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Made available in DSpace on 2014-10-09T12:53:25Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:09:32Z (GMT). No. of bitstreams: 1 12213.pdf: 5005382 bytes, checksum: c4bde457760b3a6d6f53b64c21e33010 (MD5) / No presente trabalho, foi utilizada a técnica de aquecimento por microondas para estudar a imobilização de rejeitos radioativos de nível de atividade baixo e médio, como resinas de troca iônica exauridas, empregadas na remoção de íons indesejáveis dos circuitos primários de refrigeração de reatores nucleares refrigerados a água, e aquelas usadas em colunas de separação química e radionuclídica no controle de qualidade de radioisótopos. Matrizes betuminosas reforçadas com alguns tipos de borrachas (Neoprene®, Silicone e Etileno Vinil Acetato - EVA), provenientes de material descartado ou sobras de produção, foram utilizadas para incorporação dos rejeitos radioativos. As irradiações das amostras foram feitas em um forno de microondas caseiro, que opera com freqüência de 2.450MHz e possui potência de 1.000W. As amostras foram caracterizadas, empregando-se ensaios de penetração, resistência à lixiviação, pontos de amolecimento, fulgor e combustão, termogravimetria e microscopia óptica. Os resultados obtidos mostraram-se compatíveis com os padrões dos componentes das matrizes, indicando que esta técnica é uma alternativa bastante útil aos métodos de imobilização convencionais e para esses tipos de rejeitos radioativos. / Dissertação (Mestrado) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

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Experimental data on SuperCritical-Water (SCW) cooled bundles are very limited. Major problems with performing such experiments are: 1) small number of operating SCW experimental setups and 2) difficulties in testing and experimental costs at very high pressures, temperatures and heat fluxes. However, SuperCritical Water-cooled nuclear Reactor (SCWRs) designs cannot be finalized without such data. Therefore, as a preliminary approach experiments in SCW-cooled bare tubes and in bundles cooled with SC modeling fluids can be used. One of the SC modeling fluids typically used is Freon-12 (R-12) where the critical pressure is 4.136 MPa and the critical temperature is 111.97ºC. These conditions correspond to a critical pressure of 22.064 MPa and critical temperature of 373.95ºC in water. A set of experimental data obtained in a Freon-12 cooled vertical bare bundle at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia) was analyzed. This set consisted of 20 cases of a vertically oriented 7-element bundle installed in a hexagonal flow channel. To secure the bundle in the flow channel 3 thin spacers were used. The dataset was obtained at equivalent parameters of the proposed SCWR concepts. Data was collected at pressures of about 4.65 MPa for several different combinations of wall and bulk-fluid temperatures that were below, at, or above the pseudocritical temperature. Heat fluxes ranged from 9 kW/m2 to 120 kW/m2 and mass fluxes ranged from 440 kg/m2s to 1320 kg/m2s. Also inlet temperatures ranged from 70ºC – 120ºC. The test section consisted of fuel elements that were 9.5 mm in diameter with the total heated length of 1 m. Bulk-fluid and wall temperature profiles were recorded using a combination of 8 different thermocouples.The data was analyzed with respect to its temperature profile and heat transfer coefficient along the heated length of the test section. In a previous study it was confirmed that there is the existence of three distinct regimes for forced convention with supercritical fluids. (1) Normal heat transfer; (2) Deteriorated heat transfer, characterized by higher than expected temperatures; and (3) Improved heat transfer, characterized by lower than expected temperatures. All three regions were observed for the 7 rod bundle experiments. This work compares the experimental data to predictions based upon current 1-D correlations for heat transfer in supercritical fluids. Results show that no current 1-D correlation was able to accurately predict heat transfer coefficients within ±50%. A parametric analysis of the data was also completed to determine if continuity in the experiment was present. Results of this study show that two distinct regions are present in the data. For cases with a mass flux below 1200 kg/m2s wall temperature profiles appear to be normal while in cases with mass flux above 1200 kg/m2s temperature given by the wall thermocouples were higher than normal. This phenomenon occurred regardless of heat flux-to-mass flux ratios. / UOIT

Pressurizer modeling for predicting the dynamic pressure of the PIUS system is presented. The transient behavior of this model for the PIUS system was investigated. The validity of this model for the PIUS system is limited to transients that are neither too large nor too long in duration. For example, the model is not capable of describing events following a complete loss of liquid for the pressurizer. However, the model can be used for qualitative prediction of the PIUS system behavior for a wide variety of severe transients. A review of pressurizer modeling indicates that the neglecting of the change in the internal energy of the subcooled water during transients is an acceptable assumption. The inherently safe feature of the PIUS system was confirmed through the self-shutdown of the reactor or, in some cases, through reactor power reduction as a result of the ingress of the pool boric acid solution into the primary system. This dynamic model was constructed of three major components: (1) The primary loop, (2) The secondary loop, and (3) The natural convection loop through the pool. A lumped parameter model, uniform heat transfer, and point kinetics have been the main approximations in this model. Other approximations are mentioned during the modeling of each component of the model. The dynamic model was simulated using the DARE-P continuous system simulation language which was developed in the Electrical Engineering Department at the University of Arizona.

SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. SuperCritical Water (SCW) Nuclear Power Plants (NPPs) are expected to have much higher operating parameters compared to current NPPs, i.e., pressure of about 25 MPa and outlet temperature up to 625 oC. This study presents the heat transfer analysis of an intermediate Heat exchanger (HX) design for indirect-cycle concepts of Pressure-Tube (PT) and Pressure-Vessel (PV) SCWRs. Thermodynamic configurations with an intermediate HX gives a possibility to have a single-reheat option for PT and PV SCWRs without introducing steam-reheat channels into a reactor. Similar to the current CANDU and Pressurized Water Reactor (PWR) NPPs, steam generators separate the primary loop from the secondary loop. In this way, the primary loop can be completely enclosed in a reactor containment building. This study analyzes the heat transfer from a SCW primary (reactor) loop to a SCW and Super-Heated Steam (SHS) secondary (turbine) loop using a double-pipe intermediate HX. The numerical model is developed with MATLAB and NIST REFPROP software. Water from the primary loop flows through the inner pipe, and water from the secondary loop flows through the annulus in the counter direction of the double-pipe HX. The analysis on the double-pipe HX shows temperature and profiles of thermophysical properties along the heated length of the HX. It was found that the pseudocritical region has a significant effect on the temperature profiles and heat-transfer area of the HX. An analysis shows the effect of variation in pressure, temperature, mass flow rate, and pipe size on the pseudocritical region and the heat-transfer area of the HX. The results from the numerical model can be used to optimize the heat-transfer area of the HX. The higher pressure difference on the hot side and higher temperature difference between the hot and cold sides reduces the pseudocritical-region length, thus decreases the heat-transfer surface area of the HX. / UOIT

A large set of experimental data, obtained in Russia, was analyzed and a new heat-transfer correlation for supercritical water was developed. This experimental dataset was obtained within conditions similar to those for proposed SuperCritical Water-cooled nuclear Reactor (SCWR) concepts. Thus, this new correlation, for forced convective heat transfer in the normal heat-transfer regime, can be used for preliminary heat-transfer calculations in SCWR fuel channels. It has demonstrated a good fit for Heat Transfer Coefficient (HTC) values (±25%) and for wall temperature calculations (±15) for the analyzed dataset. This correlation can be used for supercritical water heat exchangers linked to indirectcycle concepts and the co-generation of hydrogen, for future comparisons with other independent datasets, with bundle data, as the reference case, for the verification of computer codes for SCWR core thermalhydraulics and for the verification of scaling parameters between water and modeling fluids. / UOIT