Der Reaktordynamikcode DYN3D wird für Kernanalysen von Natrium-gekühlten schnellen Reaktoren (SFR) erweitert. In diesem Bericht werden neu implementierte thermomechanische Modelle für die adäquate Simulation von SFR-Transienten beschrieben, die die Simulation der axialen Wärmeausdehnung von Brennstäben und die radiale Ausdehnung des Reaktorkerns umfassen. Darüber hinaus wurde das Verfahren zur Erstellung von Querschnittsbibliotheken für DYN3D für SFR-Analysen erweitert. Die Verifizierung der neuen Modelle und der Querschnittserstellung erfolgte auf Vollkern-Ebene mit stationären Experimenten von der BFS-Testanlage des IPPE Obninsk und Daten des großen oxidischen Kerns des OECD/NEA-Benchmark und den Experimenten zum Zyklusende des Phenix-Kerns. Die DYN3D-Ergebnisse wurden mit der Monte-Carlo-Referenzlösung verglichen, die durch den SERPENT-Code berechnet wurde. Die Testergebnisse zeigen, dass die neu entwickelten Modelle die Wärmeausdeh-nungseffekte der Kernstruktur genau berücksichtigen können. Das neu entwickelte Verfahren zur Erstellung von Querschnittsbibliotheken wurde ebenfalls auf der Basis von SERPENT-Ergebnissen erfolgreich verifiziert. Zur Validierung wurden mehrere Tests, die sowohl stationäre als auch transiente Fälle aus den Phenix-Experimenten enthalten, mit DYN3D berechnet. Die DYN3D-Lösungen weisen eine gute Übereinstimmung mit den experimentellen Daten auf, was die Anwendbarkeit der Codes für Kernanalysen von Natrium-gekühlten schnellen Reaktoren bestätigt.

Using an inverter-driven compressor in variable capacity heat pump systems has a main drawback, which is the extra loss in the inverter. The present experimental study aims to recover the inverter losses by using brine-cooled and water-cooled inverters, thereby improving the total efficiency of the heat pump system. In order to achieve this goal, a test rig with the air-cooled, water-cooled and brine-cooled inverters is designed and built, and a comparative analysis of the recovered heat, inverter losses and system performance is conducted when the compressor is driven by the water-cooled, brine-cooled and air-cooled inverters at three different switching frequencies for each inverter. The experimental results show that the inverter losses as a magnitude and as a ratio of the total consumed power are lowest in the brine-cooled inverter and highest in the air-cooled one at all the compressor speeds and all the inverter switching frequencies. Moreover, the recovered energy varies between 45 and 125 (W) in the water-cooled inverter, which corresponds to 63 and 69 (%) of the inverter losses; while it varies between 61 and 139 (W) in the brine-cooled inverter, which corresponds to 79 and 90 (%) of the inverter losses. It is also proved that the improvement of the system coefficient of performance (COPsys) is almost the same when the water-cooled or the brine-cooled inverter is used and varies between 0.54 and 3 (%) in comparison with using the air-cooled one. Indeed, the total isentropic efficiency of the compressor is improved slightly when using the water-cooled inverter and little more when using the brine-cooled one at the same running conditions. In addition, the total isentropic efficiency of the compressor is improved by increasing the inverter switching frequency when any of the inverters is used. The experimental results also show that cooling the inverter by the water, which comes out from the condenser, increases the maximum temperature of the base plate of the inverter about 10 °C which could cause a two-fold deterioration in the inverter median life in comparison with cooling the inverter by air. On the contrary, using the brine for cooling the inverter decreases the maximum temperature of the base plate of the inverter about 30 °C which could cause about a six-fold improvement in the inverter median life. / Capacity-controlled Ground Source Heat Pump single-family dwellings

Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety analysis as a tool that enables safety related physical phenomena occurring in the reactor coolant system to be described in more detail and accuracy. Validation is a necessary step in improving predictive capability of a computationa code or coupled computational codes. Validation refers to the assessment of model accuracy incorporating any uncertainties (aleatory and epistemic) that may be of importance. The uncertainties must be identi ed, quanti ed and if possible, reduced. In the rst part of this thesis, a discussion on the development of an approach and experimental facility for the validation of coupled Computational Fluid Dynamics codes and System Thermal Hydraulics (STH) codes is given. The validation of a coupled code requires experiments which feature signi cant two-way feedbacks between the component (CFD sub-domain) and the system (STH sub-domain). Results of CFD analysis that are used in the development of a exible design of the TALL-3D experimental facility are presented. The facility consists of a lead-bismuth eutectic (LBE) thermal-hydraulic loop operating in forced and natural circulation regimes with a heated pool-type 3D test section. Transient analysis of the mixing and strati cation phenomena in the 3D test section under forced and natural circulation conditions in the loop show that the test section outlet temperature deviates from that predicted by analytical solution (which the 1D STH solution essentially is). Also an experimental validation test matrix according to the key physical phenomena of interest in the new experimental facility is developed. In the second part of the thesis we consider the risk related to steam generator tube leakage or rupture (SGTL/R) in a pool-type design of lead-cooled reactor (LFR). We demonstrate that there is a possibility that small steam bubbles leaking from the SGT will be dragged by the turbulent coolant ow into the core region. Voiding of the core might cause threats of reactivity insertion accident or local damage (burnout) of fuel rod cladding. Trajectories of the bubbles are determined by the bubble size and turbulent ow eld of lead coolant. The main objective of such study is to quantify likelihood of steam bubble transport to the core region in case of SGT leakage in the primary coolant system of the ELSY (European Lead-cooled SYstem) design. Coolant ow eld and bubble motion are simulated by CFD code Star-CCM+. First, we discuss drag correlations for a steam bubble moving in liquid lead. Thereafter the steady state liquid lead ow eld in the primary system is modeled according to the ELSY design parameters of nominal full power operation. Finally, the consequences of SGT leakage are modeled by injecting bubbles in the steam generator region. An assessment of the probability that bubbles can reach the core region and also accumulate in the primary system, is performed. The most dangerous leakage positions in the SG and bubble sizes are identi ed. Possible design solutions for prevention of core voiding in case of SGTL/R are discussed.

Thesis (Nucl.E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1981. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by El-Khider Si Ahmed. / Nucl.E.

Thesis. 1975. Nucl.E.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / Bibliography: leaves 153-157. / Nucl.E.

Thesis. 1976. Ph.D.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / Microfiche copy available in Archives and Science. / Includes bibliographical references. / by Pascal De Laquil, III. / Ph.D.

The aim of this work was to rationalize the corrosion resistance of candidate austenitic iron-chromium-nickel (Fe-Cr-Ni) alloys in supercritical water (SCW) for use as fuel claddings within the in-core structure of the Canadian supercritical water-cooled reactor (SCWR) concept. High chromium (Cr)-containing alloys (Alloy 800HT with 20.6 wt.% Cr and 30.7 wt.% Ni and Alloy 33 with 33.4 wt.% Cr and 31.9 wt.% Ni) in the mill annealed condition were chosen for this purpose. Coupons were exposed on a short-term basis (500 h) in a static autoclave containing 25 MPa SCW at 550 °C and 625 °C. Gravimetric measurements and electron microscopy techniques were then used to study the oxidation/corrosion resistance of two alloys. Alloy 33 was found to exhibit the higher corrosion resistance at both temperatures. The improved corrosion resistance of Alloy 33 was attributed to two factors: (i) the formation of a continuous Cr-rich corundum-type M2O3 (M= Cr and Fe) oxide layer that prevented the diffusion of Fe and the formation of a less-protective Fe/Mn-Cr spinel ([Fe,Mn]Cr2O4) outer layer, (ii) a sufficient residual bulk Cr in the Cr-depleted layer adjacent to the alloy/scale interface that prevented any localized internal oxidation from occurring. A mass balance conducted on the corroded Alloy 33 material suggested that volatilization of the corundum-type oxide layer did not occur, at least not within the short-term exposure in the essentially deaerated SCW. A key issue requiring further study was the observation of intermetallic precipitates that formed below the Cr-depleted zone adjacent to the alloy/scale interface in both alloys when exposed for 500 h at 625 °C and their possible influence on the in-service mechanical integrity. / Thesis / Master of Applied Science (MASc) / The supercritical water-cooled reactor (SCWR) is one of the six reactor design concepts developed by the Generation-IV International Forum (GIF). Canada is planning to build the SCWR within the next decades. However, selection of proper materials that perform well within such high pressure high temperature circumstances inside the reactor core with minimum degradation is a very imperative challenge. The current work has addressed this issue by studying the corrosion behaviour of Fe-Cr-Ni alloys in similar environment using electron microscopy techniques.

A study was undertaken to create a finite element model of a cooled throat converging/diverging rocket nozzle to be used as a tool in designing a cooled aerospike nozzle. Using ABAQUS, a simplified 2D axisymmetric model was created featuring only the copper throat and stainless steel support ring, which were brazed together for the experimental test firings. This analysis was a sequentially coupled thermal/mechanical model. The steady state thermal data matched closely to experimental data. The subsequent mechanical model predicted a life of over 300 cycles using the Manson-Halford fatigue life criteria. A mesh convergence study was performed to establish solution mesh independence. This model was expanded by adding the remainder of the parts of the nozzle aft of the rocket motor so as to attempt to match the transient nature of the experimental data. This model included variable hot gas side coefficients in the nozzle calculated using the Bartz coefficients and mapped onto the surface of the model using a FORTRAN subroutine. Additionally, contact resistances were accounted for between the additional parts. The results from the preliminary run suggested the need for a parameter re-evaluation for cold side gas conditions. Parametric studies were performed on contact resistance and cold side film coefficient. This data led to the final thermal contact conductance of k=0.005 BTU/s•in.•°R for contact between metals, k=0.001 BTU/s•in.•°R for contact between graphite and metal, and h=0.03235 BTU/s2•in.•°R for the cold side film coefficient. The transient curves matched closely and the results were judged acceptable. Finally, a 3D sector model was created using identical parameters as the 2D model except that a variable cold side film condition was added. Instead of modeling a symmetric one or two inlet/one or two outlet cooling channel, this modeled a one inlet/one outlet nozzle in which the coolant traveled almost the full 360° around the cooling annulus. To simplify the initial simulation, the model was cut at the barrier between inlet and outlet to form one large sector, rather than account for thermal gradients across this barrier. This simplified nozzle produced expected data, and a 3D full nozzle model was created. The cold side film coefficients were calculated from previous experimental data using a simplified 2D finite difference approach. The full nozzle model was created in the same manner as the 2D full nozzle model. A mesh convergence study was performed to establish solution mesh independence. The 3D model results matched well to experimental data, and the model was considered a useful tool for the design of an oxidizer cooled aerospike nozzle.

With the advent of terrestrial digital broadcasting, new and improved digital transmitter technologies are required since existing analog transmitter technology is, for the most part, unable to adequately transmit a decodable digital television signal. This study focuses on the design and construction of a solid-state, liquid-cooled UHF digital television transmitter. Emphasis is placed on the design of the amplifier module including the amplifier card, Wilkinson splitter and combiner, input and output matching circuits, DC bias network and the system mask filter. The results of this research are also presented for two television transmitters that are installed and continue to be in use today, including analyses of specific failures that have occurred while in the field. The overall objective of this study is not only to document the research that is behind the design of this system, but also to document the construction of the transmitter for reference in system maintenance and repair as well as a basis for future design.

No description available.