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Universidad Andrés Bello
1.
Hernandez Meruane, Nicolás Alonso.
Evaluación sísmica y analisis de envejecimiento de la estructura del block de piscina del reactor nuclear Rech-1
.
Degree: 2015, Universidad Andrés Bello
URL: http://repositorio.unab.cl/xmlui/handle/ria/2677 
► El Reactor Experimental Chileno uno. Fue diseñado a finales de la década de los 60’ y su construcción comenzó a principio de los 70’, el…
(more)
▼ El Reactor Experimental Chileno uno. Fue diseñado a finales de la década de los 60’ y su construcción comenzó a principio de los 70’, el diseño estructural del RECH-1, es una copia de los reactores de piscina británicos. Con más de cuarenta años en operación y situado en un país como el nuestro, donde la sismicidad es alta, surge de manera imperante, la necesidad de conocer el estado actual de la estructura, para ello se evalúa su envejecimiento mediante un documento propuesto por la Agencia Internacional de Energía Atómica, además de un análisis sísmico.
La función que desempeña la estructura no permite realizar ensayos invasivos para poder conocer el estado de los hormigones, por lo que el estudio de los mismos se desarrolla mediante modelos que permitan reflejar los fenómenos que los afectan. Siendo lo de mayor importancia el estudio de las grietas que presenta. En análisis sísmico es necesario para contextualizar la estructura y observar cómo se comporta el diseño bajo la normativa nacional.
El modelo de elementos finitos que se realizara para obtener los esfuerzos en los elementos de hormigón armado, incorpora el acople de las masas de agua presentes en la piscina y es sometido a un proceso de calibración para luego proceder a la realización de un análisis modal espectral bajo el espectro de la norma NCH 2369.
Para calibrar el modelo, se procede a comprar los registros de salidas del modelo versus los registros monitoreados en la estructura, se comparan algunos parámetros de los registros, como energía liberada, picos de aceleración contenidos de frecuencias etc. Se recurre a la experiencia de los ingenieros chilenos, como Arturo Arias, quien participo en el diseño del edificio de contención del RECH-1, y Raúl Saragoni quien reviso la estructura luego del terremoto del 27 de febrero del 2010.
Advisors/Committee Members: Vargas Cárdenas, Eugenio (advisor).
Subjects/Keywords: Ingenieria Sismica;
Reactores Nucleares;
Fusión Nuclear
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APA (6th Edition):
Hernandez Meruane, N. A. (2015). Evaluación sísmica y analisis de envejecimiento de la estructura del block de piscina del reactor nuclear Rech-1
. (Thesis). Universidad Andrés Bello. Retrieved from http://repositorio.unab.cl/xmlui/handle/ria/2677
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Chicago Manual of Style (16th Edition):
Hernandez Meruane, Nicolás Alonso. “Evaluación sísmica y analisis de envejecimiento de la estructura del block de piscina del reactor nuclear Rech-1
.” 2015. Thesis, Universidad Andrés Bello. Accessed March 04, 2021.
http://repositorio.unab.cl/xmlui/handle/ria/2677.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
MLA Handbook (7th Edition):
Hernandez Meruane, Nicolás Alonso. “Evaluación sísmica y analisis de envejecimiento de la estructura del block de piscina del reactor nuclear Rech-1
.” 2015. Web. 04 Mar 2021.
Vancouver:
Hernandez Meruane NA. Evaluación sísmica y analisis de envejecimiento de la estructura del block de piscina del reactor nuclear Rech-1
. [Internet] [Thesis]. Universidad Andrés Bello; 2015. [cited 2021 Mar 04].
Available from: http://repositorio.unab.cl/xmlui/handle/ria/2677.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Council of Science Editors:
Hernandez Meruane NA. Evaluación sísmica y analisis de envejecimiento de la estructura del block de piscina del reactor nuclear Rech-1
. [Thesis]. Universidad Andrés Bello; 2015. Available from: http://repositorio.unab.cl/xmlui/handle/ria/2677
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
University of Tennessee – Knoxville
2.
Griswold, Justin Reed.
Thick Target Yield of Th-229 via Low Energy Proton Bombardment of Th-232.
Degree: MS, Nuclear Engineering, 2014, University of Tennessee – Knoxville
URL: https://trace.tennessee.edu/utk_gradthes/2820
► Actinium-225 is one of the more effective radioisotopes used in alpha radioimmunotherapy. Due to its ten-day half-life, it is more efficient to create its…
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▼ Actinium-225 is one of the more effective radioisotopes used in alpha radioimmunotherapy. Due to its ten-day half-life, it is more efficient to create its precursor,
229Th [Thorium-229] (t
1/2[half-life] = 7932 ± 55 years). In this work,
229Th was produced via 40 MeV [Mega electron Volts] proton bombardment of a thick
232Th [Thorium-232] target. The irradiation took place at the Holifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Lab (ORNL). The target, consisting of 23 stacked natural thorium foils (137 mg/cm
2 [milligrams per square centimeter] each), was irradiated with 50 nA [nanoamps] of protons from HRIBF’s 25 MV [Mega Volt] tandem electrostatic accelerator for approximately 143 discontinuous hours. After 215 days post bombardment, allowing for the decay of short-lived protactinium and actinium isotopes and fission products, the target was chemically purified by a series of ion chromatography techniques. Thorium-229 was measured directly by γ-ray [gamma-ray] spectroscopy immediately after separation of the thorium fraction from the decay daughters of
228Th [Thorium-228] (t
1/2 = 1.9 years) and long long-lived fission products. The effective thick target cross section of
229Th is 205 ± 18 mb [millibarns] at a proton energy of 26.1 MeV. Variation of the effective cross-section as a function of proton energy is also reported.
Advisors/Committee Members: Lawrence H. Heilbronn, Howard L. Hall, Laurence F. Miller.
Subjects/Keywords: Nuclear; Nuclear Engineering
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APA (6th Edition):
Griswold, J. R. (2014). Thick Target Yield of Th-229 via Low Energy Proton Bombardment of Th-232. (Thesis). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_gradthes/2820
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Chicago Manual of Style (16th Edition):
Griswold, Justin Reed. “Thick Target Yield of Th-229 via Low Energy Proton Bombardment of Th-232.” 2014. Thesis, University of Tennessee – Knoxville. Accessed March 04, 2021.
https://trace.tennessee.edu/utk_gradthes/2820.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
MLA Handbook (7th Edition):
Griswold, Justin Reed. “Thick Target Yield of Th-229 via Low Energy Proton Bombardment of Th-232.” 2014. Web. 04 Mar 2021.
Vancouver:
Griswold JR. Thick Target Yield of Th-229 via Low Energy Proton Bombardment of Th-232. [Internet] [Thesis]. University of Tennessee – Knoxville; 2014. [cited 2021 Mar 04].
Available from: https://trace.tennessee.edu/utk_gradthes/2820.
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Council of Science Editors:
Griswold JR. Thick Target Yield of Th-229 via Low Energy Proton Bombardment of Th-232. [Thesis]. University of Tennessee – Knoxville; 2014. Available from: https://trace.tennessee.edu/utk_gradthes/2820
Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation
Texas A&M University
3.
Ragusa, Jean Concetto.
Accelerator driven production of tritium: target and blanket design.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R34
► Tritium is an essential component of thermonuclear weapons in the US arsenal. Unfortunately, tritium is a radioactive form of hydrogen, and one-half of the inventory…
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▼ Tritium is an essential component of thermonuclear weapons in the US arsenal. Unfortunately, tritium is a radioactive form of hydrogen, and one-half of the inventory disappears through radioactive decay every 12 years; therefore, it must be replenished. Until a few years ago, the only way to accomplish the tritium production mission was to use fission reactors. Recently, thanks to the development of new accelerator technologies (SDI and SSC studies) and to the post cold war era (international treaties limiting the number of warheads and therefore the tritium requirements), accelerator-based production of tritium seems feasible and is being investigated. The production of tritium using accelerators is a two step process: the production of neutrons in the 'target' and the use of these neutrons in the 'blanket assembly'. The systems described in this thesis employ a linear accelerator (1 GeV protons, I 00 mA beam current), lead targets for the production of neutrons via spallation reactions, and tritium breeding regions (blankets containing '6Li in various mixtures). The high energy interactions and the particle transport were modeled with the LAHET computer code system. Heterogeneous and homogeneous spallation target/blanket systems were investigated. The target designs in the heterogeneous systems were 1 / liquid lead, and 2/ layers of solid lead plates cooled by heavy water. The tritium breeding blanket assemblies contained either lithium oxide or molten fluorine salt with or without UF4' The tritium production rates achieved were-1 5 tritium atoms per incident proton for the L'20 blanket,-1 6 tritium atoms per incident proton for the LiF BeF2ZrF4blanket, and-215 tritium atoms per incident proton for the LiF BeF2ZrF4UF4blanket. An homogeneous target/blanket system consisting of molten lithium lead eutectic (L',7Pb83) was also considered. This design was the most promising with-24 to-29 tritium atoms per incident proton, upgradable to-32 tritium atoms per incident proton.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Ragusa, J. C. (2012). Accelerator driven production of tritium: target and blanket design. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R34
Chicago Manual of Style (16th Edition):
Ragusa, Jean Concetto. “Accelerator driven production of tritium: target and blanket design.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R34.
MLA Handbook (7th Edition):
Ragusa, Jean Concetto. “Accelerator driven production of tritium: target and blanket design.” 2012. Web. 04 Mar 2021.
Vancouver:
Ragusa JC. Accelerator driven production of tritium: target and blanket design. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R34.
Council of Science Editors:
Ragusa JC. Accelerator driven production of tritium: target and blanket design. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R34
Texas A&M University
4.
Ramone, Gilles Lionel.
A Transport Synthetic Acceleration method for transport iterations.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R365
► We present a family of Transport Synthetic Acceleration (TSA) methods to iteratively solve within-group scattering problems. A single iteration in these schemes consists of a…
(more)
▼ We present a family of Transport Synthetic Acceleration (TSA) methods to iteratively solve within-group scattering problems. A single iteration in these schemes consists of a transport sweep followed by a low-order calculation which is itself a simplified transport problem. We describe the development and the realization of the method for an isotropic source in XY geometry. We carry out a Fourier Analysis for a continuous set of equations and report TSA behavior. We show that a previously proposed TSA method is unstable in two dimensions but that our modifications make it stable and rapidly convergent. We follow the same procedure for descritized transport equations, using Step-Characteristics and two Bilinear Discontinuous methods, and find that discretization enhances TSA performance. We then propose to implement a Conjugate Gradient method on the low-order problem, to use a crude quadrature set in the low-order problem and to set the number of low-order iterations per transport sweep to a finite value. We prove that these features represent simple and efficient improvements to the method. We test TSA on a series of physical problems and propose a set of parameters for which the method behaves especially well. We further demonstrate that TSA achieves a substantial reduction in computational cost over Source Iteration, regardless of discretization parameters or materials and emphasize that this gain is an increasing function of the scattering ratio. We devote the final section to some conclusions and suggestions for future work.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Ramone, G. L. (2012). A Transport Synthetic Acceleration method for transport iterations. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R365
Chicago Manual of Style (16th Edition):
Ramone, Gilles Lionel. “A Transport Synthetic Acceleration method for transport iterations.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R365.
MLA Handbook (7th Edition):
Ramone, Gilles Lionel. “A Transport Synthetic Acceleration method for transport iterations.” 2012. Web. 04 Mar 2021.
Vancouver:
Ramone GL. A Transport Synthetic Acceleration method for transport iterations. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R365.
Council of Science Editors:
Ramone GL. A Transport Synthetic Acceleration method for transport iterations. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R365
Texas A&M University
5.
Reyes, Joseph Patrick.
Simulation of condensation in the presence of noncondensible gases using the French Thermal Hydraulic computer code CATHARE.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R49
► The modeling of the complex thermal hydraulics of reactor systems involves the use of experimental test systems as well as numerical codes. A simulation of…
(more)
▼ The modeling of the complex thermal hydraulics of reactor systems involves the use of experimental test systems as well as numerical codes. A simulation of the phenomena of condensation in the presence of noncondensible gas was performed using the CATHARE thermal hydraulic code. The experiments were conducted at the University of California, Berkeley, and at the Energy Research Laboratory in Japan; supervised by Hitachi, Ltd.. The experiments involved the condensation of a steam/air mixture in downward flow through pipes. The test sections approximate the design of isolation condensers included in the passive containment cooling system in the new generation of nuclear reactors. Both the University of California at Berkeley and Hitachi test sections consisted of a stainless steel pipe connected to an upper and lower plenum. An annulus containing cooling water was present in both experiments. The steam/air mixture was introduced into the test section by means of mixers connected to the upper plenum. The transient runs were conducted on two separate platforms. The University of California at Berkeley simulation was conducted on a HP 9000/720 Workstation at Texas A&M University. The Hitachi simulation was performed on the "ventoux" branch of the HP network located at the Centre D'Etudes Nucleaires de Grenoble, located in Grenoble, France. The position in which condensation began to occur was overpredicted in the Berkeley simulation, but was accurate in the Hitachi simulation. The liquid heat flux was heavily underpredicted in both the Berkeley and Hitachi simulation. As a result of the underprediction in the heat flux profile, the condensing tube wall temperature and the local overall temperature coefficients were underpredicted in comparison to the experimental data. Secondary side cooling water temperature was also underpredicted as a result of the underprediction of the primary side heat flux. Overall, the CATHARE simulation were in reasonable agreement with both the low and high pressure cases.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Reyes, J. P. (2012). Simulation of condensation in the presence of noncondensible gases using the French Thermal Hydraulic computer code CATHARE. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R49
Chicago Manual of Style (16th Edition):
Reyes, Joseph Patrick. “Simulation of condensation in the presence of noncondensible gases using the French Thermal Hydraulic computer code CATHARE.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R49.
MLA Handbook (7th Edition):
Reyes, Joseph Patrick. “Simulation of condensation in the presence of noncondensible gases using the French Thermal Hydraulic computer code CATHARE.” 2012. Web. 04 Mar 2021.
Vancouver:
Reyes JP. Simulation of condensation in the presence of noncondensible gases using the French Thermal Hydraulic computer code CATHARE. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R49.
Council of Science Editors:
Reyes JP. Simulation of condensation in the presence of noncondensible gases using the French Thermal Hydraulic computer code CATHARE. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-R49
Texas A&M University
6.
Troshko, Andrey Arthurovich.
Simulation of the loss of the residual heat removal of an integral test facility using computer code Cathare7.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-T76
► Assessment of the ability of thermal hydraulic codes to correctly predict complex phenomena during reactor transients can be performed only by means of comparison against…
(more)
▼ Assessment of the ability of thermal hydraulic codes to correctly predict complex phenomena during reactor transients can be performed only by means of comparison against experimental data. The thermal hydraulic CATHARE V1.3U computer program has been used to simulate International Standard Problem (ISP38) experiment conducted at BETHSY Integral Test Facility located in Grenoble, France. BETHSY is an integral test facility modeling the three-loop, 2775 thermal NM Framatome Pressurized Water Reactor (PWR). Its main objectives are to contribute to the assessment of CATHARE code and of the physical basis of PWR Emergency Operating Procedures. This investigation dealt with experimental simulation of the loss of the Residual Heat Removal System (RHRS) during midloop operation. It involved opening of the pressurizer manway and the steam generator outlet plenum manway simultaneously with switching on the power of heating rods, thus simulating loss of the RHRS. The total power was kept unchanged with the level 138 kW throughout the test. Mass discharge through both manways led to core boiling and uncovery. A gravity feed injection of cold water was actuated in the cold leg when cladding temperature at the top of heating rods reached 250 C. The test was stopped when the primary cooling system was filled back to rnidloop level. The transient was executed on the HP6400 workstation at Texas A&M University. The code predictions underestimated the time of the core uncovery and the actuation of the gravity feed injection due to the ovepredicted mass discharge through steain generator manway during the initial stage of the transient. This was probably caused by CATHARE's miscalculation of the phase separation effect at the hot leg/surge line tee junction and significant water entrainment into the surge line in the beginning of the test. It was found that the model of the upward tee junction needs to be refined for the low pressure range. Overall, the code's predictions were in a qualitative agreement with the experimental data.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Troshko, A. A. (2012). Simulation of the loss of the residual heat removal of an integral test facility using computer code Cathare7. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-T76
Chicago Manual of Style (16th Edition):
Troshko, Andrey Arthurovich. “Simulation of the loss of the residual heat removal of an integral test facility using computer code Cathare7.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-T76.
MLA Handbook (7th Edition):
Troshko, Andrey Arthurovich. “Simulation of the loss of the residual heat removal of an integral test facility using computer code Cathare7.” 2012. Web. 04 Mar 2021.
Vancouver:
Troshko AA. Simulation of the loss of the residual heat removal of an integral test facility using computer code Cathare7. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-T76.
Council of Science Editors:
Troshko AA. Simulation of the loss of the residual heat removal of an integral test facility using computer code Cathare7. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-T76
Texas A&M University
7.
Walkowicz, Joshua Peter.
Investigation of a thermoluminescent dosimeter mixture between LiF:Mg,Ti and Li2B4O7 in a solid form.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-W355
► Thermoluminescent dosimetry is one of the most popular methods in use today for the monitoring of personnel exposures to radiation. This research investigates a different…
(more)
▼ Thermoluminescent dosimetry is one of the most popular methods in use today for the monitoring of personnel exposures to radiation. This research investigates a different type of dosimeter, a 3:1 mixture of lithium fluoride (TLD-I 00) and lithium borate (TLD800) pressed into tablet form. This mixture was previously combined within PTFE tubing which limited the capabilities of the dosimeter. These limitations led to this investigation. Several methods of making the dosimeters were studied, including sintering and the addition of a binder. The dosimeter chips were exposed to both beta and gamma radiation with a slightly higher response in the beta measurements, about 19% higher. The response for the beta and gamma exposures were 6.012 ︢ 2.032 and 5.246 ︢1.648 RC for a dose of' I Gy, respectively. The main set of chips also were exposed to several dose levels to check for linearity. The best fit line for this data had a slope of 7.147 for thermoluminescent output versus dose. Based on this data the minimal detectable dose was calculated to be 0.58 mGy, or three times the standard deviation of the background. A fading study revealed a percent loss of 65.6% in 7 days for one set of dosimeters, and 18.4% in 26 days in another. The dosimeters were then checked for light sensitivity, but there was none observed.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Walkowicz, J. P. (2012). Investigation of a thermoluminescent dosimeter mixture between LiF:Mg,Ti and Li2B4O7 in a solid form. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-W355
Chicago Manual of Style (16th Edition):
Walkowicz, Joshua Peter. “Investigation of a thermoluminescent dosimeter mixture between LiF:Mg,Ti and Li2B4O7 in a solid form.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-W355.
MLA Handbook (7th Edition):
Walkowicz, Joshua Peter. “Investigation of a thermoluminescent dosimeter mixture between LiF:Mg,Ti and Li2B4O7 in a solid form.” 2012. Web. 04 Mar 2021.
Vancouver:
Walkowicz JP. Investigation of a thermoluminescent dosimeter mixture between LiF:Mg,Ti and Li2B4O7 in a solid form. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-W355.
Council of Science Editors:
Walkowicz JP. Investigation of a thermoluminescent dosimeter mixture between LiF:Mg,Ti and Li2B4O7 in a solid form. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-W355
Texas A&M University
8.
Dixon, John Leslie.
Electrostatic coalescence of used automotive crankcase oil as an alternative to other separation processes.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-D59
► This thesis presents an initial investigation of using electrostatic coalescence as an alternative to conventional separation processes to purify used automotive crankcase oil. Specific emphasis…
(more)
▼ This thesis presents an initial investigation of using electrostatic coalescence as an alternative to conventional separation processes to purify used automotive crankcase oil. Specific emphasis of this study was the feasibility of this approach, verified by separating and analyzing a used oil emulsion. The metal removal efficiency was compared to that of a five day gravity settling. Separation experiments were performed in a 2.26 L coalescer with a flat parallel insulated electrode configuration. The used oil emulsion, composed of used oil, Isopar M, and water (no noticeable phase separation for 12 hours) followed the electrostatic coalescence characteristic of higher applied voltages or frequencies allowing higher feed rates. Metal removal efficiencies for iron, calcium and zinc were 3.57, 47.1, and 46.7 %, respectively, using Nalco 7715 at a peak a.c. voltage of 7 kV/cm and a frequency of 1000 Hz at the maximum rate of coalescence. For gravity settlement, metal removal efficiencies for iron, calcium and zinc were 11.2, 15.6, and 57.1 %, respectively. Considering the residence time of a moderate emulsion feed rate is a fraction of an hour, electrostatic coalescence offers an advantage over gravity settling. Oil phase water content varied between 0.05 and 7.2 wt %.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Dixon, J. L. (2012). Electrostatic coalescence of used automotive crankcase oil as an alternative to other separation processes. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-D59
Chicago Manual of Style (16th Edition):
Dixon, John Leslie. “Electrostatic coalescence of used automotive crankcase oil as an alternative to other separation processes.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-D59.
MLA Handbook (7th Edition):
Dixon, John Leslie. “Electrostatic coalescence of used automotive crankcase oil as an alternative to other separation processes.” 2012. Web. 04 Mar 2021.
Vancouver:
Dixon JL. Electrostatic coalescence of used automotive crankcase oil as an alternative to other separation processes. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-D59.
Council of Science Editors:
Dixon JL. Electrostatic coalescence of used automotive crankcase oil as an alternative to other separation processes. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-D59
Texas A&M University
9.
Hamm, Trenton Allen.
A steady state analysis code for prediction of behavior in loop heat pipes.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H354
► The purpose of this work is to prepare an analysis raphics. code for the prediction of Loop Heat Pipe (LHP) behavior in steady-state operation. The…
(more)
▼ The purpose of this work is to prepare an analysis raphics. code for the prediction of Loop Heat Pipe (LHP) behavior in steady-state operation. The FORTRAN program is then benchmarked with experimental data obtained in two orientations: 1) the evaporator-over- condenser and 2) the condenser-over-evaporator configurations. The results are quite promising in that the behavior of the LHP was well predicted given certain controllable assumptions. These results also show a strong dependence on the ambient and coolant temperatures and whether the LHP is insulated from the environment or not. The fact that these three items are easily controlled validates the modeling techniques employed here. It is interesting to note that the knowledge of the internal characteristics of the LHP wick is not absolutely necessary to accurately predict behavior as well.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Hamm, T. A. (2012). A steady state analysis code for prediction of behavior in loop heat pipes. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H354
Chicago Manual of Style (16th Edition):
Hamm, Trenton Allen. “A steady state analysis code for prediction of behavior in loop heat pipes.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H354.
MLA Handbook (7th Edition):
Hamm, Trenton Allen. “A steady state analysis code for prediction of behavior in loop heat pipes.” 2012. Web. 04 Mar 2021.
Vancouver:
Hamm TA. A steady state analysis code for prediction of behavior in loop heat pipes. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H354.
Council of Science Editors:
Hamm TA. A steady state analysis code for prediction of behavior in loop heat pipes. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H354
Texas A&M University
10.
Hari, Sridhar.
Analysis of a research reactor under anticipated transients without scram events using the RELAP5/MOD3.2 computer program.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H373
► Simulations for two series of anticipated transients phics. without scram (ATWS) events have been carried out for a small, hypothetical, research reactor based on the…
(more)
▼ Simulations for two series of anticipated transients phics. without scram (ATWS) events have been carried out for a small, hypothetical, research reactor based on the High Flux Australian Reador HIFAR using the RELAPS/MOD3.Z computer program. The first series simulated the ATWS events associated with pump transients and the second series simulated the ATWS events associated with heat exchanger transients. In the first series, two cases were simulated: (1) the non-availability of one of the two primary pumps and (2) the non-availability of one pump and 50 % operation of the other pimp. In the second series, three cases were simulated: (1) the non-availability of one of the three heat enchanters (2) the non-availability of two of the three heat enchanters and (3) the non-availability of all the three heat enchanters. To carry out the above simulations, modifications were incorporated to the critical heat flux (CHF) correlations of the RELAP program to account for the annular geometry of the fuel elements. The results of the simulated plump transients indicated that the mass flow rate would decrease to 270 Kg/s in the first transient and to 133 Kg/s in transient 2. The maximum rise in the coolant temperatures in the primly circuit would be 50C and the maximum rise in the central fuel element temperature would be 23[]C. The results of the first two cases of the heat exchanger transients indicated that the maximum rise in the coolant temperatures in the primary circuit would be 18.5[]C and the maximum rise in the central fuel element temperature would be 6[]C. However, during the simulation of the non-availability of all the three heat enchanters, primary coolant temperature started increasing rapidly. Void generation occurred in the central fuel element annuli at 408.5 seconds into the transient with saturation conditions being attained in the annulus. The phase change resulted in the expulsion of fluid from the annulus. Consequently, temperature in the second fuel ring exceeded 800K, resulting in its failure.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Hari, S. (2012). Analysis of a research reactor under anticipated transients without scram events using the RELAP5/MOD3.2 computer program. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H373
Chicago Manual of Style (16th Edition):
Hari, Sridhar. “Analysis of a research reactor under anticipated transients without scram events using the RELAP5/MOD3.2 computer program.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H373.
MLA Handbook (7th Edition):
Hari, Sridhar. “Analysis of a research reactor under anticipated transients without scram events using the RELAP5/MOD3.2 computer program.” 2012. Web. 04 Mar 2021.
Vancouver:
Hari S. Analysis of a research reactor under anticipated transients without scram events using the RELAP5/MOD3.2 computer program. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H373.
Council of Science Editors:
Hari S. Analysis of a research reactor under anticipated transients without scram events using the RELAP5/MOD3.2 computer program. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1998-THESIS-H373
Texas A&M University
11.
Dorsey, Daniel John.
Electrical resistance of gases in explosive magnetic flux compression generator environments.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-D68
► Gases that are electrical insulators at STP can become conductors when subjected to the high pressure and temperature environment in explosive magnetic flux compression generators…
(more)
▼ Gases that are electrical insulators at STP can become conductors when subjected to the high pressure and temperature environment in explosive magnetic flux compression generators (FCGs). This thesis describes experiments performed to determine the electrical properties of several gases in the FCG environment. The hydrodynamics in a helical explosive magnetic flux compression generator (FCG) are modeled using the Gurney method and a shock physics code, CTH, developed at Sandia National Laboratory. The armature in a typical FCG is calculated to approach the stator at approximately 3 km/s. To simulate FCG operating conditions in the volume between the armature and stator, expendable, stagnated shock, explosively driven shock tubes are designed to propel aluminum flyer plates towards dense stainless steel plugs. Two opposing copper probes are inserted into the shock tube walls and charged to 2 kV by an external capacitor bank. The voltage across these probes is tracked by oscilloscope and the current is measured with a Pearson transformer at the capacitor bank. The current and voltage measurements are used to calculate a bulk resistance for the gas between the probes. Current limited experiments with series resistors have also been conducted. Experiments were performed using argon, helium, sulfur hexafluoride, and synthetic air (20% oxygen/80% nitrogen) as the shock tube fill gas. Argon readily ionizes throughout its volume in the shock tube. This effect is believed to be due to photo-ionization from radiation emitted in the shocked region. Helium only becomes ionized when the initial shock wave reflects off the stainless steel plug and stagnates. Air ionizes in the shock wave and maintains a resistance near 3 Ohms when the initially shocked region is measured. SF6 is ionized and becomes less resistive as the shocked region crosses the probes. The SF6 resistance measurement is 200 Ohms for the initial shock wave. In all cases, the reflected shock wave is expected to have sufficient energy to begin disassembling the shock tube, and calculations based on measurements made after the reflected shock wave reaches the voltage probes are unreliable.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Dorsey, D. J. (2012). Electrical resistance of gases in explosive magnetic flux compression generator environments. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-D68
Chicago Manual of Style (16th Edition):
Dorsey, Daniel John. “Electrical resistance of gases in explosive magnetic flux compression generator environments.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-D68.
MLA Handbook (7th Edition):
Dorsey, Daniel John. “Electrical resistance of gases in explosive magnetic flux compression generator environments.” 2012. Web. 04 Mar 2021.
Vancouver:
Dorsey DJ. Electrical resistance of gases in explosive magnetic flux compression generator environments. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-D68.
Council of Science Editors:
Dorsey DJ. Electrical resistance of gases in explosive magnetic flux compression generator environments. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-D68
Texas A&M University
12.
Fu, Chun.
RELAP5/MOD3.2 analysis of a VVER-1000 reactor with UO[2] fuel and MOX fuel.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-F8
► A RELAP5/MOD3.2 model of a VVER-1000/MODEL V320 nuclear power plant, Balakovo Unit 4, was updated, improved and validated on the basis of an input deck…
(more)
▼ A RELAP5/MOD3.2 model of a VVER-1000/MODEL V320 nuclear power plant, Balakovo Unit 4, was updated, improved and validated on the basis of an input deck prepared by the Kurchatov Institute of Moscow. The RELAP5 model includes both the primary and the secondary systems. The Emergency Core Cooling System (ECCS) is modeled according to the plant configuration. The feedwater system, along with the emergency feedwater system, is included in the model. The point reactor kinetics model, in which the decay heat is calculated with ANS decay heat data, enables the model to be used for analysis of a large spectrum of transients and accidents. The plant model is used for analysis and prediction of a cold leg Large Break Loss-of-Coolant Accident (LBLOCA). The RELAP5/MOD3.2 results showed a good agreement with calculations obtained with TECH-M computer program. The cladding temperatures of the MOX assembly have been compared with that of the hot UO₂ assembly. The peak cladding temperature of MOX assembly is about 55 K higher than that of UO₂ assembly. An uncertainty analysis has been performed for the peak cladding temperature, in which Monte Carlo calculations have been performed using the response surface built up from fifteen sets of RELAP5 calculations. The result shows that the ECCS would be sufficient to keep the cladding temperature during the scenario of a LBLOCA well below the required licensing limit.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Fu, C. (2012). RELAP5/MOD3.2 analysis of a VVER-1000 reactor with UO[2] fuel and MOX fuel. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-F8
Chicago Manual of Style (16th Edition):
Fu, Chun. “RELAP5/MOD3.2 analysis of a VVER-1000 reactor with UO[2] fuel and MOX fuel.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-F8.
MLA Handbook (7th Edition):
Fu, Chun. “RELAP5/MOD3.2 analysis of a VVER-1000 reactor with UO[2] fuel and MOX fuel.” 2012. Web. 04 Mar 2021.
Vancouver:
Fu C. RELAP5/MOD3.2 analysis of a VVER-1000 reactor with UO[2] fuel and MOX fuel. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-F8.
Council of Science Editors:
Fu C. RELAP5/MOD3.2 analysis of a VVER-1000 reactor with UO[2] fuel and MOX fuel. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-F8
Texas A&M University
13.
Carlisle, Bruce Scott.
An evaluation of the neutron radiography facility at the Nuclear Science Center for dynamic imaging of two-phase hydrogenous fluids.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C283
► Though both film and video radiographic image techniques are available in neutron radiography, radiographic cameras are commonly used to capture the dynamic flow patterns in…
(more)
▼ Though both film and video radiographic image techniques are available in neutron radiography, radiographic cameras are commonly used to capture the dynamic flow patterns in a rapid sequence of images. These images may be useful to verify two-phase flow models in small diameter flow channels. An initial series of real-time neutron radiography experiments were performed at the Texas A&M University System, Texas Engineering Experiment Station, Nuclear Science Center Reactor (NSCR) to determined the image resolution of two-phase water and air flow regimes through small diameter metal flow channels. After evaluating these initial images, research was conducted to determine cost effective enhancements that would increase the dimensional accuracy and contrast of these flow images. Modifications were completed to the beam collimator and the radiography camera video processing board was realigned to provide a stronger vidio signal with less noise. Several hydrogenous-media reference standards were designed and constructed to evaluate the effectiveness of the modifications. The beamport collimator was redesigned and the radiography calibration methodology was changed. The post-modification images demonstrate that a smaller, more focused neutron beam and a more sensitive video camera provide clearer images with excellent dimensional characteristics. Specific research to quantify both the resolution and sensitivity limits is proposed and a change in dynamic target imaging methodology is proposed.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Carlisle, B. S. (2012). An evaluation of the neutron radiography facility at the Nuclear Science Center for dynamic imaging of two-phase hydrogenous fluids. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C283
Chicago Manual of Style (16th Edition):
Carlisle, Bruce Scott. “An evaluation of the neutron radiography facility at the Nuclear Science Center for dynamic imaging of two-phase hydrogenous fluids.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C283.
MLA Handbook (7th Edition):
Carlisle, Bruce Scott. “An evaluation of the neutron radiography facility at the Nuclear Science Center for dynamic imaging of two-phase hydrogenous fluids.” 2012. Web. 04 Mar 2021.
Vancouver:
Carlisle BS. An evaluation of the neutron radiography facility at the Nuclear Science Center for dynamic imaging of two-phase hydrogenous fluids. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C283.
Council of Science Editors:
Carlisle BS. An evaluation of the neutron radiography facility at the Nuclear Science Center for dynamic imaging of two-phase hydrogenous fluids. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C283
Texas A&M University
14.
Castrianni, Christopher Lee.
A one-dimensional nonlinear corner balance method for solving the neutron transport equation.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C355
► The one-dimensional nonlinear comer balance method is a new spatial discretization scheme for solving the transport equation on grids consisting of arbitrarily connected polygonal meshes.…
(more)
▼ The one-dimensional nonlinear comer balance method is a new spatial discretization scheme for solving the transport equation on grids consisting of arbitrarily connected polygonal meshes. It is a conceptually simple method based on particle balance over half-cells. We describe the development and realization of this method in slab geometry with isotropic scattering. We prove that it is a positive method given positive sources. We show that the method is exact in purelyabsorbing and source-free half-space problems. We then compare the method to the linear comer balance and the nonlinear characteristics methods in a source-free deep-penetration problem. We then discuss the effects of downwardly-concave sources on the interior solution. We analyze the method in the thick, diffusive limit and show that the nonlinear comer balance method satisfies a discrete diffusion equation with Marshak boundary conditions. We further demonstrate that this discrete diffusion equation becomes the standard three-point finite difference approximation in the fine mesh limit. We demonstrate the lack of importance of the loss of superposition and Marshak boundary conditions by solving several simple transport problems. Finally, we show that the method displays fourth order truncation error through the use of a simple numerical experiment and offer some conclusions and suggestions for future work
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Castrianni, C. L. (2012). A one-dimensional nonlinear corner balance method for solving the neutron transport equation. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C355
Chicago Manual of Style (16th Edition):
Castrianni, Christopher Lee. “A one-dimensional nonlinear corner balance method for solving the neutron transport equation.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C355.
MLA Handbook (7th Edition):
Castrianni, Christopher Lee. “A one-dimensional nonlinear corner balance method for solving the neutron transport equation.” 2012. Web. 04 Mar 2021.
Vancouver:
Castrianni CL. A one-dimensional nonlinear corner balance method for solving the neutron transport equation. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C355.
Council of Science Editors:
Castrianni CL. A one-dimensional nonlinear corner balance method for solving the neutron transport equation. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C355
Texas A&M University
15.
Cebull, Peter Patrick.
Simulation of the SPE-4 small-break loss-of-coolant accident using RELAP5/MOD 3.1.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C3875
► A small-break loss-of-coolant accident experiment conducted at the PMK-2 integral test facility in Hungary is analyzed using the RELAP5/MOD3.1 thermal-hydrauhc code. The experiment simulated a…
(more)
▼ A small-break loss-of-coolant accident experiment conducted at the PMK-2 integral test facility in Hungary is analyzed using the RELAP5/MOD3.1 thermal-hydrauhc code. The experiment simulated a 7.4% break in the cold leg of a VVER-440/213-type nuclear power plant as part of the International Atomic Energy Agency's Fourth Standard Problem Exercise (SPE-4). One distinguishing characteristic of this type of power plant 'LS the horizontal steam generator. Nineteen countries participated in the exercise, with Texas A&M representing the U. S. Blind calculations of the exercise are presented, and the timing of various events throughout the transient is discussed. A post-test analysis is performed in which the sensitivity of the calculated results is investigated. RELAP5 predicts most of the transient events well, although the predicted time of occurrence of several events during the accident scenario is adversely affected by an underprediction of system pressure. A few problems are noted, particularly the failure of RELAP5 to predict dryout in the core even though collapsed liquid level fell below the top of the heated portion. A discrepancy between the predicted primary mass inventory distribution and the experimental data is identified. Finally, the primary and secondary pressures calculated by RELAP5 fell too rapidly during the latter part of the transient, resulting in rather large errors in the predicted timing of some pressure-actuated events.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Cebull, P. P. (2012). Simulation of the SPE-4 small-break loss-of-coolant accident using RELAP5/MOD 3.1. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C3875
Chicago Manual of Style (16th Edition):
Cebull, Peter Patrick. “Simulation of the SPE-4 small-break loss-of-coolant accident using RELAP5/MOD 3.1.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C3875.
MLA Handbook (7th Edition):
Cebull, Peter Patrick. “Simulation of the SPE-4 small-break loss-of-coolant accident using RELAP5/MOD 3.1.” 2012. Web. 04 Mar 2021.
Vancouver:
Cebull PP. Simulation of the SPE-4 small-break loss-of-coolant accident using RELAP5/MOD 3.1. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C3875.
Council of Science Editors:
Cebull PP. Simulation of the SPE-4 small-break loss-of-coolant accident using RELAP5/MOD 3.1. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C3875
Texas A&M University
16.
Costes, Sylvain Vincent.
Development of a three-dimensional particle image velocimetry algorithm and analysis of synthetic and experimental flows in three-dimensions.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C842
► This study concerns with the development of a simple three-dimensional technique to determine the velocity of fluid by tracing the motion of seeded particles in…
(more)
▼ This study concerns with the development of a simple three-dimensional technique to determine the velocity of fluid by tracing the motion of seeded particles in a flow in three-dimensions. A correction for light refraction at water-air boundary plane is applied. This technique is an extension of the two-dimensional Pulsed Laser Velocimetry (PLV) to three-dimensions. With this new scheme, one will be able to analyze non-planar and turbulent flows. To validate the scheme, firstly the algorithm will be tested with synthetic data, and secondly a three-dimensional experimental study will be performed. Calibration is necessary in order to calculate the camera parameters that are needed in the algorithm. Which with the followed by a complete analysis of data obtained from a three-dimensional experiment. Vector velocity field, utilizing a three-dimensional cross correlation technique, will be mapped. This method has been proposed in order to provide an easier experiment and calibration than the methods used previously. Indeed, it is a combination of two different techniques. The Yamainoto's scheme is a complex experimental set up with three cameras located on a sphere but which requires easy equations for the calibration. But the Nishino's technique has a simpler set up, with no requirement for the cameras positions, but it involves complex equations concerning the determination of cameras parameters To summarize, this study demonstrates the set up, data acquisition, calibration, coordinate transforms, and tracking of experimental data in three-dimensions.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Costes, S. V. (2012). Development of a three-dimensional particle image velocimetry algorithm and analysis of synthetic and experimental flows in three-dimensions. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C842
Chicago Manual of Style (16th Edition):
Costes, Sylvain Vincent. “Development of a three-dimensional particle image velocimetry algorithm and analysis of synthetic and experimental flows in three-dimensions.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C842.
MLA Handbook (7th Edition):
Costes, Sylvain Vincent. “Development of a three-dimensional particle image velocimetry algorithm and analysis of synthetic and experimental flows in three-dimensions.” 2012. Web. 04 Mar 2021.
Vancouver:
Costes SV. Development of a three-dimensional particle image velocimetry algorithm and analysis of synthetic and experimental flows in three-dimensions. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C842.
Council of Science Editors:
Costes SV. Development of a three-dimensional particle image velocimetry algorithm and analysis of synthetic and experimental flows in three-dimensions. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-C842
Texas A&M University
17.
Eaton, Thomas Lance.
A generalized simple corner-balance transport method for 1-D problems.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-E145
► A new generalized simple Comer Balance transport method was developed in an attempt to produce accurate solutions to all one-dimensional transport problems. The method allows…
(more)
▼ A new generalized simple Comer Balance transport method was developed in an attempt to produce accurate solutions to all one-dimensional transport problems. The method allows only one parameter, 0, to vary according to the values of the scattering ratio and the cell optical thickness. The new method was compared too commonly used methods including Diamond Differencing, standard Linear Discontinuous. and modified Linear Discontinuous. These methods were compared to the solutions resulting! from the new method for various problems. In most cases, the new method provided greater accuracv. Some problems with thick cells and moderate scattering ratios did. however, cause the new method to produce negative solutions in the presence of positive sources. The method provided extremely good accuracy in several regions: pureiv absorbing, highly scattering, and thin cells.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Eaton, T. L. (2012). A generalized simple corner-balance transport method for 1-D problems. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-E145
Chicago Manual of Style (16th Edition):
Eaton, Thomas Lance. “A generalized simple corner-balance transport method for 1-D problems.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-E145.
MLA Handbook (7th Edition):
Eaton, Thomas Lance. “A generalized simple corner-balance transport method for 1-D problems.” 2012. Web. 04 Mar 2021.
Vancouver:
Eaton TL. A generalized simple corner-balance transport method for 1-D problems. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-E145.
Council of Science Editors:
Eaton TL. A generalized simple corner-balance transport method for 1-D problems. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1994-THESIS-E145
Texas A&M University
18.
Yoon, Churl.
Development of a genetic algorithm tracking technique for the particle image velocimetry and comparison with other tracking models.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-Y66
► The Particle Image Velocimetry (plV) flow measu rement technique is a very efficient technique for studying the structure of various fluid flows. It provides quantitative…
(more)
▼ The Particle Image Velocimetry (plV) flow measu rement technique is a very efficient technique for studying the structure of various fluid flows. It provides quantitative and qualitative full-field velocity information which can be analyzed to find the flow's velocity field, vorticity components, turbulent intensities, etc. The measurement technique relies on fast and efficient methods to accurately track images of neutral density particles which have been suspended in the flow. Since a large quantity of data has to be analyzed, the. tracking process must be fast and reliable. A new tracking method, known as the Genetic Algorithm, was developed for the two-dimensional PIV technique. This method utilizes a combinatorial approach to identify corresponding particles between sequential frames. The new tracking method was compared with three other existing tracking methods to evaluate their efficiency, i.e., reliability and yield. Synthetic data from a Large Eddy Simulation (LES) computational fluid dynamic code (GUST) were generated. All four methods were used to track the synthetic data. Then the simulated which are determine vectors were compared with the reconstructed vectors, the tracking, results of four tracking techniques, to the yield and reliability.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Yoon, C. (2012). Development of a genetic algorithm tracking technique for the particle image velocimetry and comparison with other tracking models. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-Y66
Chicago Manual of Style (16th Edition):
Yoon, Churl. “Development of a genetic algorithm tracking technique for the particle image velocimetry and comparison with other tracking models.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-Y66.
MLA Handbook (7th Edition):
Yoon, Churl. “Development of a genetic algorithm tracking technique for the particle image velocimetry and comparison with other tracking models.” 2012. Web. 04 Mar 2021.
Vancouver:
Yoon C. Development of a genetic algorithm tracking technique for the particle image velocimetry and comparison with other tracking models. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-Y66.
Council of Science Editors:
Yoon C. Development of a genetic algorithm tracking technique for the particle image velocimetry and comparison with other tracking models. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1996-THESIS-Y66
Texas A&M University
19.
Caldwell, Amy Baker.
Risk analysis of shipping plutonium pits and mixed oxide fuel.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C35
► With the end of the cold war, there no longer seems to be a credible threat of war between nuclear superpowers, with its possible consequence…
(more)
▼ With the end of the cold war, there no longer seems to be a credible threat of war between nuclear superpowers, with its possible consequence of billions of fatalities. However, the residue of the cold war, most notably the now excess weapons plutonium, has been identified as the source of a number of potential catastrophes. For example, just a single crude nuclear weapon in the hands of a terrorist organization or rogue state and detonated in even a medium-sized city could lead to hundreds of thousands of deaths. For this reason, the ultimate disposition of this excess plutonium has been identified as a national priority. The process of carrying out this disposition itself carries some risks, and even though any conceivable consequences clearly will be much smaller in magnitude than those cited above, U.S. federal law (the National Environmental Protection Act) mandates that such risks must be analyzed. The ability to carry out one type of such an analysis is demonstrated in this thesis. Specifically, one possible option that has been identified for disposition of excess U.S. weapons plutonium is the transformation into mixed oxide (MOX) fuel, that then would be used as fuel in a commercial nuclear power plant. Any such process will involve the transportation of the MOX fuel from the MOX fuel fabrication facility to the nuclear power plant, and possibly transportation of the plutonium from a storage site to the fuel fabrication facility. This thesis is intended to demonstrate the capability to analyze the risks associated with such transportation campaigns. The primary tool used for these analyses was RADTRAN, a code developed by Sandia National Laboratories for evaluating risk associated with the transportation of radioactive materials. Two sample scenarios were explored relative to the transformation of plutonium pits to MOX fuel. First, the pits would be converted to MOX fuel at a fuel fabrication facility located either at the Pantex Plant or the Savannah River Site (SRS), and then the MOX fuel would be ultimately shipped to a final destination of a commercial power plant, the Palo Verde Generating Station in Arizona. For the scenario of placing the MOX fuel fabrication facility at SRS, pits would need to be shipped from Pantex to SRS and then the MOX fuel would be shipped to Palo Verde. The total number of expected fatalities over a 25 year campaign duration for this scenario would be 1.06, with 0. 1 73 fatalities resulting from latent cancer fatalities due to radiation exposure and 0.89 resulting from traffic accidents. For the placement of the MOX fuel fabrication facility at Pantex, only the MOX fuel would need to be transported from one facility to another, in this case from Pantex to Palo Verde. The total fatalities for this scenario over 25 years would be 0.413, resulting from 5.29 x 10-2 latent cancer fatalities and 0.36 traffic accident fatalities. The maximum exposed individual along any of the three routes would receive 1.0 X 10-5 rem per year or 0.25 mrem over 25 years.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Caldwell, A. B. (2012). Risk analysis of shipping plutonium pits and mixed oxide fuel. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C35
Chicago Manual of Style (16th Edition):
Caldwell, Amy Baker. “Risk analysis of shipping plutonium pits and mixed oxide fuel.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C35.
MLA Handbook (7th Edition):
Caldwell, Amy Baker. “Risk analysis of shipping plutonium pits and mixed oxide fuel.” 2012. Web. 04 Mar 2021.
Vancouver:
Caldwell AB. Risk analysis of shipping plutonium pits and mixed oxide fuel. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C35.
Council of Science Editors:
Caldwell AB. Risk analysis of shipping plutonium pits and mixed oxide fuel. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C35
Texas A&M University
20.
Chang, Jae Ho.
Statistical comparison of two-phase flow, void fraction fluctuations in a microgravity environment.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4423
► A two-phase flow experiment was flown aboard the NASA KC-135 zero gravity aircraft to test void fraction sensors and collect void fraction data under the…
(more)
▼ A two-phase flow experiment was flown aboard the NASA KC-135 zero gravity aircraft to test void fraction sensors and collect void fraction data under the unique conditions of microgravity. Void fraction measurements were made by two capacitance void fraction sensors and trapped liquid between two quick closing valves. A statistical method involving a probability density function and moments of a distribution were developed to analyze the void fraction fluctuations for uses as a possible flow regime identifier. Results show that slug flows exhibit both unimodal distribution and multi-modal distribution in the probability density function while annular flows have unimodal distribution with a peak at high void fractions. It was found that the variance of void fluctuations for slug flows tended to be larger than annular flows. Annular flows have negative coefficients of skewness and kurtosis. Slug flows exhibited both positive and negative coefficients of skewness and kurtosis. The combination of probability density function and variance of the void fraction fluctuations was found to be the best flow regime identification tool.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Chang, J. H. (2012). Statistical comparison of two-phase flow, void fraction fluctuations in a microgravity environment. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4423
Chicago Manual of Style (16th Edition):
Chang, Jae Ho. “Statistical comparison of two-phase flow, void fraction fluctuations in a microgravity environment.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4423.
MLA Handbook (7th Edition):
Chang, Jae Ho. “Statistical comparison of two-phase flow, void fraction fluctuations in a microgravity environment.” 2012. Web. 04 Mar 2021.
Vancouver:
Chang JH. Statistical comparison of two-phase flow, void fraction fluctuations in a microgravity environment. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4423.
Council of Science Editors:
Chang JH. Statistical comparison of two-phase flow, void fraction fluctuations in a microgravity environment. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4423
Texas A&M University
21.
Charlton, William S.
Delayed neutron measurements from fast fission of actinide waste isotopes.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4427
► A study was performed to determine the delayed neutron emission properties from fast fission of several actinide waste isotopes. The specific isotopes evaluated were U-235,…
(more)
▼ A study was performed to determine the delayed neutron emission properties from fast fission of several actinide waste isotopes. The specific isotopes evaluated were U-235, Np-237, and Am-243. A calculational technique based on the microscopic method was used to predict initial guesses for the delayed neutron parameters (group decay constants and yields). Based on these calculations, an alternate "seven-group" structure, in contrast to the traditional "six-group" structure used previously, was suggested which would yield a superior fit to the measured data. A series of measurements were performed to test the hypothesis suggested by this alternate group structure. Using a set of highly purified actinide samples (provided by Oak Ridge National Laboratory), the delayed neutron emission decay constants and yields for six groups of the "seven-group" structure were measured for U-235, Np-237, and Am-243. These experiments were performed using the Texas A&M University Nuclear Science Center Reactor, a quick pneumatic transfer system, an integrated computer control and counting system, and a specially designed in-core irradiation device. The values for the total delayed neutron yield (per 100 fissions) from fast-neutron induced fission of U-235, Np237, and Am-243 were determined to be 1.67 ︢0.08, 1.14 ︢0.07, 0.86 ︢0.05, respectively. The newly measured values were compared with other values recommended by Keepin et al., Waldo et al., Saleh et al., and Brady and England. Good agreement was found in all cases. The "seven-group" structure was shown to yield a superior fit to the measured data, as well as, provide a more direct correlation between delayed neutron groups and their associated delayed neutron precursors.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Charlton, W. S. (2012). Delayed neutron measurements from fast fission of actinide waste isotopes. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4427
Chicago Manual of Style (16th Edition):
Charlton, William S. “Delayed neutron measurements from fast fission of actinide waste isotopes.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4427.
MLA Handbook (7th Edition):
Charlton, William S. “Delayed neutron measurements from fast fission of actinide waste isotopes.” 2012. Web. 04 Mar 2021.
Vancouver:
Charlton WS. Delayed neutron measurements from fast fission of actinide waste isotopes. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4427.
Council of Science Editors:
Charlton WS. Delayed neutron measurements from fast fission of actinide waste isotopes. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1997-THESIS-C4427
Texas A&M University
22.
Ortensi, Javier.
Rutherford backscattering analysis of gallium implanted 316 stainless steel.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-O78
► Ion implantation of Ga ions into 316 stainless steel was performed at fluences ranging from 8x10¹⁶ to 10¹⁸ ions/cm². The depth profile of Ga in…
(more)
▼ Ion implantation of Ga ions into 316 stainless steel was performed at fluences ranging from 8x10¹⁶ to 10¹⁸ ions/cm². The depth profile of Ga in the steel was analyzed via Rutherford Backscattering and ToFSIMS. The surface effects were characterized with SEM analysis. Results indicate that Ga saturation was reached at fluences between 2-6x10¹⁷ ions/cm². The maximum Ga concentration occurred near the surface and was between 20 and 25 atomic percent. A constant Ga concentration of 5% was attained at 300 [] and deeper. The possible enhanced diffusion of Ga was observed, but not necessarily through the grain boundaries. Although there was no indication of compound formation, significant pitting was observed at high fluences. Repassivation characteristics of stainless steel may be inhibited at high fluences; therefore future studies are recommended.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Ortensi, J. (2012). Rutherford backscattering analysis of gallium implanted 316 stainless steel. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-O78
Chicago Manual of Style (16th Edition):
Ortensi, Javier. “Rutherford backscattering analysis of gallium implanted 316 stainless steel.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-O78.
MLA Handbook (7th Edition):
Ortensi, Javier. “Rutherford backscattering analysis of gallium implanted 316 stainless steel.” 2012. Web. 04 Mar 2021.
Vancouver:
Ortensi J. Rutherford backscattering analysis of gallium implanted 316 stainless steel. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-O78.
Council of Science Editors:
Ortensi J. Rutherford backscattering analysis of gallium implanted 316 stainless steel. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2000-THESIS-O78
Texas A&M University
23.
Clarno, Kevin Taylor.
Development of a RELAP5-3D three-dimensional model of a VVER-1000 Nuclear Power Plant for analysis of a large-break loss-of-coolant accident.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-C53
► In order to analyze the benefits of the multi-dimensional hydrodynamic modeling capability of the RELAP5-3D system code in the VVER-1000 Nuclear Power Plant, a three-dimensional…
(more)
▼ In order to analyze the benefits of the multi-dimensional hydrodynamic modeling capability of the RELAP5-3D system code in the VVER-1000 Nuclear Power Plant, a three-dimensional model of the core, downcomer, and lower plenum have been created to replace their one-dimensional counterparts in a complete plant model. This multi-dimensional model has been validated with plant operational data and other computer simulations of a thermal-hydraulic transient. The simulated transient considered was a large-break loss-of-coolant accident (LB LOCA). A validated, one-dimensional control of the nuclear power plant, for the study of the effects of mixed oxide (MOX) fuel, was modified to include a standard fuel loading of UO₂. The development of the three-dimensional sections of the reactor vessel consisted of ensuring geometrical fidelity with the design of the modeled plant, the Balacovo Unit 4, Nuclear Power Plant in Saratov, Russia. A stable operational steady-state was obtained and the calculated plant conditions compared well with the design values of the Balacovo Plant. Transient results verified that the simulated thermal-hydraulic conditions of the multi-dimensional model agreed well with both the control and analyses that have been performed separately from this report. It was found that the multi-dimensional model has shown a reduction in the calculated hot-spot peak-clad temperature (PCT) during the blowdown stage of a LB LOCA and an increase in PCT during the reflood stage. A preliminary uncertainty analysis of the PCT during blowdown stage was performed using a response surface method of the Code Scaling, Applicability, and Uncertainty Method and a significant number of relevant input variables. From the preliminary analysis, the PCT reduction during blowdown appears to be significant, but a further, more detailed analysis should be performed, along with an uncertainty analysis of the PCT during the reflood stage. The enhanced depiction of the flow patterns and temperature distributions in the transient situation allowed the user further understanding of the thermal-hydraulic conditions throughout the transient. The developed model proved to be suitable for analysis of the VVER-1000 plant, but to further the applicability of the model a three-dimensional kinetics model of the neutronics and three-dimensional hydrodynamic models of the horizontal steam generators should be included.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Clarno, K. T. (2012). Development of a RELAP5-3D three-dimensional model of a VVER-1000 Nuclear Power Plant for analysis of a large-break loss-of-coolant accident. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-C53
Chicago Manual of Style (16th Edition):
Clarno, Kevin Taylor. “Development of a RELAP5-3D three-dimensional model of a VVER-1000 Nuclear Power Plant for analysis of a large-break loss-of-coolant accident.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-C53.
MLA Handbook (7th Edition):
Clarno, Kevin Taylor. “Development of a RELAP5-3D three-dimensional model of a VVER-1000 Nuclear Power Plant for analysis of a large-break loss-of-coolant accident.” 2012. Web. 04 Mar 2021.
Vancouver:
Clarno KT. Development of a RELAP5-3D three-dimensional model of a VVER-1000 Nuclear Power Plant for analysis of a large-break loss-of-coolant accident. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-C53.
Council of Science Editors:
Clarno KT. Development of a RELAP5-3D three-dimensional model of a VVER-1000 Nuclear Power Plant for analysis of a large-break loss-of-coolant accident. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-C53
Texas A&M University
24.
Garcia, Richard Michael.
Neon time-of-flight backscattering spectrometry for surface analysis.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-G366
► Several time-of-flight backscattering spectrometry experiments using singly ionized neon for purposes of analyzing a bismuth coated silicon target were conducted. This work builds upon prior…
(more)
▼ Several time-of-flight backscattering spectrometry experiments using singly ionized neon for purposes of analyzing a bismuth coated silicon target were conducted. This work builds upon prior work, done with lighter ions, with the goal of comparing the observed system resolutions with those obtained previously in similar analyses. Discriminator thresholds, electronic start/stop logic, aperture sizes, target tilt and beam energies were varied and studied to determine their effect on system resolution. Depth resolutions below 10 A[], energy resolutions less than 1 keV and sensitivities of 10¹¹ (atoms/cm²)were attained. Comparison to prior research indicates the depth resolutions are significantly improved while energy resolutions are slightly improved. However, there does appear to be room to further improve upon these results. This work also indicates the presence of a previously unobserved and unexpected target tail feature on the analysis spectra. A possible cause for the tail feature is suggested and recommendations are given for further research to both improve the system resolution and to further analyze the tail feature.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Garcia, R. M. (2012). Neon time-of-flight backscattering spectrometry for surface analysis. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-G366
Chicago Manual of Style (16th Edition):
Garcia, Richard Michael. “Neon time-of-flight backscattering spectrometry for surface analysis.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-G366.
MLA Handbook (7th Edition):
Garcia, Richard Michael. “Neon time-of-flight backscattering spectrometry for surface analysis.” 2012. Web. 04 Mar 2021.
Vancouver:
Garcia RM. Neon time-of-flight backscattering spectrometry for surface analysis. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-G366.
Council of Science Editors:
Garcia RM. Neon time-of-flight backscattering spectrometry for surface analysis. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-G366
Texas A&M University
25.
Henderson, Ashley David.
The radiological impact of the 2000 Hanford Fire (24-Command Fire).
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-H457
► The range fire at the Hanford facility in late June 2000 coupled with the fire at Los Alamos during the same year have raised a…
(more)
▼ The range fire at the Hanford facility in late June 2000 coupled with the fire at Los Alamos during the same year have raised a number of questions about the potential migration and/or transport of radioactive materials off U.S. nuclear sites into more populated areas. This paper examines the radiological impact of the 24-Command Fire, which occurred on the Hanford Site in late June 2000. Several different approaches are compared against each other to determine the validity of the results. The approaches include physical calculations from collected data as well as estimates from current transport and diffusion software. The analysis begins with the estimation of release. There are sufficient data on the concentrations of radionuclides in the most contaminated areas of the Hanford Site, but very little on the land in between. Once soil concentrations were determined, resuspension factors were applied to estimate releases of material from these areas. A Hanford-specific diffusion and dispersion program, a dose assessment program, and a calculation by hand were used to determine the estimated transport of material to areas populated by the general public. These results are compared against each other as well as the air monitoring results obtained and reported by the United States Environmental Protection Agency and the Washington State Department of Health. Air concentrations from all three methods were used to calculate the associated doses and risks to individuals in these areas. From the analyses, the radiological impact of the fire was determined to be minimal. The ensuing wind events, resuspending particulate matter from the contaminated areas burned during the fire, resulted in a committed effective dose of approximately 10 []Sv (0.01 mrem) from the inhalation of contaminated air. This dose is insignificant when compared to the 360 mrem per year average dose of a member of the general public from indoor and outdoor sources of background radiation. The ingestion pathway was analyzed but found to contribute less than 2 Bq yr⁻¹ for the most important foodstuffs: vegetables and fruits.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Henderson, A. D. (2012). The radiological impact of the 2000 Hanford Fire (24-Command Fire). (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-H457
Chicago Manual of Style (16th Edition):
Henderson, Ashley David. “The radiological impact of the 2000 Hanford Fire (24-Command Fire).” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-H457.
MLA Handbook (7th Edition):
Henderson, Ashley David. “The radiological impact of the 2000 Hanford Fire (24-Command Fire).” 2012. Web. 04 Mar 2021.
Vancouver:
Henderson AD. The radiological impact of the 2000 Hanford Fire (24-Command Fire). [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-H457.
Council of Science Editors:
Henderson AD. The radiological impact of the 2000 Hanford Fire (24-Command Fire). [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-H457
Texas A&M University
26.
Stone, Joseph C.
Delayed neutron measurements for Th-232, Np-237, Pu-239, Pu-241 and depleted uranium.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-S7454
► The neutron emission rates from five very pure actinide samples (Th-232, Np-237, Pu-239, Pu-241 and depleted uranium) were measured following equilibrium irradiation in fast and…
(more)
▼ The neutron emission rates from five very pure actinide samples (Th-232, Np-237, Pu-239, Pu-241 and depleted uranium) were measured following equilibrium irradiation in fast and thermal neutron fluxes. The relative abundances (alphas) for the first four groups were calculated from the delayed neutron emission (counts vs. time) data using Keepin's 6-group decay constants (lambdas) for Th-232, Pu-239 and depleted uranium (both fast and thermal neutron induced fissions). The relative abundances (alphas) for the first five groups were calculated for the fast neutron induced fission of Np-237 using the 7-group lambdas obtained by Charlton (1997). The relative abundances for the first five groups were also calculated using the 7-group lambdas proposed by Loaiza and Haskin (2000), the 8-group lambdas proposed by Campbell and Spriggs (1998) and the 8-group lambdas proposed by Piksaikin (2000) for all of the samples (fast neutron induced fission only for Th-232 and Np-237, fast and thermal neutron induced fission for the remainder). Fission product yield and delayed neutron emission probability data from the ENDF-349 and JEF 2.2 nuclear data libraries were also used to simulate neutron emission data from the samples. The calculated neutron yield curves were used to obtain group relative abundances for each of the five actinide samples (fast neutron induced fission only for Th-232 and Np-237, fast and thermal neutron induced fission for the remainder) based on each set of proposed lambdas. The relative abundances obtained from the experiments and calculations are compared and the differences are noted and discussed.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Stone, J. C. (2012). Delayed neutron measurements for Th-232, Np-237, Pu-239, Pu-241 and depleted uranium. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-S7454
Chicago Manual of Style (16th Edition):
Stone, Joseph C. “Delayed neutron measurements for Th-232, Np-237, Pu-239, Pu-241 and depleted uranium.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-S7454.
MLA Handbook (7th Edition):
Stone, Joseph C. “Delayed neutron measurements for Th-232, Np-237, Pu-239, Pu-241 and depleted uranium.” 2012. Web. 04 Mar 2021.
Vancouver:
Stone JC. Delayed neutron measurements for Th-232, Np-237, Pu-239, Pu-241 and depleted uranium. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-S7454.
Council of Science Editors:
Stone JC. Delayed neutron measurements for Th-232, Np-237, Pu-239, Pu-241 and depleted uranium. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2001-THESIS-S7454
Texas A&M University
27.
Furukawa, Toru.
Three-dimensional reconstruction of bubble distribution in two-phase bubbly flows with the dynamic programming method.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-F96
► A three-dimensional bubble reconstruction method is proposed in this thesis to analyze two-phase bubbly flows. Gas/liquid two-phase flows have important roles in the nuclear and…
(more)
▼ A three-dimensional bubble reconstruction method is proposed in this thesis to analyze two-phase bubbly flows. Gas/liquid two-phase flows have important roles in the nuclear and chemical industries and other engineering fields, but they are not completely understood yet. This indicates more experimental investigation is required. Bubble distribution is very important for the accurate description of bubbly flows and many methods have been proposed. However, conventional methods assume a fixed shape or size of bubbles, and this is not a realistic assumption for the flows with bubbles of complex shapes. The fundamental strategy of the proposed method is similar to tomography, which is an iteration process of guessing and evaluating. The advantages of the proposed method are that it does not need to separate a phantom of a single bubble from another in the shadow image and it does not need to assume fixed shapes or sizes of the bubbles. In the proposed method, a bubble distribution is represented as a metaball object with few parameters needed to describe three-dimensional complex bubble shape. Thus, the proposed method is able to analyze limited data acquired from digital cameras. In contrast, ordinary tomography requires large amount of data. The method is an application of dynamic programming. In this case, the problem of searching the optimal metaball object that represents the bubble distribution the best is divided into smaller problems of searching a semi-optimal object. This approach reduces the hypothetical solution space, and thus the number of match testing processes. Knowledge base is employed here for efficiency. The search algorithm consults the knowledge base to determine whether a solution candidate is worth performing a match testing. This approach isolates the search algorithm from the knowledge base so that a developer can easily add new rules. The method has been applied to synthetically generated shadow image sets in various conditions to evaluate the performance and the images acquired from the real experiment. The result indicates that the reconstructed bubble distribution can be very accurate.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA (6th Edition):
Furukawa, T. (2012). Three-dimensional reconstruction of bubble distribution in two-phase bubbly flows with the dynamic programming method. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-F96
Chicago Manual of Style (16th Edition):
Furukawa, Toru. “Three-dimensional reconstruction of bubble distribution in two-phase bubbly flows with the dynamic programming method.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-F96.
MLA Handbook (7th Edition):
Furukawa, Toru. “Three-dimensional reconstruction of bubble distribution in two-phase bubbly flows with the dynamic programming method.” 2012. Web. 04 Mar 2021.
Vancouver:
Furukawa T. Three-dimensional reconstruction of bubble distribution in two-phase bubbly flows with the dynamic programming method. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-F96.
Council of Science Editors:
Furukawa T. Three-dimensional reconstruction of bubble distribution in two-phase bubbly flows with the dynamic programming method. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-F96
Texas A&M University
28.
Gautier, Vincent Charles.
Development of Nuclear Reactor remote Monitoring software (NRM) for the Star project.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-G385
► As a response to the needs of developing countries to meet their rapidly growing energy requirements, the Safe, Transportable, Autonomous Reactor (STAR) program originated. This…
(more)
▼ As a response to the needs of developing countries to meet their rapidly growing energy requirements, the Safe, Transportable, Autonomous Reactor (STAR) program originated. This concept relies on small, passively safe, and highly autonomous nuclear power stations to make nuclear energy available to countries not containing the infrastructure to support a conventional reactor. Nuclear Reactor remote Monitoring software (NRM), a part of this project, has been developed to centralize and coordinate operations support for multiple plants, guarantee safety and nonproliferation, and improve cost-effectiveness. NRM allows soft real time monitoring of the reactor power, primary and secondary coolant temperatures, control rod movement, and secondary coolant zones in the steam generator (subcooled, saturated, and superheated). The coherence of the data sets is checked with QUASIMODO, a thermohydraulics computational code. NRM receives the measurement of physical variables from a data acquisition system. This data is then encrypted and transferred to the monitoring center using Secure SHell (SSH). The main program, written in C and Java embedded with Java Native Interface (JNI) processes the data, which is displayed in a graphic user interface along with calculated results.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA ·
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APA (6th Edition):
Gautier, V. C. (2012). Development of Nuclear Reactor remote Monitoring software (NRM) for the Star project. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-G385
Chicago Manual of Style (16th Edition):
Gautier, Vincent Charles. “Development of Nuclear Reactor remote Monitoring software (NRM) for the Star project.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-G385.
MLA Handbook (7th Edition):
Gautier, Vincent Charles. “Development of Nuclear Reactor remote Monitoring software (NRM) for the Star project.” 2012. Web. 04 Mar 2021.
Vancouver:
Gautier VC. Development of Nuclear Reactor remote Monitoring software (NRM) for the Star project. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-G385.
Council of Science Editors:
Gautier VC. Development of Nuclear Reactor remote Monitoring software (NRM) for the Star project. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-G385
Texas A&M University
29.
Helton, Donald McLean.
Calculation of unsteady turbulent flow around obstacles using the large eddy simulation turbulence model.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-H455
► The premise of the work presented here is to use a common analytical tool, Computational Fluid Dynamics (CFD), along with a prevalent turbulence model, Large…
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▼ The premise of the work presented here is to use a common analytical tool, Computational Fluid Dynamics (CFD), along with a prevalent turbulence model, Large Eddy Simulation (LES), to study the flow past rectangular cylinders. In an attempt to use CFD simulations to model the cylinder flow phenomena, a suitable CFD code (Trio_U) was selected, and implemented. A validation test of flow around a cylinder in an open channel was studied, and grid / model sensitivities were explored. Trio's LES capabilities were then extended to study obstacle flow in a closed channel. Mesh resolution tests indicated that an instability in the solution procedure disallows a grid independent solution. However, it was determined that fine and moderate grid resolutions were able to reproduce experimental results. Through further testing, it was determined that the 4th order central difference scheme, the Smagorinsky sub-grid scale model and the standard law of the wall model were most suitable for their respective applications. In addition, inlet condition tests were run which indicated that a constant inlet velocity condition is the most suitable boundary condition for these simulations. Lastly, extension to a larger grid in the z-direction resulted in increased turbulence dissipation. When the same obstacle is placed at the center of a closed channel, results were very similar to the open channel case, with slight changes in the near-wake velocities, and a slight increase of the drag coefficient. Interaction between the obstacle and the wall was negligible. However, at distances greater than x = 5D behind the obstacle, cylinder-generated vortices were observed to perturb the boundary layer slightly. When the obstacle was moved to a distance of 1D from the lower wall, significant changes in the results were observed. The drag coefficient increased by 17%, and the near-wake turbulent velocity was also shown to have increased. Vortex formation behind the cylinder was observed to occur at an angle to the channel centerline. Furthermore, substantial interaction between the obstacle and the wall was noted, both near the obstacle and further downstream. The boundary layer downstream of the obstacle was dramatically affected by the obstacle's presence. The cumulative result is a flow field which contains physical behavior that is much more complex than that of the freely-stationed obstacle.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA ·
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APA (6th Edition):
Helton, D. M. (2012). Calculation of unsteady turbulent flow around obstacles using the large eddy simulation turbulence model. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-H455
Chicago Manual of Style (16th Edition):
Helton, Donald McLean. “Calculation of unsteady turbulent flow around obstacles using the large eddy simulation turbulence model.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-H455.
MLA Handbook (7th Edition):
Helton, Donald McLean. “Calculation of unsteady turbulent flow around obstacles using the large eddy simulation turbulence model.” 2012. Web. 04 Mar 2021.
Vancouver:
Helton DM. Calculation of unsteady turbulent flow around obstacles using the large eddy simulation turbulence model. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-H455.
Council of Science Editors:
Helton DM. Calculation of unsteady turbulent flow around obstacles using the large eddy simulation turbulence model. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-H455
Texas A&M University
30.
Jiltchenkov, Dmitri Victorovich.
The development of a remote monitoring system for the Nuclear Science Center reactor.
Degree: MS, nuclear engineering, 2012, Texas A&M University
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-J57
► With funding provided by Nuclear Energy Research Initiative (NERI), design of Secure, Transportable, Autonomous Reactors (STAR) to aid countries with insufficient energy supplies is underway.…
(more)
▼ With funding provided by Nuclear Energy Research Initiative (NERI), design of Secure, Transportable, Autonomous Reactors (STAR) to aid countries with insufficient energy supplies is underway. The development of a new monitoring system that allows remote access to data from the reactor site is an important part of this project. The two goals of this monitoring system are to control the use of nuclear materials and to monitor the performance of the facility from a remote location. I have designed a prototype system for this NERI project that utilizes LabVIEW software and global network technologies to monitor the Nuclear Science Center (NSC) reactor at Texas A&M University. LabVIEW and its applications have all the needed features to build a monitoring system for many types of facilities, including STAR reactors. This system takes data from reactor cooling systems, power monitoring channels, fuel temperature indicators, control rod drives, security alarm sensors and stores it on local and remote hard drives, sends it through an output port to remote clients, and graphically displays these data in the reactor control room. Data from NSC TRIGA reactor is fed to a computer program that analyzes and predicts reactor performance in real time. To provide a remote observation of the working area and fissile material, this system uses cameras, triggered by alarm sensors and LabVIEW vision applications. Operators at the local and remote control stations may view and store all the images from these cameras. The system has been in operation for many months at the NSC with outstanding results and further development is continuing.
Subjects/Keywords: nuclear engineering.; Major nuclear engineering.
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APA ·
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MLA ·
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APA (6th Edition):
Jiltchenkov, D. V. (2012). The development of a remote monitoring system for the Nuclear Science Center reactor. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-J57
Chicago Manual of Style (16th Edition):
Jiltchenkov, Dmitri Victorovich. “The development of a remote monitoring system for the Nuclear Science Center reactor.” 2012. Masters Thesis, Texas A&M University. Accessed March 04, 2021.
http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-J57.
MLA Handbook (7th Edition):
Jiltchenkov, Dmitri Victorovich. “The development of a remote monitoring system for the Nuclear Science Center reactor.” 2012. Web. 04 Mar 2021.
Vancouver:
Jiltchenkov DV. The development of a remote monitoring system for the Nuclear Science Center reactor. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 04].
Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-J57.
Council of Science Editors:
Jiltchenkov DV. The development of a remote monitoring system for the Nuclear Science Center reactor. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2002-THESIS-J57
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