Mechanical Engineering
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Item Physics-of-Failure Methodology for Accelerated Thermal Cycling of LCC Solder Joints(1995) Rothman, Timothy Paul; Dasgupta, Abhijit; Mechanical Engineering; University of Maryland (College Park, Md); Digital Repository at the University of MarylandThis research presents a case study were existing physics-of-failure models and Bayesian statistical methods are used in conjunction to quantify the test-time compression achieved during accelerated temperature cycling tests on leadless solder joints. Different combinations of substrate materials and package styles are evaluated with physics-of-failure models and calculable information is obtained from a relatively small population of test specimens under accelerated stresses, because the critical variables are identified, and their influences on the stress magnitude are quantified. Bayesian statistical analysis is employed to obtain an acceleration transform, determine the confidence on the calculations, and determine which outliers are contaminating the database. In addition to accelerating the stress levels, the total test time is further minimized by tailoring the stress drivers in each sample such that multiple stress levels can be achieved under a single loading, which eliminates the need for repeating the test at multiple load levels. This research presents the details of how the models and statistical methods are applied, the results of evaluating different material combinations and package styles, problems encountered during the test, and a summary of the acceleration transforms obtained from the test. Analytical predicative models for life predictions are essential and will obviously result in significant savings of cost and time. The methods used in this are general enough to be applied to screening, qualification, and reliability enhancement tests of a wide range of new or existing electronics assemblies.Item 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.Item 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.Item 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.Item 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.Item 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.Item 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.Item A Comparison of Some Analytical Models with Experiment for the University of Maryland Reactor(1968) Ross, Malcolm Fred Jr; Duffey, Dick; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)It is often desirable in the operation of a research reactor to make adjustments in the nuclear core configuration. These adjustments may occur for a variety of reasons, such as the rearrangement of fuel to perform a particular experiment. It is beneficial to the reactor operator and experimenter to have an adequate analytical model with which to predict the changes in nuclear characteristics which occur with core rearrangement. Several analytical models have been investigated and compared with experimental results for the semipermanent, or normal, core configuration for the University of Maryland Reactor. These models were selected because, while somewhat time consuming with respect to the use of computers, the computer time utilized is much less than needed by more complex methods. At the same time, the methods used tend to minimize the large inherent error associated with simple hand calculations . The methods used consist of a two-dimensional few group diffusion theory coupled with several cross section models from which macroscopic cross sections were obtained. The cross section models used for the above thermal energy groups were the volume integrated P-1 method and the Fourier transform B-1 method. Thermal energy group cross sections we reobtained using the Wigner-Wilkins model and the Maxwell-Boltzmann model. The volume integrated P-1 model and the Wigner-Wilkins model coupled with the two-dimensional group diffusion method were found to give the best agreement with experiment for the semi-permanent core configuration. This model was then tested over a range of experiments. The conclusion of this analysis was that the model was capable of predicting, with reasonable accuracy, the changes in core reactivity with core rearrangement.Item Reactivity Measurements on the University of Maryland Reactor by Conventional Methods and Statistical Processes and Comparison with Calculational Methods(1970) Zubieta, Agustin Diaz; Duffey, Dick; Nuclear Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)The accurate experimental determination of nuclear reactor core physics parameters is of great importance for its safe operation. In particular, the accurate determination and prediction of criticality during the initial fuel loading of a nuclear reactor are essential for a safe nuclear reactor startup. Also, the degree of subcriticality or shutdown margin of a nuclear core with its control rods inserted is an important parameter for the operation of a nuclear reactor throughout the core operating lifetime. There are several methods utilized to determine both the criticality and the shutdown margins. All of these methods depend on measuring the response of neutron detectors and the calibration of the control rods. However, neutron detectors respond only to the neutron flux in the cor e in the vicinity of the nuclear detectors. During initial reactor core fuel loading, the reactivity of the core is determined from the multiplication of the neutrons of the startup source by the addition of nuclear fuel. Reactivity is determined from the multiplication factor by a constant which is related to the source-detector geometry in the core. In this research a method was studied which allowed the determination of reactivity independent of the source-detector geometry. Reactivity measurements of the 10 kw University of Maryland pool training reactor (UMR) were made by conventional methods and by a statistical process, the variance-to-mean ration method, and the results were compared with calculational methods. The theoretical method selected to determine the UMR core reactivity was based on the multigroup, multiregion, diffusion theory. The accuracy of the theoretical model was determined for the just critical UMR core. Agreement to within 0.2% ΔK/K was obtained between the control rod measured and the calculated reactivity for the UMR full core, and smaller UMR supercritical cores. The statistical technique of the variance-to-mean ratio of the number of counts for various counting gate openings, as a means to determine the degree of subcriticality, or shutdown margin, has been proven to be an effective method. A BF3, thermal neutron proportional detector, with a sensitivity of 12.1 counts/sec per n/cm2/sec, was placed inside of an eleven feet long aluminum tubing. The end of the tubing containing the detector was inserted in the center Glory Hole of the UMR core. The current pulses from the proportional detector was amplified and fed to a TMC 1024-channel pulse analyzer. The pulses were counted for different Δt gate openings from 10-4 seconds to 10 seconds. For each Δt gate opening, 1023 samples were taken. The printed output from the TMC-1024 was collected, giving the number of counts received per Δt, as well as the integral of all the counts received during a period of time, equal to 1023 x Δt seconds. From the integrated value for the number of counts the average count c for the gate opening Δt was obtained. The printed output was transferred to IBM cards acceptable to a "Reactor Noise" code written for the IBM-7090. This code calculated the average value, c, for each Δt (which gave a check on the validity of the data transferred to the IBM cards by comparing it with the value obtained during the measurement), the standard deviation o, and the variance-to-mean ratio for all the data taken for each Δt seconds gate opening. Plots of the values of the variance-to-mean ratio versus gate openings were obtained for several UMR full core with the rods banked at various degrees of insertion (shutdown margins), and also for various UMR subcritical cores. Measurements of the shutdown margins by the variance-to-mean technique were in agreement with the values obtained from the rod calibration for negative reactivities of less than -1. 00% ΔK/K, and within ten percent for negative reactivities of approximately -2. 0% ΔK/K. Measurements of the reactivity of small UMR cores indicated that for UMR core conditions of 0. 5% ΔK/K subcritical, experiment and theory for 1-Keff were found to be only 6 parts in 100 apart, and for 2. 0% ΔK/K subcritical, experiment and theory were found to be only 8 parts in 100 apart. The variance-to-mean technique was compared to the inverse multiplication method for determination of criticality during the UMR fuel loading, and was found to be a more accurate method, primarily, because of its independence of the source-detector geometry effects. The system utilized for the statistical data processing is exact, however cumbersome, due to the amount of data to process and the amount of peripheral hardware utilized in the reduction of the data. It appears from this study that greater overall counting efficiency for the same amount of statistical data would permit more accurate measurements at larger degrees of subcriticality, perhaps, in the region of -4. 0% ΔK / K to -5. 0% ΔK / K . A system is proposed in this study to measure negative reactivity continuously, and directly, by means of a small computer capable of accepting the output of a multichannel scaler. The computer would have a fixed internal logic capable of calculating reactivity from the variance-to-mean analysis of the neutron detector counts.Item The Stress Field Surrounding the Tip of a Crack Propagating in a Finite Body(1987) Chona, Ravinder; Irwin, George R.; Mechanical Engineering; Digital Repository at the University of Maryland; University of Maryland (College Park, Md)The goal of this dissertation was to establish the relationship between a parameter descriptive of the trajectory of a smoothly curving crack, such as the curvature of the crack path, and the local stress state in the close vicinity of the crack tip. The behavior of fast -running cracks propagating along straight and smoothly curving paths in fracture specimens of various geometries was examined using dynamic photoelasticity and representations of the running crack stress field we redeveloped in terms of the coefficients of a set of infinite series, for both opening and shear mode loading conditions. Analysis of the isochromatic patterns, using local collocation methods based on this stress field representation, allowed the stress state in the neighborhood of the propagating crack-tip to be modelled with a high degree of accuracy and results were obtained for the variations with crack tip position of both the singular and leading non- singular stress field coefficients of interest. The results obtained for quasi-static and rapid crack propagation under opening mode conditions in a ring segment revealed the importance of retaining terms of order (at a minimum) r^1/2 even when only the singular term was to be determined accurately. Furthermore, it was found that the non-singular stress field coefficients varied similarly in both static and dynamic situations, with some variations in magnitude that could be attributed to crack speed. The results from the curved crack experiments also showed systematic variation of the non-singular terms, but more importantly, it was found that the instantaneous curvature of the crack path was related to the magnitude of the lowest order non-singular stress component (the coefficient of the r^1/2 term) associated with the local shear mode of deformation in the vicinity of the tip of the running crack. Furthermore, the results established that the only singularity associated with a crack propagating along a smoothly curving path in a brittle, isotropic material was that associated with the opening mode stress intensity factor, K1, and that the shear mode singularity, KII, was identically equal to zero.