A. James Clark School of Engineering

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The collections in this community comprise faculty research works, as well as graduate theses and dissertations.

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End date: 1979

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    Battery Studies with Particular Reference to Organic Depolarizers
    (1955) Monson, William L.; Huff, W. J.; Chemical Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, MD)
    Since Volta's invention of the first primary cell, using silver and zinc, numerous other cell combinations have been studied, covering a wide variety of anode and cathode materials. The latter have included both inorganic and organic substances capable of electrochemical reduction, although, historically, organic cathode materials have received very much less attention than the inorganic. It was the purpose of this investigation to study the actual behavior of a selected number of quinones as depolarizers in primary cells. Performance of experimental cells was compared with cells of the usual dry cell composition but of the same size and construction as cells of experimental composition. The results show that certain substituted anthraquinones possess good depolarizing ability as measured by discharge voltage and coulombic capacity. Energy output in some cases was higher than that of the manganese dioxide control cells (zinc anodes in all cases) because of higher effective coulombic capacities. A qualitative study of the effect of substituents on the discharge voltages of various quinones showed that cell working voltages were much more sensitive to quinone substitution than were the calculated reversible potentials. Also, in the case of nitro-substituted anthraquinones more coulombic capacity was obtained than could be accounted for by the simple reduction to the corresponding hydroquinone. The possibility of a reduction of the nitro-group of this compound was considered. Substances investigated were benzoquinone, naphthoquinone, anthraquinone, and certain of their derivatives, using various electrolytes. The size of the experimental cells was such that about 0.2 gram of the various depolarizers could be studied conveniently.
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    Hinge-Bill Orientation Techniques for Automated Oyster Processing
    (1977) Gird, John; Wheaton, F.W.; Bioengineering; Digital Repository at the University of Maryland; University of Maryland (College Park, MD)
    The width and thickness dimensions of oysters and an inclined V-shaped trough were studied as means for achieving end orientation. Two series of experiments were conducted on 2,430 oysters sampled from three different locations in the Chesapeake Bay. Both width and thickness were measured every 0.2 inch along the oyster length from the hinge to the bill end. A width to thickness ratio was found to be the best dimensional combination for distinguishing between the hinge and bill ends. Less than 0.50 percent of all oysters failed the ratio test conditions. Statistical analysis on five width to thickness ratio tests with failure rates between 0.25 and 0.49 percent showed there to be no differences in the percent oyster failure over all bars and across all tests. Results indicate that comparable oyster orienting efficiencies can be attained by width to thickness ratios with orienting points located 0.4 to 1.0 inches in from the oyster ends. Negative results occurred when an inclined V-shaped trough was used for orienting oysters. There were significant differences in the proportion of hinge and bill leading oysters exiting the trough for each trough loading position over all bars and oyster axes. The tendency for the oyster axes to behave differently explained some of the differences in the trough's orienting efficiency. However, there were no significant relationships between orienting efficiency and oyster axes.
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    The Determination of Preferred Orientation in Rolled Electrical Steels Using Single Diffraction of Neutrons
    (1963) Eugenio, Manuel; Duffey, Dick; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, MD)
    Preferred orientation in rolled electrical steels has been determined using single diffraction of neutrons from the University of Maryland pool-type nuclear reactor (DMR) operating at 10 KW thermal . X-rays are used extensively to determine preferred orientations in metallic wires and rolled sheets, but X-rays suffer the disadvantage of high absorption and cannot be used effectively on thick samples without chemical or mechanical treatment which ultimately results in the destruction of the samples. The use of reactor neutrons for this purpose is believed to offer particular advantages such as the use of thicker samples and wider beams. To this end, neutrons from the UMR were scattered directly from metallic sheet samples to obtain diffraction patterns from which preferred orientations of the crystallographic axes could be deduced. The neutron diffraction data were obtained in the form of : 1) Maxwellian curves; and 2) rocking curves. To obtain the first type of curve, the sample and neutron detector were rotated at a 1-to-2 angular ratio, respectively, and the diffraction pattern was essentially the Maxwellian neutron energy distribution. From the maximum of the Maxwellian curve, the crystallographic plane mainly responsible for the reflection was calculated; from this, the main orientation was deduced. For the second type of curve, the sample was rocked back and forth, with the neutron detector fixed, and the resulting pattern was used to infer the variation of a given crystallographic direction about its main orientation. The results of this study, particularly on grain-oriented and cube-textured silicon-iron (Si-Fe) alloy sheets demonstrate that single diffraction techniques can be used to determine preferred orientation in highly oriented materials. The results on Si-Fe sheets described as non-oriented indicate the possibility that these techniques may be applicable to ordinary rolled metallic sheets, which are not highly oriented.
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    Local Atomic Arrangments and Solution Strengthening of Ta-Mo and Ta-Nb Alloys
    (1975) Predmore, Roamer Edward; Arsenault, Richard J.; Chemical Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)
    Ta-Nb alloys are shown to form random solid solutions by x-ray diffuse scattering measurements. These alloys have equal size atoms in their pure state with lattice parameters that are invariant in composition, obey Vegard's law, and exhibit an absence of solid solution hardening and an absence of fracture embrittlement at high solute concentrations. Ta-Mo atoms of about 5% difference in atomic size form short range ordered solid solutions with large atomic displacement effects. The Ta-Mo, and Ta-W, Nb-Mo and Nb-W alloys have in common a lattice parameter that varies in composition with a negative deviation from Vegard's Law. There is also a negative heat of mixing which is well correlated with short range order. In addition, all these alloys show linear solid solution hardening to high solute concentrations at room temperature and fracture embrittlement at high solute concentrations. Diffuse ex-ray scattering measurements on Ta-Mo alloys give the short range order parameters and atomic size displacements. The hardening is attributed to a combination of size effect induced substitutional solid solution hardening and short range order induced hardening.
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    Kinetic Parameters of the University of Maryland Reactor by the Interval-Distribution Method
    (1976) Simonson, Simon Christian III; Munno, Frank J.; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)
    The Rossi alpha of the University of Maryland Reactor was measured at criticality and at shutdown by the Babala interval-distribution method. At criticality, α= 188.8 ± 4.7 sec^-1, and at shutdown, α= 1026.6 ± 4.1 sec^-1. The shutdown reactivity was found to be ρ/β = -4.44 ± 0.14 dollars.
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    A Computer Study of Nuclear Characteristics of the University of Maryland Reactor
    (1967) Altomare, Philip M.; Duffey, Dick; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)
    An analytical study of the University of Maryland Reactor was performed using an IBM 7094 computer. The nuclear analysis codes employed in the study were GAM-I, for fast group cross-section evaluation, Tempest-II, for thermal cross-sections, WANDA, for one-dimensional neutron transport calculations and PDQ, for two-dimensional transport calculations. A computer program, XS-1, for the calculation of effective control rod cross-sections was written as part of this study. The method employed in the determination of fast energy broad group cross-sections for non-fuel regions from that normally used. The neutron leakage from the fuel was used as a source of neutrons in lieu of the fission spectrum. This method was felt to give a better estimate of the neutron flux distribution in energy. The analytical results compared favorably with experimental data. The predicted K- effective was 0.993 versus 1.006 actual. The thermal neutron flux calculated and measured at the center of fuel elements generally agreed within 10%. The calculated neutron flux above 3 Mev, which is more difficult to match, was found to be low by approximately 30%. Control rod worth was predicted to be 5.6% Δρ for the center shim rod and 2.7% Δρ for the outside shim rod. Measured values are reported as 3.6% Δρ and 2.0% Δρ respectively.
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    Gamma Ray Spectra and Shielding Survey of the University of Maryland Reactor
    (1963) Nhiep, Nguyen; Duffey, Dick; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)
    The gamma ray spectra of the University of Maryland reactor were measured at the beam tube which extends from the side of the reactor to the core, the thermal column and the top of the reactor. There were three gamma ray spectra measurements at the beam tube : gamma ray spectra recorded by a single-channel spectrometer when the reactor was shut down and when the reactor was operated at different power levels and the gamma ray spectra recorded by a 256-channel spectrometer when the reactor was at different levels of power. However, only the single channel spectrometer was used to measure gamma ray spectra at the thermal column and the top of the reactor; some gamma ray spectra were recorded at the thermal column when the reactor was operated at various power levels and a gamma ray spectrum was recorded at the top of the reactor when the reactor was operated at full licensed power of 10 kw. The gamma ray shielding survey of the reactor was done by a Geiger-Muller survey meter when the reactor was running at full power of 10 kw. The highest level found was about 0.8 millirem per hour which was at the side of the reactor at core level. A calculated gamma ray level was near this value.
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    The Effect of Nuclear Reactor Radiation on the Electrical and Mechanical Properties of Epoxide Polymers
    (1965) Kincaid, Charles Vernon; Duffey, Dick; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)
    The effect of nuclear reactor radiation on the volume resistivity, bulk compressibility, specific volume, and hardness of three epoxide polymers was studied. The polymers were thermally polymerized and then subjected to varying radiation doses in the enriched uranium, water moderated, heterogeneous nuclear reactor located at the University of Maryland. Changes in the volume resistivity as a function of temperature for the various radiation doses was then determined by the use of an electrometer to measure the potential drop across the specimen. The effect of radiation on the bulk compressibility was determined by means of a special compressibility tester over the pressure range one to 8,000 atmospheres and temperatures 25°C to 250°C. The specific volume and hardness were measured for each sample before and after irradiation using standard techniques. The temperature dependence of the electrical volume resistivity was initially increased by exposure to 2.5 Mrads. Further irradiation up to 25 Mrads caused severe degradation of the temperature dependence of the electrical resistivity for all epoxide polymers studied . The magnitude of the compressibility for the polymers was found to be relatively unaffected by irradiation up to 25 Mrads; however, determination of the pressure-induced glass transition yielded anomalous results. No significant change in the hardness of the polymers was noted after varying amounts of radiation. The specific volume of an aromatic cured epoxide was not affected to any extent , but the aliphatic cured polymer showed a decreased specific volume.
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    Cerenkov Light Production in a Water Moderated Nuclear Reactor
    (1963) Madey, Robert William; Duffey, Dick; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)
    An experimental investigation of the production of Cerenkov radiation in a water moderated nuclear reactor is conducted using a photomultiplier as a light sensor. The variations in light intensity are studied during various phases of reactor operation, namely: startup, steady state and shutdown. The relevant theory is presented as an aid in interpreting and extrapolating the experimental results. It is found that for transients such as startup, the light signal is directly related to reactor power for periods (e - folding time) faster than about 20 seconds. Additional transient data acquired from measurements performed on a TRIGA pulsed-type reactor illustrate the excellent agreement between the Cerenkov detector and a conventional ionization chamber for measuring pulse characteristics such as peak power, pulse half-width, and prompt period. The proportionality between reactor power and Cerenkov signal is no longer valid for whole core measurements made at steady state power level because of the gradual increase of the Cerenkov signal as a result mainly of fission product contributions. Selective scanning of the Cerenkov spectrum through the use of interference filters over the wavelength range 3500 Å to 5530 Å results in a lower buildup fraction. Indications are that measurements further into the short wavelength region may yield a light sensor, and hence a good power detector, independent of any fission product buildup. The decrease in the Cerenkov light intensity after shutdown is measured for reactor operating times from 20 minutes to 4 hours. Comparison of the empirical data with theoretical considerations results in good agreement for shutdown times ranging from 500 seconds to 10,000 seconds. Spectral measurements made through 17 feet of water with a Hilger quartz spectrograph show a spectral distribution ranging from 2500 Å to 6000 Å. A calculated spectral distribution is compared with the measured spectrum after correcting for water attenuation.
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    Fast Neutron Flux in the University of Maryland Reactor
    (1965) Berman, Philip Gilbert; Duffey, Dick; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)
    The fast neutron flux distribution in the experimental facilities of the University of Maryland Reactor (UMR) was studied by means of neutron activation threshold foils with the reactor at a power level of 10 kw. The foils used were sulfur, magnesium, and aluminum which detected neutrons of energies greater than 3.0 Mev, 6.3 Mev, and 8.1 Mev, respectively. The activities of the foils were measured by calibrated beta and gamma scintillation detection systems. A central row of fuel elements, the west beam port, and the through tube were studied using all three types of foils. The fast neutron flux above 3.0 Mev was measured in each element of the core. The maximum fast neutron flux above 3.0 Mev available in the central ''glory hole " was about 2x10^10 n/cm^2 -sec. The flux above 6.3 Mev and above 8.1 Mev was about 5.5x10^8 n/cm^2 -sec and about 4.5x10^8n/cm^2 -sec, respectively. Available in the west beam port was a maximum fast neutron flux above 3.0 Mev of about 2.6x10^9 n/cm^2 -sec adjacent to the core and of about 1.2x10^7 n/cm^2 -sec at the inner face of the beam port shield plug. The fast neutron flux above 3.0 Mev available in the center of the through tube was 1.7x10^9 n/cm^2 -sec. The minimum at both east and west ends, at the inner faces of the shield plugs, was about 7x10^5 n/cm^2 -sec.