Report (M.S.)--Virginia Polytechnic Institute and State University, 1994. / Vita. Abstract. Includes bibliographical references (leaf 130). Also available via the Internet.

Doctor of Philosophy / Department of Mechanical and Nuclear Engineering / Hitesh Bindra / Helium-cooled, graphite moderated reactors have been considered for a future fleet of high temperature and high efficiency nuclear power plants. Nuclear-grade graphite is used in these reactors for structural strength, neutron moderation, heat transfer and, within a helium environment, has demonstrated stability at temperatures well above HTGR operating conditions. However, in the case of an air ingress accident, the oxygen introduced into the core can affect the integrity of the fuel graphite matrix. In this work a combination of computational models and mixed effects experiments were used to better understand the air ingress process and its potential effects on the heat removal capabilities of an HTGR design following an air-ingress accident. Contributions were made in the understanding of the air-ingress phenomenon, its potential effects on graphite, and in experimental and computational techniques. The first section of this thesis focuses on experimental and computational studies that were undertaken to further the understanding of the Onset of Natural Convection (ONC) phenomenon expected to occur inside of an HTGR following an air ingress accident. The effects of two newly identified factors on ONC – i.e., the existence of the large volume of stagnate helium in a reactor's upper plenum, and the possibility of an upper head leak – were investigated. Mixed-effects experimental studies were performed to determine the changes induced in nuclear grade graphite exposed to high-temperature, oxidizing flow of varying flow rates. Under all scenarios, the thermal diffusivity of the graphite test samples was shown to increase. Thermal conductivity changes due to oxidation were found to be minor in the tested graphite samples – especially compared to the large drop in thermal conductivity the graphite is expected to experience due to irradiation. Oxidation was also found to increase the graphite's surface roughness and create a thin outer layer of decreased density. The effects of thermal contacts on the passive cooling ability of an HTGR were experimentally investigated. Conduction cool down experiments were performed on assemblies consisting of a number of rods packed into a cylindrical tube. Experimental conditions were then modeled using several different methodologies, including a novel graph laplacian approach, and their results compared to the experimentally obtained temperature data. Although the graph laplacian technique shows great promise, the 2–D Finite Element Model (FEM) provided the best results. Finally, a case study was constructed in which a section of a pebble bed reactor consisting of a number of randomly packed, spherical fuel particles was modeled using the validated FEM technique. Using a discrete elements model, a stable, randomly packed geometry was created to represent the pebble bed. A conduction cool down scenario was modeled and the results from the FEM model were compared to best possible results obtainable from a more traditional, homogeneous 1–D approximation. When the graphite in the bed was modeled as both oxided and irradiated, the homogeneous method mispredicted the maximum temperature given by the 3–D, FEM model by more than 100°C.

This thesis deals with the severe accident of the gas-cooled fast reactor GFR. At the beginning of the study there is a review of the gas-cooled fast reactor subject. Next part is focused on description of possible solutions for severe accidents with emphasis on the solution applied in the Generation III+ reactors. Chapters that deal with material and thermal balance with severe accident of GFR demonstration unit, along with the chapter which analyses features of the corium, create a basis for the conceptual design of core catcher of GFR demonstration unit, which forms the final part of this thesis.

Thesis. 1977. M.S.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / M.S.

Thesis. 1975. M.S.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / Includes bibliographical references. / This is a report of the experimental results of a program in countercurrent flow critical heat flux. These experiments were performed with Freon 113 at 200 psia in order to model a high pressure water system. An internally heated annulus was used to model a fuel pin in a channel. Only low flowrates were examined. The flow regime was always bubbly or slug, with the liquid primarily on the walls. It was found that critical heat flux of about .9 of the pool boiling value can be expected for up to 20% void. Beyond this, the CHF value decreases uniformly with void. Void fractions up to 70% were investigated, with the results being slightly conservative compared to a similar experiment carried out at atmospheric pressure by Avedisian. / M.S.

Thesis. 1977. Nucl.E.--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / M̲i̲c̲ṟo̲f̲i̲c̲ẖe̲ c̲o̲p̲y̲ a̲v̲a̲i̲ḻa̲ḇḻe̲ i̲ṉ A̲ṟc̲ẖi̲v̲e̲s̲ a̲ṉḠS̲c̲i̲e̲ṉc̲e̲. / Includes bibliographical references. / by Farid Bamdad-Haghighi. / Nucl.E.

Thesis (Nucl. E.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering; and, (M.S.)--Massachusetts Institute of Technology, Dept. of Ocean Engineering, 1981. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Vita. / Includes bibliographical references. / by Guillermo Baltra Aedo. / M.S. / Nucl.E.

Thesis (B.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1982. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE / Includes bibliographical references. / by John Jay McCauley. / B.S.

<p> The feasibility of adaptive control of a nuclear reactor is investigated. For practical reasons, an actual operating power plant is chosen, and a digital computer model developed for the reactor and associated control system. The effects of parameter variations on the transient response of the overall system are studied, and the advantages of using an adaptive controller established. An algorithm for the adaptation scheme is developed, and applied successfully to control the nuclear reactor. </p> / Thesis / Master of Engineering (MEngr)

<p> This work was the first study at the beam port # 4 at the McMaster Nuclear Reactor, involving prompt gamma in vivo neutron activation analysis. The project consisted of experimental and computational parts. The computational part was done using MCNP program, which simulates the neutron and photon transport in the medium. The first thing assessed was the energy dependent neutron fluence rate in the collimated neutron beam, at the site. This was done in order to figure out the complete source (sdef) card for further MCNP calculations. This was combined experimental and computational work. For the experimental part, various activation foils were used and computational part was done by using MCNP programming.</p> <p> The second part of the project involved experimental prompt gamma in vivo activation analysis using 7 different phantoms, ranging from 30 mL to 2 L. Three different elements were observed. The prompt gamma in vivo detection of cadmium was the preliminary calibration study and the experiments were done with all seven phantoms. The calibration lines and MDL were assessed for all phantoms, with concentration ranging from 0 to 50 ppm. The prompt gamma in vivo detection of boron and mercury was done using 30 mL phantoms. Calibration lines and MDL for both elements were assessed as well.</p> <p> MCNP experimental simulations for 30 mL water phantoms were done and they were in close agreement with the experimental results. Furthermore, the MCNP gamma and neutron dose survey in the cave was done.</p> <p> The results obtained showed that there are numerous open possibilities for improvement in terms of in vivo prompt gamma analysis at the site. It predominantly includes the improvements in prompt gamma detection techniques and MCNP source definition. Furthermore, it was found that MCNP programming is the ideal tool for assessment and control of the experimental results in this case. It means that in the future research, the MCNP modeling will be the essential part of the in vivo prompt gamma activation analysis at this beam port.</p> / Thesis / Master of Science (MSc)