subject:(Nuclear fuels)

Oregon State University

1. Reardon, Patrick Thomas. A two-dimensional analysis of porosity effects on the thermal conductivity of ceramic nuclear fuels.

Degree: MS, Nuclear Engineering, 1983, Oregon State University

URL: http://hdl.handle.net/1957/41793

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Reardon, P. T. (1983). A two-dimensional analysis of porosity effects on the thermal conductivity of ceramic nuclear fuels. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/41793

Chicago Manual of Style (16^th Edition):

Reardon, Patrick Thomas. “A two-dimensional analysis of porosity effects on the thermal conductivity of ceramic nuclear fuels.” 1983. Masters Thesis, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/41793.

MLA Handbook (7^th Edition):

Reardon, Patrick Thomas. “A two-dimensional analysis of porosity effects on the thermal conductivity of ceramic nuclear fuels.” 1983. Web. 18 Apr 2021.

Vancouver:

Reardon PT. A two-dimensional analysis of porosity effects on the thermal conductivity of ceramic nuclear fuels. [Internet] [Masters thesis]. Oregon State University; 1983. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/41793.

Council of Science Editors:

Reardon PT. A two-dimensional analysis of porosity effects on the thermal conductivity of ceramic nuclear fuels. [Masters Thesis]. Oregon State University; 1983. Available from: http://hdl.handle.net/1957/41793

Oregon State University

2. Jenks, Richard P. Direct shear testing of UO₂ sphere-pac nuclear reactor fuel.

Degree: MS, Nuclear Engineering, 1982, Oregon State University

URL: http://hdl.handle.net/1957/41755

► The design, development, and utilization of a direct shear test apparatus is presented in this thesis. The effectiveness of the device for determining the friction… (more)

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Jenks, R. P. (1982). Direct shear testing of UO₂ sphere-pac nuclear reactor fuel. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/41755

Chicago Manual of Style (16^th Edition):

Jenks, Richard P. “Direct shear testing of UO₂ sphere-pac nuclear reactor fuel.” 1982. Masters Thesis, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/41755.

MLA Handbook (7^th Edition):

Jenks, Richard P. “Direct shear testing of UO₂ sphere-pac nuclear reactor fuel.” 1982. Web. 18 Apr 2021.

Vancouver:

Jenks RP. Direct shear testing of UO₂ sphere-pac nuclear reactor fuel. [Internet] [Masters thesis]. Oregon State University; 1982. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/41755.

Council of Science Editors:

Jenks RP. Direct shear testing of UO₂ sphere-pac nuclear reactor fuel. [Masters Thesis]. Oregon State University; 1982. Available from: http://hdl.handle.net/1957/41755

Oregon State University

3. Montgomery, Scott D. Thermal and mechanical behavior of sphere pac fuel.

Degree: MS, Nuclear Engineering, 1982, Oregon State University

URL: http://hdl.handle.net/1957/41981

► The use of advanced concepts can make the nuclear fuel cycle more efficient by allowing pins to be operated at high powers and to higher… (more)

▼ The use of advanced concepts can make the nuclear fuel cycle more efficient by allowing pins to be operated at high powers and to higher burnups. Among the more promising concepts in advanced fuels is sphere pac fuel. In a first attempt to bring together models describing individual phenomena, the computer code SPECKLE-I was written. In this thesis, the results obtained from modified and refined versions of this code are compared with data obtained from examinations of irradiated sphere pac fuel pins, and predictions for future irradiations are discussed. The values for fractional gas release and free gas pressure and the location of the sintering boundary obtained from SPECKLE-I calculations are compared with values derived from the destructive examination of mixed carbide sphere pac fuel pins. SPECKLE-I seems to underestimate the amount of fission gas release and the extent of restructuring that occurred in these pins. The code was modified to include the effects of fission gas release due to recoil from the fission event. While a large effect can be seen for significant release rates, it has been seen from low burnup experiments that the release rate due to this mechanism is very small. In another set of calculations, the sintering rates due to surface diffusion and volume diffusion were added directly to the sintering rate due to grain boundary diffusion. The differences in the results for higher power irradiations between the single sintering mechanism case and the cumulative sintering mechanism case were small. They were more pronounced for lower power and lower temperature experiments. SPECKLE-I was modified to model the behavior of a sphere pac fuel pin during an initial startup sequence. A series of calculations were performed to analyze the power-to-melt ratio of several initial startup sequences. The results of this work indicated that the peak pin temperatures could be manipulated to meet specified safety criteria by tailoring the startup sequence. The effect on the calculations of the use of a new correlation for the thermal conductivity of a partially restructured particle bed was also examined. The results of a set of experiments to determine the axial shear force neccessary to overcome the shear forces exerted on a simulated clad tube are presented. A binary mixture was used in the experiment. There seemed to be a relation between the force exerted by the bed on the clad and the number of contact points between the bed and the clad. Advisors/Committee Members: Peddicord, K. L. (advisor).

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Montgomery, S. D. (1982). Thermal and mechanical behavior of sphere pac fuel. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/41981

Chicago Manual of Style (16^th Edition):

Montgomery, Scott D. “Thermal and mechanical behavior of sphere pac fuel.” 1982. Masters Thesis, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/41981.

MLA Handbook (7^th Edition):

Montgomery, Scott D. “Thermal and mechanical behavior of sphere pac fuel.” 1982. Web. 18 Apr 2021.

Vancouver:

Montgomery SD. Thermal and mechanical behavior of sphere pac fuel. [Internet] [Masters thesis]. Oregon State University; 1982. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/41981.

Council of Science Editors:

Montgomery SD. Thermal and mechanical behavior of sphere pac fuel. [Masters Thesis]. Oregon State University; 1982. Available from: http://hdl.handle.net/1957/41981

Oregon State University

4. Henke, Robert K. Stochastic simulation of sphere beds and the effect of packing density on the thermal conductivity of sphere pac nuclear fuels.

Degree: MS, Nuclear Engineering, 1981, Oregon State University

URL: http://hdl.handle.net/1957/42199

► The establishment of any thermal model of reactor fuel requires a thorough understanding of the behavior of the thermal conductivity. This is especially true in… (more)

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Henke, R. K. (1981). Stochastic simulation of sphere beds and the effect of packing density on the thermal conductivity of sphere pac nuclear fuels. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/42199

Chicago Manual of Style (16^th Edition):

Henke, Robert K. “Stochastic simulation of sphere beds and the effect of packing density on the thermal conductivity of sphere pac nuclear fuels.” 1981. Masters Thesis, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/42199.

MLA Handbook (7^th Edition):

Henke, Robert K. “Stochastic simulation of sphere beds and the effect of packing density on the thermal conductivity of sphere pac nuclear fuels.” 1981. Web. 18 Apr 2021.

Vancouver:

Henke RK. Stochastic simulation of sphere beds and the effect of packing density on the thermal conductivity of sphere pac nuclear fuels. [Internet] [Masters thesis]. Oregon State University; 1981. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/42199.

Council of Science Editors:

Henke RK. Stochastic simulation of sphere beds and the effect of packing density on the thermal conductivity of sphere pac nuclear fuels. [Masters Thesis]. Oregon State University; 1981. Available from: http://hdl.handle.net/1957/42199

Oregon State University

5. Nielsen, Larry H. An assessment of describing function methods applied to reactor fuel performance analysis.

Degree: MS, Nuclear Engineering, 1980, Oregon State University

URL: http://hdl.handle.net/1957/42808

► Large digital computer reactor fuel performance codes perform repeated calculations of temperature profiles during transients. The thermal response is required to calculate the mechanical behavior.… (more)

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Nielsen, L. H. (1980). An assessment of describing function methods applied to reactor fuel performance analysis. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/42808

Chicago Manual of Style (16^th Edition):

Nielsen, Larry H. “An assessment of describing function methods applied to reactor fuel performance analysis.” 1980. Masters Thesis, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/42808.

MLA Handbook (7^th Edition):

Nielsen, Larry H. “An assessment of describing function methods applied to reactor fuel performance analysis.” 1980. Web. 18 Apr 2021.

Vancouver:

Nielsen LH. An assessment of describing function methods applied to reactor fuel performance analysis. [Internet] [Masters thesis]. Oregon State University; 1980. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/42808.

Council of Science Editors:

Nielsen LH. An assessment of describing function methods applied to reactor fuel performance analysis. [Masters Thesis]. Oregon State University; 1980. Available from: http://hdl.handle.net/1957/42808

Oregon State University

6. Baker, Ronald Blair. Modeling the thermal effects of fabricated fuel-to-cladding gap in fast reactor fuel.

Degree: MS, Mechanical Engineering, 1976, Oregon State University

URL: http://hdl.handle.net/1957/43628

► This report presents a refined method for the calculation of fuel centerline temperatures and the linear-heat-rating-to-incipient melting, ' Q'[subscript m], of mixed-oxide fuel typical of… (more)

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Baker, R. B. (1976). Modeling the thermal effects of fabricated fuel-to-cladding gap in fast reactor fuel. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/43628

Chicago Manual of Style (16^th Edition):

Baker, Ronald Blair. “Modeling the thermal effects of fabricated fuel-to-cladding gap in fast reactor fuel.” 1976. Masters Thesis, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/43628.

MLA Handbook (7^th Edition):

Baker, Ronald Blair. “Modeling the thermal effects of fabricated fuel-to-cladding gap in fast reactor fuel.” 1976. Web. 18 Apr 2021.

Vancouver:

Baker RB. Modeling the thermal effects of fabricated fuel-to-cladding gap in fast reactor fuel. [Internet] [Masters thesis]. Oregon State University; 1976. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/43628.

Council of Science Editors:

Baker RB. Modeling the thermal effects of fabricated fuel-to-cladding gap in fast reactor fuel. [Masters Thesis]. Oregon State University; 1976. Available from: http://hdl.handle.net/1957/43628

Oregon State University

7. Alberti, Anthony L. Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility.

Degree: MS, Nuclear Engineering, 2015, Oregon State University

URL: http://hdl.handle.net/1957/58315

► With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption… (more)

Subjects/Keywords: TREAT; Nuclear fuels – Testing

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APA (6^th Edition):

Alberti, A. L. (2015). Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/58315

Chicago Manual of Style (16^th Edition):

Alberti, Anthony L. “Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility.” 2015. Masters Thesis, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/58315.

MLA Handbook (7^th Edition):

Alberti, Anthony L. “Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility.” 2015. Web. 18 Apr 2021.

Vancouver:

Alberti AL. Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility. [Internet] [Masters thesis]. Oregon State University; 2015. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/58315.

Council of Science Editors:

Alberti AL. Steady State Modeling of the Minimum Critical Core of the Transient Reactor Test Facility. [Masters Thesis]. Oregon State University; 2015. Available from: http://hdl.handle.net/1957/58315

Oregon State University

8. Wen, Xianfei. An Active Interrogation Technique Based on Photofission for Non-destructive Assay of Used Nuclear Fuel.

Degree: PhD, Nuclear Engineering, 2016, Oregon State University

URL: http://hdl.handle.net/1957/58469

► High-energy delayed γ-rays from photofission were demonstrated to be signatures for detection and identification of special nuclear materials. Such γ-rays were measured in between linac… (more)

Subjects/Keywords: Photofission; Nuclear fuels – Analysis

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APA (6^th Edition):

Wen, X. (2016). An Active Interrogation Technique Based on Photofission for Non-destructive Assay of Used Nuclear Fuel. (Doctoral Dissertation). Oregon State University. Retrieved from http://hdl.handle.net/1957/58469

Chicago Manual of Style (16^th Edition):

Wen, Xianfei. “An Active Interrogation Technique Based on Photofission for Non-destructive Assay of Used Nuclear Fuel.” 2016. Doctoral Dissertation, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/58469.

MLA Handbook (7^th Edition):

Wen, Xianfei. “An Active Interrogation Technique Based on Photofission for Non-destructive Assay of Used Nuclear Fuel.” 2016. Web. 18 Apr 2021.

Vancouver:

Wen X. An Active Interrogation Technique Based on Photofission for Non-destructive Assay of Used Nuclear Fuel. [Internet] [Doctoral dissertation]. Oregon State University; 2016. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/58469.

Council of Science Editors:

Wen X. An Active Interrogation Technique Based on Photofission for Non-destructive Assay of Used Nuclear Fuel. [Doctoral Dissertation]. Oregon State University; 2016. Available from: http://hdl.handle.net/1957/58469

Oregon State University

9. Bailey, Wendell Jack. Investigation of plutonium-bearing fuel elements for the plutonium recycle test reactor (PRTR).

Degree: MS, Mechanical Engineering, 1960, Oregon State University

URL: http://hdl.handle.net/1957/50363

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Bailey, W. J. (1960). Investigation of plutonium-bearing fuel elements for the plutonium recycle test reactor (PRTR). (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/50363

Chicago Manual of Style (16^th Edition):

Bailey, Wendell Jack. “Investigation of plutonium-bearing fuel elements for the plutonium recycle test reactor (PRTR).” 1960. Masters Thesis, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/50363.

MLA Handbook (7^th Edition):

Bailey, Wendell Jack. “Investigation of plutonium-bearing fuel elements for the plutonium recycle test reactor (PRTR).” 1960. Web. 18 Apr 2021.

Vancouver:

Bailey WJ. Investigation of plutonium-bearing fuel elements for the plutonium recycle test reactor (PRTR). [Internet] [Masters thesis]. Oregon State University; 1960. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/50363.

Council of Science Editors:

Bailey WJ. Investigation of plutonium-bearing fuel elements for the plutonium recycle test reactor (PRTR). [Masters Thesis]. Oregon State University; 1960. Available from: http://hdl.handle.net/1957/50363

Oregon State University

10. Stout, Richard Bartholomew. Optimization of in-core nuclear fuel management in a pressurized water reactor.

Degree: PhD, Mechanical and Nuclear Engineering, 1972, Oregon State University

URL: http://hdl.handle.net/1957/44778

► Fuel loading patterns which have a minimum power peak are economically desirable to allow power reactors to operate at the highest possible power density and… (more)

▼ Fuel loading patterns which have a minimum power peak are economically desirable to allow power reactors to operate at the highest possible power density and to minimize the possibility of fuel failure. A computer code called SHUFLE was developed for pressurized water reactors which shuffles the fuel in search of the lowest possible power peaking factor. An iterative approach is used in this search routine. A radial power distribution is calculated from which the program logic Selects a movement of fuel elements in an attempt to lower the radial power peak. Another power calculation is made and the process repeated until a predetermined convergence is reached. The logic by which the code decides the fuel movement is presented, along with the criteria for accepting or rejecting the move after a power calculation of the new loading pattern is made. A 1.5 group course mesh diffusion theory method was used to obtain the power distribution for each SHUFLE iteration. Convergence to a final loading pattern varies from about 10 to 40 shuffling iterations depending on the initial loading presented to the code. Since the typical computer running time for a one-quarter core power distribution with this 1.5 group method is only one to a few seconds, depending on the loading, convergence to a good loading pattern takes on the order of one minute on a Univac 1108. The low computer cost plus ease of operation should make this code of considerable use in determining loading patterns with minimum power peaking for any given set of fuel elements. The program also has burnup capability which can be used to check power peaking throughout core life. A parametric analysis study of fuel cycle costs for a PWR is also presented. Cost parameters analyzed were variation in the cost of yellow cake, enrichment, money, fabrication, and reprocessing plus changes in burnup, load factors, power densities, and the effect of forced early discharge. Figures are presented to indicate total fuel costs as a function of burnup for these cost parameters. Linear relationships for minimum cost and optimum burnup are presented for each parameter. Advisors/Committee Members: Robinson, Alan H. (advisor).

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Stout, R. B. (1972). Optimization of in-core nuclear fuel management in a pressurized water reactor. (Doctoral Dissertation). Oregon State University. Retrieved from http://hdl.handle.net/1957/44778

Chicago Manual of Style (16^th Edition):

Stout, Richard Bartholomew. “Optimization of in-core nuclear fuel management in a pressurized water reactor.” 1972. Doctoral Dissertation, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/44778.

MLA Handbook (7^th Edition):

Stout, Richard Bartholomew. “Optimization of in-core nuclear fuel management in a pressurized water reactor.” 1972. Web. 18 Apr 2021.

Vancouver:

Stout RB. Optimization of in-core nuclear fuel management in a pressurized water reactor. [Internet] [Doctoral dissertation]. Oregon State University; 1972. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/44778.

Council of Science Editors:

Stout RB. Optimization of in-core nuclear fuel management in a pressurized water reactor. [Doctoral Dissertation]. Oregon State University; 1972. Available from: http://hdl.handle.net/1957/44778

Oregon State University

11. Guenther, Richard John. The applicability of sintering mechanisms in the initial stage restructuring and preferential pore migration of irradiated (U₀.₈₅Pu₀.₁₅)C sphere-pac fuel.

Degree: MS, Nuclear Engineering, 1978, Oregon State University

URL: http://hdl.handle.net/1957/42739

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Guenther, R. J. (1978). The applicability of sintering mechanisms in the initial stage restructuring and preferential pore migration of irradiated (U₀.₈₅Pu₀.₁₅)C sphere-pac fuel. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/42739

Chicago Manual of Style (16^th Edition):

Guenther, Richard John. “The applicability of sintering mechanisms in the initial stage restructuring and preferential pore migration of irradiated (U₀.₈₅Pu₀.₁₅)C sphere-pac fuel.” 1978. Masters Thesis, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/42739.

MLA Handbook (7^th Edition):

Guenther, Richard John. “The applicability of sintering mechanisms in the initial stage restructuring and preferential pore migration of irradiated (U₀.₈₅Pu₀.₁₅)C sphere-pac fuel.” 1978. Web. 18 Apr 2021.

Vancouver:

Guenther RJ. The applicability of sintering mechanisms in the initial stage restructuring and preferential pore migration of irradiated (U₀.₈₅Pu₀.₁₅)C sphere-pac fuel. [Internet] [Masters thesis]. Oregon State University; 1978. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/42739.

Council of Science Editors:

Guenther RJ. The applicability of sintering mechanisms in the initial stage restructuring and preferential pore migration of irradiated (U₀.₈₅Pu₀.₁₅)C sphere-pac fuel. [Masters Thesis]. Oregon State University; 1978. Available from: http://hdl.handle.net/1957/42739

Kansas State University

12. Giefer, David L. The irradiation effects on a model TBP-decahydranaphthalene solvent extraction system.

Degree: 1974, Kansas State University

URL: http://hdl.handle.net/2097/10983

Subjects/Keywords: Nuclear fuels

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APA (6^th Edition):

Giefer, D. L. (1974). The irradiation effects on a model TBP-decahydranaphthalene solvent extraction system. (Thesis). Kansas State University. Retrieved from http://hdl.handle.net/2097/10983

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Giefer, David L. “The irradiation effects on a model TBP-decahydranaphthalene solvent extraction system.” 1974. Thesis, Kansas State University. Accessed April 18, 2021. http://hdl.handle.net/2097/10983.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Giefer, David L. “The irradiation effects on a model TBP-decahydranaphthalene solvent extraction system.” 1974. Web. 18 Apr 2021.

Vancouver:

Giefer DL. The irradiation effects on a model TBP-decahydranaphthalene solvent extraction system. [Internet] [Thesis]. Kansas State University; 1974. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/2097/10983.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Giefer DL. The irradiation effects on a model TBP-decahydranaphthalene solvent extraction system. [Thesis]. Kansas State University; 1974. Available from: http://hdl.handle.net/2097/10983

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Tartu University

13. Auzans, Aris. Development of computational model for nuclear energy systems analysis: natural resources optimisation and radiological impact minimization .

Degree: 2018, Tartu University

URL: http://hdl.handle.net/10062/59904

► Paljud riigid püüavad piirata kasvuhoonegaaside heitkoguseid ja püüavad keskenduda süsinikuvabadele energiaallikatele, mistõttu on tuumaenergia jätkuva arutelu teema, et laiendada kogu maailma energiatootmise segmenti ja katta… (more)

▼ Paljud riigid püüavad piirata kasvuhoonegaaside heitkoguseid ja püüavad keskenduda süsinikuvabadele energiaallikatele, mistõttu on tuumaenergia jätkuva arutelu teema, et laiendada kogu maailma energiatootmise segmenti ja katta baaskoormuse nõudeid. On vaja hinnangut turul pakutavatele uraanitarnetele ja vastavat ressursside optimeerimise konteksti, tegemaks otsuseid, mis mõjutavad tuumaenergia pikaajalist arengut, ning hoiatab poliitikakujundajaid võimalikest uraaniturul toimuvatest muutustest, et aidata teha tehnoloogilisi valikuid. Käesoleva töö eesmärgiks on demonstreerida ja katsetada arenenud simulatsioonimeetodeid ja kontseptsioone, et luua realistlikumaid kütusetsüklite simulaatoreid ja seeläbi luua otsusetegijatele vajalik abivahend. Lisaks lahendab töö optimeerimisparadigma, leidmaks soodsaima tuumakütusetsükli tehnoloogia, mis aitab säästa loodusvarasid, suudab minimiseerida kõrgetasemelist radioaktiivsete jäätmete mõju loodusele ja vähendades tuumamaterjalide levikust tulenevat; Many countries are struggling to limit greenhouse gas emissions and focus on carbon-free energy sources, and therefore nuclear energy is a subject of continuing debate to enhance the worldwide energy production mix and cover the base load requirements. Evaluation of uranium supply in a market and resource optimization context is needed to inform decisions impacting the long-term development of nuclear power and warn policy makers about possible uranium market supply-side volatilities and to help choose the technology mix. The aim of this work is to demonstrate and test advanced simulation methods and conceptual ideas to enable realistic fuel cycle simulators and develop a supporting tool for decision makers. Additionally, this work facilitates the selection of the most favorable nuclear fuel cycle, using optimization with respect to multiple criteria save natural resources, minimize the impact of high level radioactive waste on nature, to and decrease nuclear material proliferation risks. Advisors/Committee Members: Tkaczyk, Alan Henry, juhendaja (advisor), Vainikko, Eero, juhendaja (advisor).

Subjects/Keywords: nuclear fuels; natural uranium fuels; nuclear technology; nuclear power energetics; economic models

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APA (6^th Edition):

Auzans, A. (2018). Development of computational model for nuclear energy systems analysis: natural resources optimisation and radiological impact minimization . (Thesis). Tartu University. Retrieved from http://hdl.handle.net/10062/59904

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Auzans, Aris. “Development of computational model for nuclear energy systems analysis: natural resources optimisation and radiological impact minimization .” 2018. Thesis, Tartu University. Accessed April 18, 2021. http://hdl.handle.net/10062/59904.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Auzans, Aris. “Development of computational model for nuclear energy systems analysis: natural resources optimisation and radiological impact minimization .” 2018. Web. 18 Apr 2021.

Vancouver:

Auzans A. Development of computational model for nuclear energy systems analysis: natural resources optimisation and radiological impact minimization . [Internet] [Thesis]. Tartu University; 2018. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/10062/59904.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Auzans A. Development of computational model for nuclear energy systems analysis: natural resources optimisation and radiological impact minimization . [Thesis]. Tartu University; 2018. Available from: http://hdl.handle.net/10062/59904

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Oregon State University

14. Azarfar, Mohammad Ali. A methodology for safeguards evaluation and its application to mixed oxide fuel fabrication.

Degree: PhD, Nuclear Engineering, 1982, Oregon State University

URL: http://hdl.handle.net/1957/41719

► This study focuses on the development of a methodology to evaluate safeguardability of various processes in the nuclear fuel cycle. National and international implementation of… (more)

▼ This study focuses on the development of a methodology to evaluate safeguardability of various processes in the nuclear fuel cycle. National and international implementation of safeguards systems has been investigated. It was concluded that proliferation is purely a political question that, hopefully, can be prevented by international norms and diplomacy; it does not depend on technical differences between the different processes. On the other hand, the prevention of diversion does depend on technical barriers implemented in nuclear facilities. Therefore, different processes with different technical fixes might have varying degrees of resistivity against diversion. Safeguardability of a process can not be evaluated in absolute terms; rather it can be compared with and ranked against another process. In addition, processes by themselves can not be compared; only plants embodying processes can be compared. This is because processes are not safeguardable but plants incorporating them are. A methodology was developed which compares and ranks various processes in the nuclear fuel cycle. It estimates the weights of different safeguard characteristics of a nuclear facility and calculates a dimensionless number – safeguards index – for each process under evaluation. By comparing these safeguards indices, one is able to rank processes against each other. As a case study, comparison of two fast reactor fuel fabrication processes, pelletization and spherepac, were selected. At present, no fast reactor fuel fabrication facility exists on a commercial scale. Hence, two fuel fabrication facilities, each incorporating one process, were designed, in enough detail to make the safeguards comparison possible. Safeguards indices were calculated for these two plants based on the information derived from their designs. The result revealed that the spherepac process is higher in rank relative to the pellet process; the spherepac process is simpler to safeguard relative to the pellet process. Derivation of some of the data used in calculation of safeguards indices was based partly upon subjective judgment. A sensitivity analysis was conducted to test the dependency of the result on subjective judgment. It was concluded that as long as the probability estimation and numerical value assignments remain in a reasonable range, the method is relatively insensitive to subjective judgment. Advisors/Committee Members: Spinrad, B. I. (advisor).

Subjects/Keywords: Nuclear fuels – Methodology

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APA (6^th Edition):

Azarfar, M. A. (1982). A methodology for safeguards evaluation and its application to mixed oxide fuel fabrication. (Doctoral Dissertation). Oregon State University. Retrieved from http://hdl.handle.net/1957/41719

Chicago Manual of Style (16^th Edition):

Azarfar, Mohammad Ali. “A methodology for safeguards evaluation and its application to mixed oxide fuel fabrication.” 1982. Doctoral Dissertation, Oregon State University. Accessed April 18, 2021. http://hdl.handle.net/1957/41719.

MLA Handbook (7^th Edition):

Azarfar, Mohammad Ali. “A methodology for safeguards evaluation and its application to mixed oxide fuel fabrication.” 1982. Web. 18 Apr 2021.

Vancouver:

Azarfar MA. A methodology for safeguards evaluation and its application to mixed oxide fuel fabrication. [Internet] [Doctoral dissertation]. Oregon State University; 1982. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/1957/41719.

Council of Science Editors:

Azarfar MA. A methodology for safeguards evaluation and its application to mixed oxide fuel fabrication. [Doctoral Dissertation]. Oregon State University; 1982. Available from: http://hdl.handle.net/1957/41719

University of Arizona

15. Smith, Adrienne Bobbette, 1960-. Nuclear excursions in aqueous solutions of fissile materials .

Degree: 1989, University of Arizona

URL: http://hdl.handle.net/10150/277150

► Fissile materials in the form of aqueous homogeneous solutions are used during the chemical processing of nuclear materials. In this form there exists the possibility… (more)

Subjects/Keywords: Nuclear fuels.; Nuclear reactor accidents

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APA (6^th Edition):

Smith, Adrienne Bobbette, 1. (1989). Nuclear excursions in aqueous solutions of fissile materials . (Masters Thesis). University of Arizona. Retrieved from http://hdl.handle.net/10150/277150

Chicago Manual of Style (16^th Edition):

Smith, Adrienne Bobbette, 1960-. “Nuclear excursions in aqueous solutions of fissile materials .” 1989. Masters Thesis, University of Arizona. Accessed April 18, 2021. http://hdl.handle.net/10150/277150.

MLA Handbook (7^th Edition):

Smith, Adrienne Bobbette, 1960-. “Nuclear excursions in aqueous solutions of fissile materials .” 1989. Web. 18 Apr 2021.

Vancouver:

Smith, Adrienne Bobbette 1. Nuclear excursions in aqueous solutions of fissile materials . [Internet] [Masters thesis]. University of Arizona; 1989. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/10150/277150.

Council of Science Editors:

Smith, Adrienne Bobbette 1. Nuclear excursions in aqueous solutions of fissile materials . [Masters Thesis]. University of Arizona; 1989. Available from: http://hdl.handle.net/10150/277150

University of Texas – Austin

16. Matavosian, Robert. Complex problems arising in the collision probability theory for neutron transport.

Degree: PhD, Mechanical Engineering., 2007, University of Texas – Austin

URL: http://hdl.handle.net/2152/3789

► Several comprehensive but time consuming neutronic codes are available for performing nuclear reactor and fuel cycle evaluations. In addition, simple models utilizing collision probability theory… (more)

▼ Several comprehensive but time consuming neutronic codes are available for performing nuclear reactor and fuel cycle evaluations. In addition, simple models utilizing collision probability theory are used to perform similar tasks with reasonable accuracy. However, the current collision probability theory treats the heterogeneous reactor configurations with a two region unit cell model. This model does not address several important reactor parameters including spatial self-shielding effects and simultaneous use of different reactor fuels within a reactor core. This dissertation studies the fidelity of expanding the collision probability theory to address the stated shortcomings through analyzing two problems. Problem 1 analyzes the effects of self-shielding. The cylindrical fuel region is divided into several sub-regions and an overall equivalent escape probability from the entire fuel region is developed based on the identified neutron transmission and escape probabilities within each fuel sub-region. The multiplication factor and radioisotopic inventory results based on modified V:BUDS (Visualize: Burnup, Depletion, Spectrum) code are in good agreement with benchmark scenarios for a reactor unit cell. The accurate multiplication factor calculation allows more accurate studies on the maximum fuel burnup and radionuclide inventories of interest in nuclear non-proliferation studies. Problem 2 analyzes the effects of simultaneous use of different fuels within a fuel lattice where the zero neutron leakage assumption across the unit cell boundaries is not valid. The developed methodology expands capabilities of the collision probability theory to a supercell model that allows existence of two different fuels. The radioisotopic inventory results for different fuels obtained from the modified V:BUDS code are in excellent agreement with the benchmark problems. These accurate results may be used in general fuel cycle and transmutation studies within power reactors. Advisors/Committee Members: Schneider, Erich A. (advisor).

Subjects/Keywords: Nuclear reactors; Nuclear fuels

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APA (6^th Edition):

Matavosian, R. (2007). Complex problems arising in the collision probability theory for neutron transport. (Doctoral Dissertation). University of Texas – Austin. Retrieved from http://hdl.handle.net/2152/3789

Chicago Manual of Style (16^th Edition):

Matavosian, Robert. “Complex problems arising in the collision probability theory for neutron transport.” 2007. Doctoral Dissertation, University of Texas – Austin. Accessed April 18, 2021. http://hdl.handle.net/2152/3789.

MLA Handbook (7^th Edition):

Matavosian, Robert. “Complex problems arising in the collision probability theory for neutron transport.” 2007. Web. 18 Apr 2021.

Vancouver:

Matavosian R. Complex problems arising in the collision probability theory for neutron transport. [Internet] [Doctoral dissertation]. University of Texas – Austin; 2007. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/2152/3789.

Council of Science Editors:

Matavosian R. Complex problems arising in the collision probability theory for neutron transport. [Doctoral Dissertation]. University of Texas – Austin; 2007. Available from: http://hdl.handle.net/2152/3789

The Ohio State University

17. Henderson, Ronald R. A nuclear reactor performance and fuel management simulator for power system planning activities.

Degree: PhD, Graduate School, 1971, The Ohio State University

URL: http://rave.ohiolink.edu/etdc/view?acc_num=osu1486727662648603

Subjects/Keywords: Engineering; Nuclear fuels; Nuclear reactors

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APA (6^th Edition):

Henderson, R. R. (1971). A nuclear reactor performance and fuel management simulator for power system planning activities. (Doctoral Dissertation). The Ohio State University. Retrieved from http://rave.ohiolink.edu/etdc/view?acc_num=osu1486727662648603

Chicago Manual of Style (16^th Edition):

Henderson, Ronald R. “A nuclear reactor performance and fuel management simulator for power system planning activities.” 1971. Doctoral Dissertation, The Ohio State University. Accessed April 18, 2021. http://rave.ohiolink.edu/etdc/view?acc_num=osu1486727662648603.

MLA Handbook (7^th Edition):

Henderson, Ronald R. “A nuclear reactor performance and fuel management simulator for power system planning activities.” 1971. Web. 18 Apr 2021.

Vancouver:

Henderson RR. A nuclear reactor performance and fuel management simulator for power system planning activities. [Internet] [Doctoral dissertation]. The Ohio State University; 1971. [cited 2021 Apr 18]. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1486727662648603.

Council of Science Editors:

Henderson RR. A nuclear reactor performance and fuel management simulator for power system planning activities. [Doctoral Dissertation]. The Ohio State University; 1971. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1486727662648603

University of South Carolina

18. Yingling, Jacob A. Bison Simulation-Based Identification of Important Design Criteria for U3SI2 Fuels With Composite-Monolithic Duplex Sic Cladding.

Degree: Degree ofMSin Nuclear Engineering, Nuclear Engineering, 2019, University of South Carolina

URL: https://scholarcommons.sc.edu/etd/5513

► Accident Tolerant Fuels (ATF) require a combination of fuel and cladding which have comparable longevity characteristics to UO2 while improving resistance to radiological release… (more)

Subjects/Keywords: Nuclear Engineering; Bison; U3SI2; U3SI2 Fuels; Composite-Monolithic Duplex; SIC; Accident Tolerant Fuels

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APA (6^th Edition):

Yingling, J. A. (2019). Bison Simulation-Based Identification of Important Design Criteria for U3SI2 Fuels With Composite-Monolithic Duplex Sic Cladding. (Thesis). University of South Carolina. Retrieved from https://scholarcommons.sc.edu/etd/5513

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Yingling, Jacob A. “Bison Simulation-Based Identification of Important Design Criteria for U3SI2 Fuels With Composite-Monolithic Duplex Sic Cladding.” 2019. Thesis, University of South Carolina. Accessed April 18, 2021. https://scholarcommons.sc.edu/etd/5513.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Yingling, Jacob A. “Bison Simulation-Based Identification of Important Design Criteria for U3SI2 Fuels With Composite-Monolithic Duplex Sic Cladding.” 2019. Web. 18 Apr 2021.

Vancouver:

Yingling JA. Bison Simulation-Based Identification of Important Design Criteria for U3SI2 Fuels With Composite-Monolithic Duplex Sic Cladding. [Internet] [Thesis]. University of South Carolina; 2019. [cited 2021 Apr 18]. Available from: https://scholarcommons.sc.edu/etd/5513.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Yingling JA. Bison Simulation-Based Identification of Important Design Criteria for U3SI2 Fuels With Composite-Monolithic Duplex Sic Cladding. [Thesis]. University of South Carolina; 2019. Available from: https://scholarcommons.sc.edu/etd/5513

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Colorado School of Mines

19. Worrall, Michael Jason. Feasibility study of a small, thorium-based fission power system for space and terrestrial applications.

Degree: PhD, Physics, 2013, Colorado School of Mines

URL: http://hdl.handle.net/11124/2

► One of the current challenges facing space exploration is the creation of a power source capable of providing useful energy for the entire duration of… (more)

▼ One of the current challenges facing space exploration is the creation of a power source capable of providing useful energy for the entire duration of a mission. Historically, radioisotope batteries have been used to provide load power, but this conventional system may not be capable of sustaining continuous power for longer duration missions. To remedy this, many forays into nuclear powered spacecraft have been investigated, but no robust system for long-term power generation has been found. In this study, a novel spin on the traditional fission power system that represents a potential optimum solution is presented. By utilizing mature High Temperature Gas Reactor (HTGR) technology in conjunction with the capabilities of the thorium fuel cycle, we have created a light-weight, long-term power source capable of a continuous electric power output of up to 70kW for over 15 years. This system relies upon a combination of fissile, highly-enriched uranium dioxide and fertile thorium carbide Tri-Structural Isotropic (TRISO) fuel particles embedded in a hexagonal beryllium oxide matrix. As the primary fissile material is consumed, the fertile material breeds new fissile material leading to more steady fuel loading over the lifetime of the core. Reactor control is achieved through an innovative approach to the conventional boron carbide neutron absorber by utilizing sections of borated aluminum placed in rotating control drums within the reflector. Borated aluminum allows for much smaller boron concentrations, thus eliminating the potential for [superscript 10]B(n,alpha)[superscript 6]Li heating issues that are common in boron carbide systems. A wide range of other reactivity control systems are also investigated, such as a radially-split rotating reflector. Lastly, an extension of the design to a terrestrial based system is investigated. In this system, uranium enrichment is dropped to 20 percent in order to meet current regulations, a solid uranium-zirconium hydride fissile driver replaces the uranium dioxide TRISO particles, and the moderating material is changed from beryllium oxide to graphite. These changes result in an increased core size, but the same long-term power generation potential is achieved. Additionally, small amounts of erbium are added to the hydride matrix to further extend core lifetime. Advisors/Committee Members: Shayer, Zeev (advisor), King, Jeffrey C. (committee member), Cecil, F. Edward (committee member), Greife, Uwe (committee member), Olson, D. L. (David LeRoy) (committee member).

Subjects/Keywords: reactor; modular; thorium; space; Thorium; Nuclear fission; Nuclear reactors; Nuclear energy; Nuclear fuels; Outer space – Exploration

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APA (6^th Edition):

Worrall, M. J. (2013). Feasibility study of a small, thorium-based fission power system for space and terrestrial applications. (Doctoral Dissertation). Colorado School of Mines. Retrieved from http://hdl.handle.net/11124/2

Chicago Manual of Style (16^th Edition):

Worrall, Michael Jason. “Feasibility study of a small, thorium-based fission power system for space and terrestrial applications.” 2013. Doctoral Dissertation, Colorado School of Mines. Accessed April 18, 2021. http://hdl.handle.net/11124/2.

MLA Handbook (7^th Edition):

Worrall, Michael Jason. “Feasibility study of a small, thorium-based fission power system for space and terrestrial applications.” 2013. Web. 18 Apr 2021.

Vancouver:

Worrall MJ. Feasibility study of a small, thorium-based fission power system for space and terrestrial applications. [Internet] [Doctoral dissertation]. Colorado School of Mines; 2013. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/11124/2.

Council of Science Editors:

Worrall MJ. Feasibility study of a small, thorium-based fission power system for space and terrestrial applications. [Doctoral Dissertation]. Colorado School of Mines; 2013. Available from: http://hdl.handle.net/11124/2

20. Ivaldete da Silva Dupim. Hidretação do Zircaloy-4 para a obtenção de pó de Zr.

Degree: 2010, Universidade Federal do ABC

URL: http://tede.ufabc.edu.br/tde_busca/arquivo.php?codArquivo=149

►

Nuclear fuels based on composite materials such as dispersion of uranium compounds in a Zirconium alloy matrix are possible alternatives for advanced nuclear generation technology.… (more)

▼

Nuclear fuels based on composite materials such as dispersion of uranium compounds in a Zirconium alloy matrix are possible alternatives for advanced nuclear generation technology. A necessary step for obtaining such fuels is producing Zr alloy powder for the metal matrix composite materials. In this study we gather data to specify the process of Zircaloy hydriding as an intermediate step for obtaining Zr powder. The hydration was carried in a tubular furnace to embrittle the material and facilitate the process of comminuting. The experiments included tests in TGA/DTA system and in a tubular furnace to study the hydration process. The H2 pressures varied from 25 and 50 kPa and the temperatures varied from 20 to 670 C. The hydriding begun at a temperature around about 450 C. The samples subjected to temperatures lower than 400 C, under the conditions presented in this work, were not hydrided. The incubation time to overcome the oxidation layer lasted about 20 minutes and, after that, the bulk of the reaction occurred in about 7 minutes. The hydrogenation level was so high that the different plateau temperatures and different pressures did not affect the samples final Hydrogen to Zirconium atom ratio. The final hydride stoichiometry, at room conditions, was found to lie between 2.74 and 2.97 H to Zr atom ratio, indicating that the material was completely hydrided. X-ray diffraction analysis, showed the predominant presence of ZrHx in ZrHx-δ e ZrHx-ε phases and some peaks ZrO2. The hydride samples could be easily comminuted. The powder obtained through the milling process had an uniform appearance with a heterogeneous size distribution. These results show that to embrittle the material aiming at comminuting it, only few minutes of hydrogenation is necessary under the conditions studied.

Combustíveis nucleares baseados em materiais compósitos como a dispersão de compostos de urânio em uma matriz de ligas de zircônio são uma alternativa possível para o avanço da tecnologia de geração nucleoelétrica. Um passo necessário para a obtenção desses combustíveis é a produção do pó da liga de Zircônio para ser utilizado como material da matriz metálica. Neste trabalho, buscou-se obter dados para especificar o processo de hidretação do Zircaloy para sua transformação em pó para uso em combustíveis nucleares do tipo dispersão. A hidretação foi realizada em fornos para fragilizar o material e facilitar o processo de cominuição. Os experimentos realizados neste trabalho incluem testes em um forno tubular e em um sistema TGA/DTA para o estudo da evolução do processo de hidretação. Vários testes foram realizados através de análise termogravimétrica, aplicando pressões de H2 de 25 e 50 kPa e temperaturas que variam de 20-670 C. O processo de hidretação tem inicio a partir de 400 C aproximadamente. As amostras submetidas a temperaturas menores que 400 C sob as condições apresentadas neste trabalho não foram hidretadas. O tempo de incubação para superar a camada de oxidação durou cerca de 20 minutos e, depois disso, a maior…

Advisors/Committee Members: João Manoel Losada Moreira.

Subjects/Keywords: Combustível Nuclear; Zircaloy-4; Zircaloy-4, Zr hydride, nuclear fuels; FISICA; Hidretação

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APA (6^th Edition):

Dupim, I. d. S. (2010). Hidretação do Zircaloy-4 para a obtenção de pó de Zr. (Thesis). Universidade Federal do ABC. Retrieved from http://tede.ufabc.edu.br/tde_busca/arquivo.php?codArquivo=149

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Dupim, Ivaldete da Silva. “Hidretação do Zircaloy-4 para a obtenção de pó de Zr.” 2010. Thesis, Universidade Federal do ABC. Accessed April 18, 2021. http://tede.ufabc.edu.br/tde_busca/arquivo.php?codArquivo=149.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Dupim, Ivaldete da Silva. “Hidretação do Zircaloy-4 para a obtenção de pó de Zr.” 2010. Web. 18 Apr 2021.

Vancouver:

Dupim IdS. Hidretação do Zircaloy-4 para a obtenção de pó de Zr. [Internet] [Thesis]. Universidade Federal do ABC; 2010. [cited 2021 Apr 18]. Available from: http://tede.ufabc.edu.br/tde_busca/arquivo.php?codArquivo=149.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Dupim IdS. Hidretação do Zircaloy-4 para a obtenção de pó de Zr. [Thesis]. Universidade Federal do ABC; 2010. Available from: http://tede.ufabc.edu.br/tde_busca/arquivo.php?codArquivo=149

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Iowa State University

21. Madsen, Larry Charles. Nuclear fuel cycle cost analysis using parametric variation.

Degree: 1966, Iowa State University

URL: https://lib.dr.iastate.edu/rtd/18487

Subjects/Keywords: Nuclear fuels – Costs; Nuclear Engineering; Engineering

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APA (6^th Edition):

Madsen, L. C. (1966). Nuclear fuel cycle cost analysis using parametric variation. (Thesis). Iowa State University. Retrieved from https://lib.dr.iastate.edu/rtd/18487

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Madsen, Larry Charles. “Nuclear fuel cycle cost analysis using parametric variation.” 1966. Thesis, Iowa State University. Accessed April 18, 2021. https://lib.dr.iastate.edu/rtd/18487.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Madsen, Larry Charles. “Nuclear fuel cycle cost analysis using parametric variation.” 1966. Web. 18 Apr 2021.

Vancouver:

Madsen LC. Nuclear fuel cycle cost analysis using parametric variation. [Internet] [Thesis]. Iowa State University; 1966. [cited 2021 Apr 18]. Available from: https://lib.dr.iastate.edu/rtd/18487.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Madsen LC. Nuclear fuel cycle cost analysis using parametric variation. [Thesis]. Iowa State University; 1966. Available from: https://lib.dr.iastate.edu/rtd/18487

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Colorado School of Mines

22. Davis, Brian Campbell. Fracture strength of the SiC layer in TRISO coated fuel particles.

Degree: MS(M.S.), Mechanical Engineering, 2013, Colorado School of Mines

URL: http://hdl.handle.net/11124/78777

► The SiC layer in the TRISO coating of next generation nuclear fuel particles is a structural element which ensures the containment of fissile products, thus… (more)

Subjects/Keywords: TRISO; NGNP; Nuclear fuels – Transportation; Particles (Nuclear physics); Coating processes; Silicon carbide

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APA (6^th Edition):

Davis, B. C. (2013). Fracture strength of the SiC layer in TRISO coated fuel particles. (Masters Thesis). Colorado School of Mines. Retrieved from http://hdl.handle.net/11124/78777

Chicago Manual of Style (16^th Edition):

Davis, Brian Campbell. “Fracture strength of the SiC layer in TRISO coated fuel particles.” 2013. Masters Thesis, Colorado School of Mines. Accessed April 18, 2021. http://hdl.handle.net/11124/78777.

MLA Handbook (7^th Edition):

Davis, Brian Campbell. “Fracture strength of the SiC layer in TRISO coated fuel particles.” 2013. Web. 18 Apr 2021.

Vancouver:

Davis BC. Fracture strength of the SiC layer in TRISO coated fuel particles. [Internet] [Masters thesis]. Colorado School of Mines; 2013. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/11124/78777.

Council of Science Editors:

Davis BC. Fracture strength of the SiC layer in TRISO coated fuel particles. [Masters Thesis]. Colorado School of Mines; 2013. Available from: http://hdl.handle.net/11124/78777

23. LEVY, DENISE S. Contribuição para informatização dos programas de proteção radiológica para instalações radiativas.

Degree: 2012, Brazil

URL: http://repositorio.ipen.br:8080/xmlui/handle/123456789/10144

►

Made available in DSpace on 2014-10-09T12:35:14Z (GMT). No. of bitstreams: 0

Made available in DSpace on 2014-10-09T14:09:36Z (GMT). No. of bitstreams: 0

Para elaborar um… (more)

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Made available in DSpace on 2014-10-09T12:35:14Z (GMT). No. of bitstreams: 0

Made available in DSpace on 2014-10-09T14:09:36Z (GMT). No. of bitstreams: 0

Para elaborar um programa de proteção radiológica, as instalações radiativas brasileiras devem considerar normas, diretrizes e recomendações nacionais e internacionais que encontram-se em documentos de diferentes organizações publicados nas últimas décadas: Comissão Internacional de Proteção Radiológica (CIPR), Organismo Internacional de Energia Atômica (OIEA) e Comissão Nacional de Energia Nuclear (CNEN). Visando a proporcionar a essas instalações o acesso às informações pertinentes de forma rápida, integrada e eficiente, este projeto propõe informatizar e disponibilizar em um só documento os programas de otimização da proteção radiológica unificados, inter-relacionados e em português, fornecendo ao público usuário um veículo completo para fins de pesquisa, consulta e informação. A partir do discernimento do que deve conter cada programa e seu real dimensionamento, foi trabalhado o inter-relacionamento das informações de maneira a satisfazer as normas e recomendações nacionais e internacionais. O projeto inclui conceitos, definições e teoria necessários, além da pesquisa detalhada do conteúdo do programa de otimização, das técnicas de ajuda para tomada de decisão, das doses de radiação e detrimento e das informações relacionadas aos custos de proteção. O conteúdo permite responder a todas as questões que devem ser colocadas na elaboração de um programa de otimização de forma a possibilitar montagem do plano de Proteção Radiológica conforme a situação específica do usuário. Para a informatização dos programas de otimização foram estudadas as possibilidades de acesso à Tecnologia da Informação e Comunicação nas empresas brasileiras, possibilitando identificar o perfil de utilização do sistema e definir a estrutura funcional adequada para a criação das melhores interfaces de ferramentas e recursos, bem como de um projeto de navegabilidade eficaz facilitando a busca de informações. O poder de processamento dos servidores aliado à tecnologia dos bancos de dados relacionais permite correlacionar informações advindas de diferentes fontes, possibilitando consultas complexas com tempo de resposta reduzido. O sistema segue o padrão WEB 2.0, que possibilita a estrutura organizacional necessária para a adequada informatização da proteção radiológica e considera os corretos critérios de indexação da informação para garantir seu reconhecimento pelos motores de busca da internet. O projeto conta com a combinação de várias tecnologias, potencializando os recursos disponíveis em cada uma delas para alcançar os objetivos propostos. Este trabalho experimental lança um cerne inicial para a informatização dos programas de proteção radiológica, informatizando inicialmente os programas de otimização. A investigação do perfil de utilização durante um período de cinco meses possibilitou o levantamento de dados importantes que apontam novas possibilidades para o desenvolvimento da…

Advisors/Committee Members: Gian Maria Agostino Angelo Sordi.

Subjects/Keywords: radiation protection; nuclear facilities; nuclear fuels; fuel cycle; computerized control systems; computer codes; program management

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APA (6^th Edition):

LEVY, D. S. (2012). Contribuição para informatização dos programas de proteção radiológica para instalações radiativas. (Masters Thesis). Brazil. Retrieved from http://repositorio.ipen.br:8080/xmlui/handle/123456789/10144

Chicago Manual of Style (16^th Edition):

LEVY, DENISE S. “Contribuição para informatização dos programas de proteção radiológica para instalações radiativas.” 2012. Masters Thesis, Brazil. Accessed April 18, 2021. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10144.

MLA Handbook (7^th Edition):

LEVY, DENISE S. “Contribuição para informatização dos programas de proteção radiológica para instalações radiativas.” 2012. Web. 18 Apr 2021.

Vancouver:

LEVY DS. Contribuição para informatização dos programas de proteção radiológica para instalações radiativas. [Internet] [Masters thesis]. Brazil; 2012. [cited 2021 Apr 18]. Available from: http://repositorio.ipen.br:8080/xmlui/handle/123456789/10144.

Council of Science Editors:

LEVY DS. Contribuição para informatização dos programas de proteção radiológica para instalações radiativas. [Masters Thesis]. Brazil; 2012. Available from: http://repositorio.ipen.br:8080/xmlui/handle/123456789/10144

Universidade Estadual de Campinas

24. Bianchessi, Luis Francisco, 1990-. Estudos relativos à síntese de biodiesel e avaliação de blendas diesel-biodiesel.

Degree: Instituto de Química; Programa de Pós-Graduação em Química, 2019, Universidade Estadual de Campinas

URL: http://repositorio.unicamp.br/jspui/handle/REPOSIP/335866

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Orientador: Matthieu Tubino

Dissertação (mestrado) - Universidade Estadual de Campinas, Instituto de Química

Made available in DSpace on 2020-01-30T18:29:58Z (GMT). No. of bitstreams: 1 Bianchessi_LuisFrancisco_M.pdf:… (more)

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Orientador: Matthieu Tubino

Dissertação (mestrado) - Universidade Estadual de Campinas, Instituto de Química

Made available in DSpace on 2020-01-30T18:29:58Z (GMT). No. of bitstreams: 1 Bianchessi_LuisFrancisco_M.pdf: 1718069 bytes, checksum: 9a65e82d5e185302a236fe2212f89d06 (MD5) Previous issue date: 2019

Resumo: O biodiesel é o principal produto proveniente da reação de transesterificação de óleos e gorduras. Dentre os problemas relacionados com a síntese e a qualidade deste biocombustível estão; (i) a eficiência do processo de transesterificação em fornecer produto de boa qualidade; (ii) distribuição de produto de má qualidade ou adulterado pela adição de, por exemplo, óleo vegetal. Neste cenário, é importante desenvolver estudos que visem a otimização do processo de esterificação para obtenção de biodiesel e elaborar métodos analíticos simples e confiáveis para a quantificação dos ésteres obtidos e presentes no biodiesel e, também, para a identificação e quantificação de possíveis adulterantes. Neste trabalho, foram desenvolvidas investigações com o objetivo de aperfeiçoar a síntese de biodiesel etílico: (i) empregando menores proporções de etanol do que aquelas relatadas na literatura; (ii) avaliando a influência da adição de óleo diesel como cossolvente na reação de transesterificação. Na proporção molar de 6:1 (etanol:triacilglicerídeo) a 30 ºC chegou-se a conversões próximas de 95%, o que pode ser considerado como muito satisfatório. Como fato adicional, observou-se que a adição de óleo diesel ao meio reacional retardou e diminuiu a formação de sabões após a síntese e purificação. No que se refere à investigação visando desenvolver novo método analítico, com o objetivo de avaliar a eficiência da reação de transesterificação, a técnica de RMN 1H mostrou-se uma ferramenta muito adequada para o monitoramento da conversão dos triacilgliceróis em ésteres. Ela fornece resultados rápidos, confiáveis e não há necessidade de pré-tratamento das amostras. Através dessa técnica, além de ser possível quantificar os ésteres presentes no biodiesel sintetizado pode-se, também, identificar e determinar a quantidade de óleos vegetais ou gorduras não esterificados, eventualmente presentes como resíduos do processo de transesterificação, ou que foram intencionalmente adicionados como adulterantes nos biodieseis metílico e etílico e em suas blendas com diesel mineral. Amostras comerciais foram analisadas e tiveram resultados estatisticamente equivalentes àqueles obtidos pelas normas de referência

Biodiesel is the main product from the transesterification reaction of oils and fats. Among the problems related to the synthesis and quality of this biofuel, it is possible to cite (i) the efficiency of the transesterification process in supplying a good quality product; (ii) distribution of poor quality or adulterated product by the addition of, for example, unesterified vegetable oil. In this scenario, it is important to develop studies aimed at optimizing the esterification process to obtain biodiesel and elaborate…

Advisors/Committee Members: UNIVERSIDADE ESTADUAL DE CAMPINAS, Tubino, Matthieu, 1947-, Silveira, Eva Lúcia Cardoso, Corbi, Pedro Paulo.

Subjects/Keywords: Biodiesel; Síntese; Blendas; Ressonância magnética nuclear; Biodiesel fuels; Synthesis; Blends; Nuclear magnetic resonance

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Bianchessi, Luis Francisco, 1. (2019). Estudos relativos à síntese de biodiesel e avaliação de blendas diesel-biodiesel. (Masters Thesis). Universidade Estadual de Campinas. Retrieved from http://repositorio.unicamp.br/jspui/handle/REPOSIP/335866

Chicago Manual of Style (16^th Edition):

Bianchessi, Luis Francisco, 1990-. “Estudos relativos à síntese de biodiesel e avaliação de blendas diesel-biodiesel.” 2019. Masters Thesis, Universidade Estadual de Campinas. Accessed April 18, 2021. http://repositorio.unicamp.br/jspui/handle/REPOSIP/335866.

MLA Handbook (7^th Edition):

Bianchessi, Luis Francisco, 1990-. “Estudos relativos à síntese de biodiesel e avaliação de blendas diesel-biodiesel.” 2019. Web. 18 Apr 2021.

Vancouver:

Bianchessi, Luis Francisco 1. Estudos relativos à síntese de biodiesel e avaliação de blendas diesel-biodiesel. [Internet] [Masters thesis]. Universidade Estadual de Campinas; 2019. [cited 2021 Apr 18]. Available from: http://repositorio.unicamp.br/jspui/handle/REPOSIP/335866.

Council of Science Editors:

Bianchessi, Luis Francisco 1. Estudos relativos à síntese de biodiesel e avaliação de blendas diesel-biodiesel. [Masters Thesis]. Universidade Estadual de Campinas; 2019. Available from: http://repositorio.unicamp.br/jspui/handle/REPOSIP/335866

University of Bath

25. Ayling, Joanna R. Evaluation of a technique for the remote measurement of surface temperature distributions in laser heated samples.

Degree: PhD, 1996, University of Bath

URL: https://researchportal.bath.ac.uk/en/studentthesis/evaluation-of-a-technique-for-the-remote-measurement-of-surface-temperature-distributions-in-laser-heated-samples(a9095f76-bfcb-486d-a5be-9eba36ed654c).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.320437

Subjects/Keywords: 621.484; Nuclear fuels; Thermal behaviour

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Ayling, J. R. (1996). Evaluation of a technique for the remote measurement of surface temperature distributions in laser heated samples. (Doctoral Dissertation). University of Bath. Retrieved from https://researchportal.bath.ac.uk/en/studentthesis/evaluation-of-a-technique-for-the-remote-measurement-of-surface-temperature-distributions-in-laser-heated-samples(a9095f76-bfcb-486d-a5be-9eba36ed654c).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.320437

Chicago Manual of Style (16^th Edition):

Ayling, Joanna R. “Evaluation of a technique for the remote measurement of surface temperature distributions in laser heated samples.” 1996. Doctoral Dissertation, University of Bath. Accessed April 18, 2021. https://researchportal.bath.ac.uk/en/studentthesis/evaluation-of-a-technique-for-the-remote-measurement-of-surface-temperature-distributions-in-laser-heated-samples(a9095f76-bfcb-486d-a5be-9eba36ed654c).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.320437.

MLA Handbook (7^th Edition):

Ayling, Joanna R. “Evaluation of a technique for the remote measurement of surface temperature distributions in laser heated samples.” 1996. Web. 18 Apr 2021.

Vancouver:

Ayling JR. Evaluation of a technique for the remote measurement of surface temperature distributions in laser heated samples. [Internet] [Doctoral dissertation]. University of Bath; 1996. [cited 2021 Apr 18]. Available from: https://researchportal.bath.ac.uk/en/studentthesis/evaluation-of-a-technique-for-the-remote-measurement-of-surface-temperature-distributions-in-laser-heated-samples(a9095f76-bfcb-486d-a5be-9eba36ed654c).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.320437.

Council of Science Editors:

Ayling JR. Evaluation of a technique for the remote measurement of surface temperature distributions in laser heated samples. [Doctoral Dissertation]. University of Bath; 1996. Available from: https://researchportal.bath.ac.uk/en/studentthesis/evaluation-of-a-technique-for-the-remote-measurement-of-surface-temperature-distributions-in-laser-heated-samples(a9095f76-bfcb-486d-a5be-9eba36ed654c).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.320437

University of Arizona

26. Friedlander, Michael Arthur. AN EXPERIMENTAL ANALYSIS OF THE FISSILE CONTENT OF THE TRIGA CORE BY TRANSFER FUNCTION MEASUREMENT.

Degree: 1984, University of Arizona

URL: http://hdl.handle.net/10150/275098

Subjects/Keywords: Nuclear fuels – Analysis.; Transfer functions.

Record Details Similar Records Cite « Share

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Friedlander, M. A. (1984). AN EXPERIMENTAL ANALYSIS OF THE FISSILE CONTENT OF THE TRIGA CORE BY TRANSFER FUNCTION MEASUREMENT. (Masters Thesis). University of Arizona. Retrieved from http://hdl.handle.net/10150/275098

Chicago Manual of Style (16^th Edition):

Friedlander, Michael Arthur. “AN EXPERIMENTAL ANALYSIS OF THE FISSILE CONTENT OF THE TRIGA CORE BY TRANSFER FUNCTION MEASUREMENT. ” 1984. Masters Thesis, University of Arizona. Accessed April 18, 2021. http://hdl.handle.net/10150/275098.

MLA Handbook (7^th Edition):

Friedlander, Michael Arthur. “AN EXPERIMENTAL ANALYSIS OF THE FISSILE CONTENT OF THE TRIGA CORE BY TRANSFER FUNCTION MEASUREMENT. ” 1984. Web. 18 Apr 2021.

Vancouver:

Friedlander MA. AN EXPERIMENTAL ANALYSIS OF THE FISSILE CONTENT OF THE TRIGA CORE BY TRANSFER FUNCTION MEASUREMENT. [Internet] [Masters thesis]. University of Arizona; 1984. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/10150/275098.

Council of Science Editors:

Friedlander MA. AN EXPERIMENTAL ANALYSIS OF THE FISSILE CONTENT OF THE TRIGA CORE BY TRANSFER FUNCTION MEASUREMENT. [Masters Thesis]. University of Arizona; 1984. Available from: http://hdl.handle.net/10150/275098

University of Arizona

27. Schwarzblat, Morris, 1945-. AN EQUATION OF STATE FOR MIXED-OXIDE FAST REACTOR FUELS .

Degree: 1974, University of Arizona

URL: http://hdl.handle.net/10150/288330

Subjects/Keywords: Equations of state.; Nuclear fuels.

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Schwarzblat, Morris, 1. (1974). AN EQUATION OF STATE FOR MIXED-OXIDE FAST REACTOR FUELS . (Doctoral Dissertation). University of Arizona. Retrieved from http://hdl.handle.net/10150/288330

Chicago Manual of Style (16^th Edition):

Schwarzblat, Morris, 1945-. “AN EQUATION OF STATE FOR MIXED-OXIDE FAST REACTOR FUELS .” 1974. Doctoral Dissertation, University of Arizona. Accessed April 18, 2021. http://hdl.handle.net/10150/288330.

MLA Handbook (7^th Edition):

Schwarzblat, Morris, 1945-. “AN EQUATION OF STATE FOR MIXED-OXIDE FAST REACTOR FUELS .” 1974. Web. 18 Apr 2021.

Vancouver:

Schwarzblat, Morris 1. AN EQUATION OF STATE FOR MIXED-OXIDE FAST REACTOR FUELS . [Internet] [Doctoral dissertation]. University of Arizona; 1974. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/10150/288330.

Council of Science Editors:

Schwarzblat, Morris 1. AN EQUATION OF STATE FOR MIXED-OXIDE FAST REACTOR FUELS . [Doctoral Dissertation]. University of Arizona; 1974. Available from: http://hdl.handle.net/10150/288330

Kansas State University

28. Johnson, William Joseph. Techniques for calculation of nuclear fuel costs.

Degree: 1976, Kansas State University

URL: http://hdl.handle.net/2097/11371

Subjects/Keywords: Nuclear fuels – Costs – Mathematics

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Johnson, W. J. (1976). Techniques for calculation of nuclear fuel costs. (Thesis). Kansas State University. Retrieved from http://hdl.handle.net/2097/11371

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Johnson, William Joseph. “Techniques for calculation of nuclear fuel costs.” 1976. Thesis, Kansas State University. Accessed April 18, 2021. http://hdl.handle.net/2097/11371.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Johnson, William Joseph. “Techniques for calculation of nuclear fuel costs.” 1976. Web. 18 Apr 2021.

Vancouver:

Johnson WJ. Techniques for calculation of nuclear fuel costs. [Internet] [Thesis]. Kansas State University; 1976. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/2097/11371.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Johnson WJ. Techniques for calculation of nuclear fuel costs. [Thesis]. Kansas State University; 1976. Available from: http://hdl.handle.net/2097/11371

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Kansas State University

29. Chen, Yuh-Fuu. A non-equilibrium LWR fuel cycle model.

Degree: 1973, Kansas State University

URL: http://hdl.handle.net/2097/10687

Subjects/Keywords: Nuclear fuels – Mathematical models

Record Details Similar Records Cite « Share

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Chen, Y. (1973). A non-equilibrium LWR fuel cycle model. (Thesis). Kansas State University. Retrieved from http://hdl.handle.net/2097/10687

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16^th Edition):

Chen, Yuh-Fuu. “A non-equilibrium LWR fuel cycle model.” 1973. Thesis, Kansas State University. Accessed April 18, 2021. http://hdl.handle.net/2097/10687.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

MLA Handbook (7^th Edition):

Chen, Yuh-Fuu. “A non-equilibrium LWR fuel cycle model.” 1973. Web. 18 Apr 2021.

Vancouver:

Chen Y. A non-equilibrium LWR fuel cycle model. [Internet] [Thesis]. Kansas State University; 1973. [cited 2021 Apr 18]. Available from: http://hdl.handle.net/2097/10687.

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

Chen Y. A non-equilibrium LWR fuel cycle model. [Thesis]. Kansas State University; 1973. Available from: http://hdl.handle.net/2097/10687

Note: this citation may be lacking information needed for this citation format:
Not specified: Masters Thesis or Doctoral Dissertation

University of New South Wales

30. Alfredson, Peter George. A study of the variables in the coating of fuel particles in a fluidised bed.

Degree: Engineering, School of Nuclear Engineering, 1964, University of New South Wales

URL: http://handle.unsw.edu.au/1959.4/56753 ; https://unsworks.unsw.edu.au/fapi/datastream/unsworks:41335/SOURCE02?view=true

Subjects/Keywords: Nuclear fuels; Thesis Digitisation Program

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6^th Edition):

Alfredson, P. G. (1964). A study of the variables in the coating of fuel particles in a fluidised bed. (Masters Thesis). University of New South Wales. Retrieved from http://handle.unsw.edu.au/1959.4/56753 ; https://unsworks.unsw.edu.au/fapi/datastream/unsworks:41335/SOURCE02?view=true

Chicago Manual of Style (16^th Edition):

Alfredson, Peter George. “A study of the variables in the coating of fuel particles in a fluidised bed.” 1964. Masters Thesis, University of New South Wales. Accessed April 18, 2021. http://handle.unsw.edu.au/1959.4/56753 ; https://unsworks.unsw.edu.au/fapi/datastream/unsworks:41335/SOURCE02?view=true.

MLA Handbook (7^th Edition):

Alfredson, Peter George. “A study of the variables in the coating of fuel particles in a fluidised bed.” 1964. Web. 18 Apr 2021.

Vancouver:

Alfredson PG. A study of the variables in the coating of fuel particles in a fluidised bed. [Internet] [Masters thesis]. University of New South Wales; 1964. [cited 2021 Apr 18]. Available from: http://handle.unsw.edu.au/1959.4/56753 ; https://unsworks.unsw.edu.au/fapi/datastream/unsworks:41335/SOURCE02?view=true.

Council of Science Editors:

Alfredson PG. A study of the variables in the coating of fuel particles in a fluidised bed. [Masters Thesis]. University of New South Wales; 1964. Available from: http://handle.unsw.edu.au/1959.4/56753 ; https://unsworks.unsw.edu.au/fapi/datastream/unsworks:41335/SOURCE02?view=true

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