Browsing by Subject "TRIGA"
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Item An easier way to a harder spectrum : generating a fast neutron flux spectrum in a Mark II TRIGA reactor using a uranium-boron-cadmium converter(2020-03-26) Pierce, Michael T., Jr.; Charlton, William S.The TRIGA Mark II research reactor has a flux spectrum with a predominance of thermal neutrons. While most research can be conducted using this thermal neutron flux, an exclusively fast neutron flux is preferred for some experiments, such as radiation damage studies. Although fast neutron flux spectrums are available using fast reactors or neutron generators, only a few fast reactors are currently in operation and neutron generators tend to have low fluxes. As a result, this report evaluated the possibility of generating a fast neutron flux spectrum in a thermal reactor through the use of an irradiator consisting of layers of uranium, boron, and cadmium. Simulations using Monte Carlo N-Particle Code (MCNP) determined that thin layers of these materials in an irradiator inserted into the 3-Element Facility of the UT-Austin TRIGA reactor will generate a fast neutron flux spectrum with negligible thermal flux and where over 91% of the total flux consists of neutrons with energy greater than 10 keV. Further research is recommended to calculate heat dissipation and to determine the reactivity worth of the irradiatorItem Characterization of neutron flux spectra for radiation effects studies(2011-05) Graham, Joseph Turner; Landsberger, SheldonThe effects of neutron displacement damage on materials are sensitive to neutron energy spectra. In controlled neutron damage experiments, a well characterized neutron flux spectrum is critical in determining the equivalent dose for displacement damage. Two techniques were used to characterize the neutron flux spectra in the University of Texas at Austin TRIGA research nuclear reactor. The first technique uses a standard method of measuring the reaction rates of two identical metal foils, one of which was irradiated in a Cd cover, the other of which was irradiated bare. Assuming an analytic form of the neutron spectrum the reaction rates were used to determine an approximate spectrum. The second technique uses the reaction rates measured from a set of activated metal foils along with two spectral unfolding techniques to approximate and then refine the neutron spectrum. A Matlab code was developed which fits radiative capture reaction rates to an approximate spectrum using a least squares approach. The result was used as an initial guess in a second Matlab code which refines the epithermal and fast energy ranges of the spectrum using reaction rates from threshold reactions. Errors in the reaction rates calculated from the resulting spectrum to the measured reaction rates were used to assess the accuracy of the final neutron spectrum.Item Determination of fission product yields of 235U using gamma ray spectroscopy(2012-12) Lu, Christopher Hing; Biegalski, Steven R.; Landsberger, SheldonIt is important to have a method of experimentally calculating fission product yields. Statistical calculations and simulations produce very large uncertainties. Experimental calculations, depending on the methods used, tend to produce lower uncertainties. This work set up a method to calculate fission product yields using gamma ray spectroscopy. In order to produce a method that was theoretically sound, a simulation was set up using OrigenArp to calculate theoretical concentrations of fission products from the irradiation of natural uranium. From these concentrations, the fission product yields were calculated to verify that they would agree with expected values. Moving forward in the work, the total flux at the point of irradiation, in the pneumatic transfer system, was calculated and determined to be 3.9070E+11 ± 6.9570E+10 n/cm^2/s at 100 kW. Once the flux was calculated, the method for calculating fission product yields was implemented and yields were calculated for 10 fission products. The yields calculated were in very good agreement (within 10.04%) with expected values taken from the ENDF-349 library. This method has strong potential in nuclear forensics as it can provide a means for developing a library of experimentally-determined fission product yields, as well as rapid post-nuclear detonation analysis.Item Investigation of possible hydrogen shielding effect on epithermal neutron activation analysis - a computation and experimental approach(2010-05) Zhou, Yang "Alex"; Erich SchneiderNeutron activation is a popular analytical technique used to determine the presence and concentration of certain elements. It has several variations, including thermal neutron, epithermal neutron, fast neutron activation, etc, for different applications; all of those variations are non-destructive, and sensitive to small quantity. While trying to determine the concentration of Cl and Br in the light water solution, Dr. Landsberger’s team found the epithermal neutron activation analysis results were 25% lower than the conventional chemical method. They were not able to determine the cause of such discrepancy. This study was motivated to re-examine such discrepancy, and to study its possible causes. Furthermore, the study tries to determine if such discrepancy, if it exists, was linked with thermal neutron cut off or hydrogen absorption of neutrons. A computer simulation using the Monte Carlo radiation transport software MCNPX was developed to radiate sample Cl & Br solutions of known mass concentrations in a simulated TRIGA reactor core at 500 KW steady state power. [1] The neutron activation rate of Br, Cl at each concentration was then calculated. Such procedure was then repeated for heavy water solutions. Finally, a cadmium shield was added to eliminate thermal neutrons; all samples were tested again using epithermal neutron activation. The actual neutron activation experiment was also carried out in the University of Texas’s TRIGA Mark II reactor. A total of 40 samples of Br & Cl solution (with and without Cd, in light water and in heavy water) were irradiated in the reactor at 500 KW steady state power.Item Simulation of cold neutron production using ordinary and deuterated cryogenic organic moderators(2000) Niset, Martin Luc; Wehring, Bernard W.The objective of this thesis is to investigate the use of deuterated organic moderator to produce cold neutron in a reactor based cold neutron source. This is of interest for two reasons: first, for possible enhancement of the operation of the Texas Cold Neutron Source located at the TRIGA reactor of the Nuclear Engineering Teaching Laboratory at the University of Texas; second, for the design of an Ultracold Neutron Source. Two computer codes were used to achieve this objective: NJOY 97 to generate the necessary thermal scattering cross sections; MCNP 4B to simulate the thermalization and transport of the neutrons in the Texas Cold Neutron Source. The generation of thermal neutron cross sections from neutron scattering laws by NJOY was intensively studied for methane (CH₄) at 22K. The influence of several key physical and numerical parameters of the input file were investigated. The results obtained gave a good understanding on how to generate the cross section in the best capability of the code NJOY. The calculated scattering cross sections were then tested by MCNP to match an experiment performed by Utsuro (1975). Although the experiment was done with a different organic moderator, 1,3,5-trimethylbenzene (mesithylene), the effects associated to the three methyl radicals should be present in the case of methane. The results of this study gave confidence that the cross sections generated by NJOY match rather well the physical cross section of organic materials. Finally, using thermal cross section for methane and deuterated methane generated by NJOY, the performance of the Texas Cold Neutron Source was simulated using MCNP under a variety of conditions. The incident thermal neutron flux coming from the core of the TRIGA reactor at NETL was simulated. Moderation of these neutrons takes place in the TCNS chamber and cold neutrons entering the neutron guide were tallied. The effect of mixing ordinary and deuterated moderators was determined