedback in the channel is negative for equilibrium density decrease transients, a temporary positive reactivity may be induced during non-equilibrium conditions such as cold-leg Loss of Coolant Accidents (LOCA). The primary objective of this study is to investigate the fuel and coolant behaviors under postulated LOCA transients, in particular those caused by cold-leg breaks, and to demonstrate the effectiveness of several proposed safety systems in the Canadian SCWR. The one-dimensional thermalhydraulic system code RELAP5 has been used for the safety analysis in many LWRs. The latest version RELAP5/MOD4 has been improved to accommodate supercritical water and is used in this study. A RELAP5 model is constructed based on the most recent Canadian SCWR design. The 336 fuel channels are split into two representative groups each with a series of hydraulic and heat structure models. A benchmark study is conducted by comparing RELAP5 to CATHENA simulations and shows good agreement for both steady-state and transient predictions. The RELAP5 model is then used to predict the system response to several postulated LOCA transients. For a 100% (single-ended) cold-leg break located in between the feedwater pump and the inlet plenum, the system pressure immediately drops followed closely by a flow reversal with rapid discharge from the break. A brief power pulse (178%FP) is observed under this non-equilibrium depressurization scenario. The transient simulations show the potential for two sheath temperature maxima, one early in the transient as a result of the power pulse and the subsequent flow-power mismatch, and another later peak resulting from the fuel heat-up under near stagnant channel flow conditions (such as in the failure of the Emergency Core Cooling Systems) as the heat transfer regime changes to radiation dominated. The Automatic Depressurization System (ADS) located on the hot-leg side mitigates the later fuel heat-up by introducing forward channel flows. This effect is enhanced by additional coolant supplied from Low Pressure Coolant Injection (LPCI) which is part of the Emergency Core Cooling System (ECCS). Under the 100% break LOCA/LOECC transient, the core inventory is depleted rapidly after the break and thermal radiation becomes the dominant heat removal mechanism. The highest MCST, 1331 K, is achieved approximately 136s after the break and meets the safety criterion (1533 K). Beyond this time the sheath temperatures gradually decrease either by the continuous LPCI from the reactor sumps, gravity driven core cooling, or in the event of a failure of those systems by the Passive Moderator Cooling System (PMCS). LOCAs initiated by break sizes varying from 5% to 100% of the cold-leg cross-section area are simulated under loss of ECCS. In this specific design, break sizes less than 15% are defined as SBLOCAs and show an early pressure increase up to the Safety Relief Valve (SRV) setpoint. During SBLOCAs, the first MCST peak is more limiting than the large LOCA case because of insufficient fuel cooling caused by relatively low reverse flow. However, these lower reverse flows prolong the period of blowdown cooling and hence help to mitigate the secondary MCST peak. The worst LOCA case occurs in the 15% break case with a maximum cladding temperature of 1450 K. The results showed the most sensitive parameters are delays associated with SDS action, emissivity and ADS actuation parameters. / Thesis / Master of Applied Science (MASc)

<p>Vid Positionsbestämning av radiosändare med TDOA(Time Difference Of Arrival) är det traditionella sättet att låta varje uppmätt tidsskillnad definiera en hyperbelgren på vilken sändaren befinner sig. Skärningspunkten mellan två eller fler hyperbelgrenar ger sändarens position.</p><p>Målet med examensarbetet är att identifiera möjliga sändarpositioner och detektera mångtydigheter. Problemet har lösts med kägelsnittsmetoden som låter tre mottagares positioner och deras uppmätta tidsskillnader definiera ett kägelsnitt där sändarens position är i en av dess fokuspunkter.</p>

Vid Positionsbestämning av radiosändare med TDOA(Time Difference Of Arrival) är det traditionella sättet att låta varje uppmätt tidsskillnad definiera en hyperbelgren på vilken sändaren befinner sig. Skärningspunkten mellan två eller fler hyperbelgrenar ger sändarens position. Målet med examensarbetet är att identifiera möjliga sändarpositioner och detektera mångtydigheter. Problemet har lösts med kägelsnittsmetoden som låter tre mottagares positioner och deras uppmätta tidsskillnader definiera ett kägelsnitt där sändarens position är i en av dess fokuspunkter.

Submitted by Almir Azevedo (barbio1313@gmail.com) on 2015-01-06T14:40:18Z No. of bitstreams: 1 dissertação mestrado ien 2014 Alexandre Soares.pdf: 4924692 bytes, checksum: 9690f1916310052f66cbc6e41f71c443 (MD5) / Made available in DSpace on 2015-01-06T14:40:18Z (GMT). No. of bitstreams: 1 dissertação mestrado ien 2014 Alexandre Soares.pdf: 4924692 bytes, checksum: 9690f1916310052f66cbc6e41f71c443 (MD5) Previous issue date: 2014-11 / O presente trabalho propões um estudo de um acidente com perda de refrigerante de um reator PWR através de um Simulador de Escopo Completo e do código computacional RELAP. Para tal, foi considerado um acidente com perda de refrigerante com área de quebra de 160 cm2 na perna fria do circuito 20 do sistema de refrigeração do reator da planta da Usina Nuclear de Angra 2, com o reator operando em condições estacionária, a 100% de potência. Foi admitido ainda, que ocorreu simultaneamente a perda de Suprimento Externo de Energia Elétrica e que a disponibilidade do Sistema de Refrigeração de Emergência do Núcleo não era plena. Os resultados obtidos apresentam-se bastante relevantes e com possibilidade de serem usados no planejamento de atividades futuras, visto que a construção de Angra 3 se apresenta em andamento e se assemelha a Angra 2. / The present paper porposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2.

Due to the Three Mile Island accident in 1979 the Swedish government took the decision in 1986 to impose a pressure relief system for Swedish BWR’s which prevents containment overpressure in case of LOCA. This pressure relief system consists of a rupture disks in two different systems, non-filtered system 361 and filtered system 362. During a steam line break it is not clear if an unjustified activation of rupture disk 361 or 362 could possibly occur. If significant amount of nitrogen will leak out from the containment then, there is a risk of low pressure in the containment (e.g. due to activation of containment spray) with leaking rupture disks, which might cause air inflow to the containment and burning of hydrogen, so conditions of activation of rupture disk must be studied. The main objective of this master thesis is the investigation of conditions of activation of rupture disk in BWR containment filtering system. In order to find out these conditions specific software called GOTHIC has been used. The methodology of this master thesis has been modeling different containments with GOTHIC software; this thesis work will go from a simple GOTHIC model, that consist in nine lumped control volumes connected by flow paths, until a more complex GOTHIC model that consist in a combination of lumped and 3D control volumes, connected among them by flow paths and 3D connectors. A large LOCA in the upper part of the reactor vessel will be considerate, due to this severe accident; conditions for the activation of the rupture disk will be complying. It has to be mentioned that pressure in the lumped modeling will be lower than pressure in the 3D volumes. Activation time for the lumped modeling will be 8,5 seconds after the steam break for system 362 and activation time for 3D modeling will be 2,8 seconds for system 362 as well. In neither case 361 system will be activated. Considering this is a nuclear safety study and accuracy must be a key point, for further investigations it might be more than advisable using 3D control volumes instead of lumped control volumes. It has to be mentioned also that due to there is no experimental data, uncertainty regarding to the results exist, and if a further safety analysis want to be done, sensitive study of the parameters implemented on GOTHIC software should be performed in the future.

In nuclear power plants, a pipe break in the cooling line releases a jet that damages other equipment in containment, and is known as a loss of coolant accident (LOCA). This report specifically focuses on boiling water reactor (BWR) applications as a guide for future studies with pressurized water reactors (PWRs). This report presents a methodology for characterizing the jet such that, given a set of upstream conditions, the pressure field and damage potential of the jet can be predicted by an end user with a minimum of computation. The resultant model has many advantages over previous models in that it is easily calculated with knowledge readily available to plant operators and it provides new metrics that allow for a quick and intuitive understanding of the damage potential of the jet. / text

This research presents the methodologies used to resolve the Nuclear Regulatory Commission Generic Safety Issue 191. The presented results are specific to South Texas Project Nuclear Operating Company (STPNOC). However, the proposed methodologies may be applicable to other nuclear power plants given the appropriate plant-specific frequencies. This research provides important inputs to CASA Grande, a computer program used to model physical phenomena and quantify uncertainties to obtain estimates of failure probabilities for post-loss-of-coolant accident events at the STPNOC containment. We provide modeling and sampling methods for loss-of-coolant accident (LOCA) frequencies and break sizes. We focus on a study known as NUREG-1829 (Tregoning et al., 2008), which includes an expert elicitation of quantiles governing the (annual) frequency of a LOCA in boiling water reactors and pressurized water reactors. We propose to model LOCA frequencies with bounded Johnson distributions and to sample break sizes using uniform distributions. We then develop a new method to distribute LOCA frequencies to different locations within a plant to account for the location-dependent differences while preserving the NUREG-1829 frequencies. We also propose to linearly interpolate the NUREG-1829 LOCA frequencies to obtain the frequencies for any break sizes other than those from NUREG-1829. In addition, we present a method to obtain the distribution of LOCA frequency within a break-size interval providing important inputs to the probabilistic risk assessment quantification for STPNOC. We review methods of combining the probability distributions of multiple experts to obtain a single probability distribution. More specifically, we describe the relative merits of the arithmetic mean (AM) and geometric mean (GM) as ways of performing this aggregation in the context of probabilities associated with rare events. Examining a set of pressurized water reactor results from NUREG-1829, we conclude that the GM represents a consistently sensible notion of the middle of the opinions expressed by nine experts. We further conclude that the AM is inappropriate for representing the center of the group's opinion for large effective break sizes. Instead, as the break size grows large a single expert's opinion dominates the combination produced by the AM. / text

Submitted by Jos? Henrique Henrique (jose.neves@ufvjm.edu.br) on 2016-12-16T13:26:38Z No. of bitstreams: 2 license_rdf: 0 bytes, checksum: d41d8cd98f00b204e9800998ecf8427e (MD5) ludmila_pires_miranda.pdf: 952225 bytes, checksum: 480197cc21b0be88145b5cb167181c71 (MD5) / Approved for entry into archive by Rodrigo Martins Cruz (rodrigo.cruz@ufvjm.edu.br) on 2016-12-19T16:46:22Z (GMT) No. of bitstreams: 2 license_rdf: 0 bytes, checksum: d41d8cd98f00b204e9800998ecf8427e (MD5) ludmila_pires_miranda.pdf: 952225 bytes, checksum: 480197cc21b0be88145b5cb167181c71 (MD5) / Made available in DSpace on 2016-12-19T16:46:22Z (GMT). No. of bitstreams: 2 license_rdf: 0 bytes, checksum: d41d8cd98f00b204e9800998ecf8427e (MD5) ludmila_pires_miranda.pdf: 952225 bytes, checksum: 480197cc21b0be88145b5cb167181c71 (MD5) Previous issue date: 2016 / Funda??o de Amparo ? Pesquisa do Estado de Minas Gerais (FAPEMIG) / O estudo analisou a influ?ncia de alocar parcelas em diferentes pontos do talh?o com a finalidade de um invent?rio pr?-corte, utilizando dois arranjos de intensidade amostral e tamanho de parcela, diferentes m?todos de amostragem (Casual Simples ? ACS ou Sistem?tica ? AS), forma de parcela (retangular ou circular) e localiza??o da parcela no talh?o (borda, borda e centro e apenas no centro). Foi realizado um censo no talh?o, onde foram mensuradas todas as circunfer?ncias ? 1,30 metros de altura. A estimativa da altura foi obtida por meio de modelo hipsom?trico. O ajuste do modelo foi feito com duas bases de dados para detectar a influ?ncia do efeito de borda na altura das ?rvores: a primeira composta por dados provenientes da ?ltima medi??o do invent?rio florestal cont?nuo (IFC) realizado no talh?o; a segunda proveniente de cinquenta ?rvores localizadas na borda do talh?o. Com as equa??es ajustadas, utilizou-se o teste de identidade de modelo para verificar o efeito de borda na altura. Para definir o efeito de borda sobre o di?metro ajustou-se a fun??o Weibull de duas formas: 1) para cada uma das 10 primeiras linhas de borda do talh?o; e 2) fez-se o ajuste das linhas de forma acumulativa, come?ando pela linha 1 at? a d?cima linha. Conhecendo-se as estimativas dos par?metros, as distribui??es diam?tricas foram estimadas, tomando como base a frequ?ncia observada da primeira linha de borda. Para verificar se a distribui??o diam?trica estimada variava conforme adentrava no talh?o utilizou-se o teste de Kolmogorov-Smirnov, a 95% de probabilidade. Para o ajuste dos modelos volum?tricos foram usados dados de cubagem de ?rvores-amostra localizadas no centro do talh?o. Combinando a localiza??o da parcela, m?todo de amostragem, formato da parcela e arranjo de intensidade amostral e tamanho da parcela, foram simulados 28 cen?rios de amostragem. A compara??o entre os procedimentos foi utilizando a precis?o e exatid?o. O resultado do teste de identidade de modelo indicou que h? influ?ncia da borda sobre a altura das ?rvores. Os testes de Kolmogorov-Smirnov demonstraram n?o haver diferen?a diam?trica entre as 10 primeiras linhas de borda do talh?o. O modelo volum?trico de Schumacher e Hall foi o que resultou nas melhores estimativas. Nos procedimentos de amostragem simulados, a ACS foi mais precisa e eficiente para um arranjo com intensidade amostral maior e parcelas de tamanho menores, j? a AS foi melhor com intensidade amostral menor e parcelas maiores; quanto ao formato, tanto circular, quanto retangular obtiveram bons resultados. Em rela??o ? localiza??o das parcelas no talh?o, os resultados foram mais exatos e precisos quando as parcelas foram alocadas no centro do talh?o, seguidos das parcelas alocadas na borda e centro e por fim aquelas alocadas na borda do talh?o. / Disserta??o (Mestrado) ? Programa de P?s-Gradua??o em Ci?ncia Florestal, Universidade Federal dos Vales do Jequitinhonha e Mucuri, 2016. / This study examined the influence of allocating plots at different points of the stand, using two arrays of sampling intensity and plot size, different methods of sampling (Simple Casual ? ACS or Systematic Sampling ? AS), plot format (rectangular or circular) and plot location in the compartment (border, border and center, and only in the center). We performed a census in the stand, where all circumferences at 1.3 meters above the ground were measured. Height estimation was through the Hypsometric model. The model fitting was conducted with two databases to detect the influence of the border effect at different tree heights: the first model fitting consisted of data from the last measurement of IFC (Continuous Forest Inventory ? IFC) conducted in the stand; the second one consisted of fifty trees located on the border of the stand. For the adjusted model, the researcher used a model identity test to determine the effect of the border on the height. To define the effect of the border on the diameter, the Weibull function was fitted in two ways: 1) for each one of the first 10 borders in the stand; and 2) for the adjustment of the lines cumulatively, beginning from the first line to the tenth line. Knowing the parameters, we estimated diameter distributions, based on the observed frequency of the first border. To verify if the estimated diameter distribution varied according to the plot, we used the Kolmogorov-Smirnov test at 95% probability. For the fitting of the volumetric models, we used cubing data of trees located in the center of the stand. So, combining the location of the plot, the sampling method, the inventory type and the plot format, we simulated 28 sampling scenarios. Comparison between procedures was through precision and accuracy. The result of the model identity test indicated that the border influences tree heights. The Kolmogorov- Smirnov tests showed no diametric difference between the first 10 border lines of the stand. The Schumacher and Hall volumetric model was the one that got the best estimates. In the simulated sampling procedures, ACS was more precise and efficient or an arrangement with greater sampling intensity and smaller size plots, while AS was better with lower sampling intensity and larger plots; as the format, either circular, or rectangular achieved good results. Regarding the location of the plots in the stand, the results were more accurate and precise when the plots were allocated in the center of the stand, followed by plots allocated on the border and center and finally those allocated in the stand border.

This diploma thesis deals with measuring the tightness of the hermetic area of nuclear power plant in the way of improving safety. Describes the layout and function of this area - the primary circuit of NPP with WWER 440/213 reactors, as well as methodology for leakage search and leakage calculation procedures. Personal contribution to the issue is a proposal improving tightness of the current state in hermetic area of NPP Jaslovské Bohunice – transition of jacketed pipe designed to collect water from the floor in the box of the steam generator to the heat exchanger of the shower system. This system belongs to the safety system designed to reduce pressure in the LOCA type of accident.

This diploma thesis deals with a pressure suppression system containment during the loss of coolant accidents. It is focused on the containment systems of the nuclear power plants with VVER-440/V-213 reactors. There is described the process of loss of coolant accident. There was designed input model which consists of the zones representing the areas which are connected with junctions and heat structures. Were then selected input parameters for the model calculations. And finally, there have been several calculations for the selected parameters.