Engineering Commons^™

Internet-Accessible Hot Cell With Gamma Spectroscopy At The Missouri University Of Science And Technology Research Nuclear Reactor, Edwin Grant

Masters Theses

A dual-chambered internet-accessible heavily shielded facility with pneumatic access to the University of Missouri Science and Technology (Missouri S&T) 200 kW Research Nuclear Reactor (MSTR) core has been built and is currently under low level irradiation testing. The facility allows distance users doing collaborative work with Missouri S&T researchers to manipulate and analyze neutron irradiated samples remotely. The system consists of two shielded compartments, one for multiple sample storage (up to 11 samples), and the other for isolated radiation measurements and spectroscopy. The second chamber has multiple detector ports, graded shielding, and has the capability to support gamma spectroscopy using …

Characterization Of The Neutron Flux Spectrum At The Missouri University Of Science And Technology Research Reactor, Zachary Andrew Kulage

Masters Theses

"A new remotely accessible shielded cell is being constructed at the Missouri University of Science and Technology Research Reactor (MSTR). The heavily shielded cell will be able to receive highly irradiated specimens directly from the reactor and will be equipped with radiation-hardened cameras, remote manipulators and gamma spectroscopy. The cell will allow the manipulation and monitoring of highly activated specimens from both a workstation at the MSTR and at remote locations using a Webbased internet interface. The ability to access and control the shielded cell via a remote internet connection will make it useful to a wide variety of users. …

Analogy Based Modeling Of Natural Convection, Vaibhav Khane

Masters Theses

"This research is an extension of previous work on the development of an integrator or resistance-capacitance circuit analogy for natural convection. As a part of a larger project to enhance transport phenomenon on a micro-scale using radiation, this work studied the phenomenon of natural convection. Using experimental techniques and numerical simulations (FLUENT code), it investigated the transient response of a natural convection system. It proposes an integrator circuit analogy for a natural convection system. Experimental investigation with three different fluids indicated that the characteristic time constant of the system is related to the Prandtl number of the fluid. The project …

Equivalent Circuit For Transient Conduction And Convection Systems, Bassam Sabry Mohammad Abdelnabi

Masters Theses

"Transient response of a natural convection system is investigated by numerical simulation using FLUENT code. An Integrator Circuit analogy was recently proposed for natural convection system. The proposed analogy was further confirmed by these recent simulations. New simulation results also suggest that a natural convection system acts as a "Low Pass" filter for transients. Transient transmission/attenuation factor was found to be a function of both fluid properties and the flow characteristics. Transmission/attenuation factor was also found to be a strong function of fluctuation frequency"--Abstract, page iv.

Effects Of Specimen Size Reduction On The Transition Curve Of The Charpy V-Notch Impact Test, Lonnie Eugene Schubert

Masters Theses

"This study was undertaken to correlate the impact properties of ASTM standard full size Charpy V-notch impact specimens to the impact properties of subsize specimens. An ASTM A533-B quenched and tempered plate steel was examined. The fine grained martensitic material had a coarse prior austenite grain size, which resulted in a low upper shelf energy for this material. Three sizes of specimens were included. Specimens were irradiated in the TRIGA Reactor Facility operated by the U. S. Geological Survey in Denver, Colorado, to 1 x 10¹⁹ n/cm² (E > 1 MeV) at 150°C (302°F). Unirradiated and irradiated specimens were …

Dose Rate Determination Of The Highly-Enriched Uranium Fuel At The University Of Missouri-Rolla Reactor In Preparation For Transportation, Alice Ann Netzer

Masters Theses

"An analytical model was developed in order to accurately estimate the dose rates in air of the highly-enriched uranium (HEU) fuel at the University of Missouri - Rolla Reactor. Based upon these results, a method is proposed to remove the fuel from the reactor.

First thermoluminescent dosimeters (TLDs) were exposed under water to each of the fuel elements in the reactor pool, except for the four control rodded elements. A model was developed to use the in-water TLD readings to calculate the dose rate of the fuel elements in air at 1 foot and 3 feet. The fuel was modeled …

Mixed Convection In Laminar/Turbulent Separated Backward-Facing Step Flows, Jen-Tai Lin

Doctoral Dissertations

"Mixed convective heat transfer results for two-dimensional laminar/turbulent flow in a duct with a backward-facing step are reported. The two walls forming the duct are maintained at uniform but different temperatures. The straight wall is maintained at the inlet air temperature, while the downstream section of the stepped wall is heated to a higher temperature. This study focuses on a backward-facing step geometry with an expansion ratio of two. The present laminar study examines a wide range of inlet flow and wall temperature conditions to cover the domain from buoyancy-assisting mixed convective flow, pure forced convective flow, buoyancy-opposing mixed convective …

Mechanical And Neutronic Properties Of Neutron Absorbing Materials, Donald James Buth

Masters Theses

“The primary neutron absorbing material examined in this project was Boraflex™, which consists of a silicone polymer with boron carbide acting as the neutron absorber. Two aluminum based neutron absorbers, Boral™ and Al-B-Ten, were also examined. Both X-ray and neutron radiography were used to examine the Boraflex™. The neutron radiography gave better resolution in detecting fluctuations in the boron carbide distribution; however, X-ray radiography did perform better in detecting flaws in the base material. Tensile tests were performed to determine the uniformity of Boraflex™'s tensile properties. The average tensile strength was calculated to be 303 ± 15 psi. A small …

Fuel Loading For A Low-Leakage Core In A Pressurized Water Reactor, Omar Ivan Aguilar

Masters Theses

"A study of low-leakage core fuel loading as a means to reduce fast neutron flux incident upon the pressure vessel is presented. The specific case under study is Cycle 5 of Arkansas Nuclear One - Unit Two (ANO-2), a Combustion Engineering Pressurized Water Reactor. Neutron transport calculations were performed using the DOT two-dimensional transport code. The results from DOT provide the angular distribution of the fast neutron flux at the surface of the pressure vessel.

Seven different fuel loading schemes were studied and compared to the Reference Case. The angular distribution of the fast neutron flux for all cases studied …

Numerical Simulation Of Mixing Processes In Two-Dimensional Channel Flows, Milan Karel Straka

Doctoral Dissertations

"Recently, considerable interest has arisen in the numerical simulation of laminar and turbulent transport phenomena. More knowledge in this field is needed to improve predictions of heat and mass transfer processes in a variety of technical fields.

In this work a new computationally efficient transport model is developed. It consists of partially parabolized time-dependent equations of motion, continuity, and concentration of nonconservative species. Pertinent equations are solved numerically for both laminar and turbulent flow.

Results of calculations of laminar flow in the entry region of a parallel-plate channel agree well with the work of other authors. For turbulent flows, the …

Analyzing And Predicting Natural Circulation In A Snupps's Pwr Using "Retran-01", Roger Lee Boyer

Masters Theses

"The purpose of this project is to predict the natural circulation start-up test for the Standardized Nuclear Unit Power Plant System (SNUPPS). In order to make this prediction, a best-estimate, thermal-hydraulic computer code called RETRAN-01 is used.

The start-up test procedures require the simultaneous tripping of all four reactor coolant pumps (RCPs). Natural circulation is confirmed when the three test acceptance criteria have been met. Since this transient is symmetrical, a one-loop approximation of the Nuclear Steam Supply System (NSSS) is normally considered adequate in the nuclear industry. However, there is a concern over how to model the four actual …

A Theoretical And Experimental Investigation Of Single Phase Flow Through A Small Centrifugal Pump In Three Quadrants Of Operation, Lee Charles Cadwallader

Masters Theses

"The operating characteristics of a conventional centrifugal pump have been investigated, both theoretically and experimentally, in three modes of operation with single phase liquid water at atmospheric pressure and room temperature.

Simple one-dimensional theory has been used to evaluate the theoretical pump performance, and these evaluations are compared to actual experimental data. The actual data shows the degradation effects of flow friction and shock losses that the theory does not take into account. The comparison of the curves illustrates these differences.

The data obtained compares well to and reinforces data from experiments found in the literature review.

Based on the …

An Alternate High Pressure Injection System For Pressurized-Water Reactors, Gary Edward Mueller

Doctoral Dissertations

"To evaluate the In-Core Injection System (ICIS) performance during a loss-of-coolant accident (LOCA) due to a major rupture in the primary system of a pressurized water reactor (PWR), tests were conducted using the thermal-hydraulic computer code RELAP4/MOD5.

The results for the overall plant system show that the maximum core average fuel cladding temperature reached during an intermediate size cold-leg break in a Combustion Engineering PWR with ICIS is equal to the initial operating temperature of 335°c (636°F). The maximum core average fuel cladding temperature reached during a large size cold-leg break with ICIS is 478°c (893°F). This temperature is approximately …

Utilization Of Waste Heat From The University Of Missouri Research Reactor., Anek Punyakumleard

Masters Theses

"An analysis has been conducted to determine the economic as well as technical feasibility of using waste heat from the University of Missouri Research Reactor (MURR) to heat the reactor containment building and the associated laboratory building.

Several methods of waste heat utilization were reviewed, including open field agriculture, animal shelter heating, aquaculture, waste water treatment, and space heating. Based on this review, it was concluded that several of these waste heat utilization methods are feasible for the MURR. However, the most practical means of utilizing the MURR waste heat appears to be for space heating in the reactor containment …

A Fluid Control System For The University Of Missouri--Rolla Nuclear Reactor, Billy Roger Mcpherson

Masters Theses

"Solid control rods and their movement through a reactor core present problems inherent in any moving machinery. The problems of binding, wear, and noise have to be dealt with in the usage of a solid control rod system. Therefore, an alternate system that eliminates or minimizes these problems would lower maintenance costs, extend the lifetime of the control system, and produce safer operating conditions.

This thesis presents a control system that has no moving parts inside the reactor core. The system is based upon a neutron-absorbing fluid containe􀁠d in hollow tubes in the reactor core. The fluid level is operated …

Effect Of Photoneutrons On Transients Of Beryllium Reflected Reactors, Safdar Ahmad Ali

Doctoral Dissertations

"An analytical model is developed to account for the photoneutron sources present in a beryllium reflector. The time-dependent, two- dimensional and three energy-group diffusion equations are written with an additional photoneutron source term. Then an approximate flux composed of two spatial shapes chosen beforehand, each having an unknown time coefficient, is inserted into time-dependent multigroup equations and the weighted residual criterion is applied. This yielded multimode kinetics equations with generalized definitions for conventional matrix parameters: generation time, reactivity, delayed neutron and delayed photoneutron fraction matrices.

Two transients are analyzed for the University of Missouri Research Reactor: simulation of rod drop …

Experimental Study Of Electron Density And Confinement Time In An Electrostatically-Plugged Cusp Device Using A Microwave Interferometer, Robert Louis Hayward

Masters Theses

"The average electron density and electron confinement time in an electrostatically plugged magnetic spindle cusp device are studied using a 20 GHz microwave interferometer. The results are compared with the predictions from approximate theoretical equations. With a magnetic induction in the point cusp of 5000 Gauss, a plasma is produced by injecting a 5 mA, 500 eV electron beam into hydrogen gas at 10^-5 Torr. The measured electron density is 2 x 10^-10 cm^-3 and the measured confinement time is 80 ysec, which agree to within a factor of two with values predicted by theory"--Abstract, page ii.

Calculation Of The Neutron Flux In The Side Regions Of The Prestressed Concrete Reactor Vessel Of A High Temperature Gas Cooled Reactor, Richard Robert Kent Jr.

Masters Theses

"Neutron flux in the side regions of the High Temperature Gas Cooled Reactor's (HTGR) Prestressed Concrete Reactor Vessel (PCRV) was determined using the Monte Carlo method. One quarter of the PCRV was modeled, with all major structures inside it, e.g. steam generator, auxiliary cooling loop, etc. imaged at their exact position and size. Initial neutron energies range from .001 eV to 3 MeV. Three neutron interactions were considered: absorbtion, elastic scattering, and inelastic scattering.

The flux was calculated for two different PCRV temperatures, 155° F and 647° F. The thermal flux showed a minor increase at the higher temperature. In …

Possible Damage Mechanism In Dresden 2 Class Boiling Water Reactor Fuel, Donald Lee Moffett

Masters Theses

"A thermal hydraulic analysis of a typical hot channel of a Dresden 2 class boiling water reactor is studied for possible severe local overheating. The COBRA-II thermal hydraulic analysis code is modified to include critical heat flux calculations for each subchannel. The effects of flow distribution, inlet mass velocity variations, dimensional tolerances, enrichment variations, and input parameters are examined in detail. Bulk channel results are in good agreement with the published data, but the assembly wall side of the corner fuel rod has a minimum critical heat flux ratio of less than unity for a number of the situations examined. …

Waste Heat Utilization From A Direct Cycle High Temperature Gas Cooled Nuclear Reactor For District Heating And Air Conditioning, John Joseph Blase

Masters Theses

"An analysis was conducted to determine the economic as well as technical feasibility of waste heat utilization from the proposed direct cycle high temperature gas cooled nuclear reactor, as designed by the General Atomic Company.

The rejected heat from this system is at considerably higher temperatures than those normally encountered in conventional steam-electric Rankine cycles. By taking advantage of these higher rejection temperatures, heat was translated into energy available to a district heating and air conditioning service. The transportation of this energy was considered to be in the form of heated and chilled water.

A refrigeration capacity on the order …

The Effects Of Washout In Polydispersed Metalic Aerosols, Jerry N. Burford

Masters Theses

"Experimental Collection Efficiencies are presented for drops of approximately 0.1cm radius falling through aluminium, magnesium and molybdenum oxide polydispersed aerosols. The experimentation was conducted in two identical cylindrical chambers, one being used for aerosol washout, and the other for allowing the aerosol to decay normally. The Collection Efficiency data were thus determined by comparing the aerosol concentration changes in the two chambers. The rather strong effects that induced micro-turbulence and relative humidity have on aerosol collection and aerosol decay were observed and are discussed"--Abstract, page ii.

Reliability Analysis Of The University Of Missouri--Rolla Reactor, Jimmy Dale Schottel

Masters Theses

"The reliability of Nuclear Installations is becoming an important area of research. In this report an approach to reliability analysis was made. The actual failure data was utilized to predict theoretical failure distributions. These distributions were then used to determine future reliability of components, subsystems and system"--Abstract, p. ii

Post Irradiation Changes Of The Elastic Modulus Of Aluminum, Dale A. Powers

Masters Theses

"Aluminum samples of 99.99% and 99.999% purity were irradiated to neutron fluences of 3·8·10¹⁸ nvt and 1.4·10¹⁹ nvt (E greater than 0.1 Mev) at one-third of the melting temperature. The change in Young's modulus of elasticity after isothermal annealing in the temperature range of 110⁰C to 130⁰C was measured. After 2.5 hours of annealing, the modulus reached a steady state value which was decreased by 0.25% ± 0.17% from the unannealed value. Electron microscopy suggests the hardening mechanisms were dispersion hardening by silicon precipitates and void hardening"--Abstract, page ii.

Calculation Of Gamma-Ray Albedo Using The Monte Carlo Method, Ezatholah Aslani-Amoli

Masters Theses

"Gamma-ray number, energy and dose albedo are calculated with the Monte Carlo technique. A plane monoenergetic beam was considered at normal incidence upon single material shields at initial gamma energy up to 10 MeV. Aluminum, iron and lead are used as shield material. The contribution of secondary photons which are produced as a result of various photon interactions in the shield are taken into account. These secondary photons are annihilation gammas, X rays and bremsstrahlung gammas. In certain cases, they constitute a considerable fraction of the reflected radiation.

Annihilation gammas predominate in the reflected radiation for all energies above 1.02 …

Neutronics Calculations Of Two Fusion Reactor Blankets, James Edward Struve

Masters Theses

"At the present state of engineering development, the most probable fusion reactor fuel cycle will use deuterium and tritium in the reaction ²₁D + ³ ₁T → ¹₀n + ⁴₂He. This reaction is exothermic and provides 17.6 Mev of energy of which 14.1 Mev are given to the neutron. The reaction will take place in a plasma environment inside an evacuated chamber. This chamber will be surrounded by a "blanket" which will serve to transform the kinetic energy of the neutron into heat and produce tritium for subsequent use in the fuel cycle. …

Break-Even Criteria For A Pulsed D-T Reactor, Kenneth Franklin Keller

Masters Theses

"To those who keep track of such things and are in a position to make predictions, it appears that the use of controlled thermonuclear reactions to economically produce electric power is well within a lifetime from now. Indeed, experiments presently being conducted are putting us closer and closer to the development and construction of demonstration fusion reactors. Understandably, the people working on such products and processes are optimistic about their results. This is as it should be if our nation is to avoid major energy shortages in the future.

Unfortunately, the tenor of this paper is pessimistic. This is not …

Use Of Radioactive Iodine Evolved Into Reactor Cooling Water To Locate Defective Fuel Elements, Carl Mozel Stroud

Masters Theses

"The relative amount of radioactive iodine evolved in reactor cooling water by each fuel element provides an excellent indication of defective cladding on fuel elements. The radioactive iodine evolved into the reactor cooling water from a defective fuel element should be higher than the average radioactive iodine evolved by other fuel elements into the cooling water by an order of magnitude or more.

Radioactive iodine with non-radioactive iodine carrier is adsorbed on DOWEX SBR resin in the iodide form. The iodide is eluted from the resin with sodium hypochlorite and extracted into carbon tetrachloride. The iodine is reduced to iodide …

Computer Simulation Of The Primary Recoil Spectra And Damage Cross Section Of Fe₃Al Using Endf/B Ii Data, Daniel Willard Smith

Masters Theses

"A computer program, using Evaluated Nuclear Data File (ENDF/B II) information, was used to develop energy-exchange kernels which delineate the energy transfer probabilities between neutrons and Fe₃Al lattice atoms. The kernels combine all available information on elastic and inelastic scattering contained in the ENDF/B II data.

The computer program used to generate the primary recoil spectra is SAKI, a modified version of RICE (1). SAKI is designed to calculate energy-exchange probabilities in any binary alloy as well as damage cross sections and optimum cutoff energies for use in comparing displacement effects in different reactor spectra."--Abstract, page ii.

Calculation Of Buildup Factors For Multilayer Slab Shields Using The Monte Carlo Method, John Paul Kuspa

Masters Theses

"Gamma-ray beams normally incident upon slab shields of varying compositions are studied using a computer program based upon the Monte Carlo technique. Initial gamma energies of 1, 4, 6, and 8 MeV are considered. The shielding materials used were aluminum, iron, and lead. Both single and double material shields were investigated, at a variety of thicknesses up to a maximum of five mean free paths. Two secondary gamma processes are included in this simulation: annihilation gammas from pair production and bremsstrahlung from the electrons. The primary effort is the calculation of the dose, energy, and number buildup factors for these …

Use Of Isotopes To Reduce Neutron-Induced Radioactivity And Augment Thermal Quality Of The Environment Of An Underground Nuclear Explosion, Nathaniel Fred Colby

Masters Theses

"The use of isotopes to include radioactive waste products to reduce the neutron-induced activity of an underground nuclear explosion and its application in the field of geothermal power stimulation is discussed. A shield composed of selected isotopes surrounding a fusion device will capture excess neutrons producing isotopes with short half-lives. Subsequent rapid decay will prolong the high temperature in the vicinity of the explosion and decrease the activity. Long-lived isotopes created by neutron capture in the host rock would be minimized. The investigation is conducted for strontium-90, cesium-137, cerium-144, technetium-99, rhodium-103 and -105. Of these, the first three appear to …