subject:(nuclear fuel)

University of California – Berkeley

1. Terrani, Kurt Amir. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors.

Degree: Nuclear Engineering, 2010, University of California – Berkeley

URL: http://www.escholarship.org/uc/item/48d78064

► This dissertation intends to examine basic materials properties, identify optimized fabrication techniques, model behavior under relevant environments, and experimentally quantify kinetic phenomena associated with hydride… (more)

▼ This dissertation intends to examine basic materials properties, identify optimized fabrication techniques, model behavior under relevant environments, and experimentally quantify kinetic phenomena associated with hydride nuclear fuels. Hydride fuels have been examined extensively for application in light water reactors (LWR) from the neutronics and thermal hydraulic standpoints, the benefits of this fuel have been underscored through such studies. This manuscript provides the background for understanding materials aspects of hydride fuel incorporation in LWR environments. The proposed LWR hydride fuel concept consists of uranium-zirconium hydride pellets clad in Zircaloy and bonded with a lead-bismuth alloy. The fuel material consists of metallic uranium particles dispersed in a zirconium hydride matrix, although thorium and/or other minor actinide hydride matrices could be utilized. The eutectic lead-bismuth alloy is liquid during reactor operating temperatures and replaces the conventionally-used helium gas in the fuel-cladding gap, thereby providing a thermal conductivity increase of two orders of magnitude. Initially uranium-thorium-zirconium hydrides were fabricated and extensively characterized. This provided detailed insight into fuel properties and the influence of fabrication methodology. A modeling approach was undertaken to examine hydride fuel behavior under steady-state and transient-power conditions in a typical LWR. This study outlined the operating parameters and fuel-response characteristics under various reactor operating conditions that support the feasibility of hydride fuel incorporation into LWRs. The kinetics of hydrogen release from the fuel, associated with one of the most severe accident scenarios, was investigated in detail. Mechanisms were identified for hydrogen desorption from and adsorption on zirconium hydride and the rates associated with each process were quantified. Hydrogen diffusivity in the thorium-zirconium hydride matrix, which is one of the critical parameters affecting fabrication and in-reactor fuel behavior, was experimentally determined by the means of incoherent quasielastic neutron scattering. Finally experiments were conducted to examine compatibility of hydride fuel with Zircaloy cladding when bonded by liquid-metal. A thin oxide grown on the surface of the cladding coupled with liquid metal was tentatively identified as adequate to limit hydrogen transport form the fuel to the cladding. Recognizing the necessity of a shift from laboratory scale experiments to more relevant fuel-operating environments, an irradiation experiment was conceived to examine the liquid-metal-bonded LWR hydride fuel concept.

Subjects/Keywords: Nuclear Engineering; Hydride Fuel; Nuclear Fuel

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

APA (6^th Edition):

Terrani, K. A. (2010). Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors. (Thesis). University of California – Berkeley. Retrieved from http://www.escholarship.org/uc/item/48d78064

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):

Terrani, Kurt Amir. “Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors.” 2010. Thesis, University of California – Berkeley. Accessed March 03, 2021. http://www.escholarship.org/uc/item/48d78064.

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

MLA Handbook (7^th Edition):

Terrani, Kurt Amir. “Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors.” 2010. Web. 03 Mar 2021.

Vancouver:

Terrani KA. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors. [Internet] [Thesis]. University of California – Berkeley; 2010. [cited 2021 Mar 03]. Available from: http://www.escholarship.org/uc/item/48d78064.

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

Council of Science Editors:

Terrani KA. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors. [Thesis]. University of California – Berkeley; 2010. Available from: http://www.escholarship.org/uc/item/48d78064

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

University of New Mexico

2. Edelmann, Paul. Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel.

Degree: Nuclear Engineering, 2013, University of New Mexico

URL: http://hdl.handle.net/1928/23107

► There have been numerous attempts to model fast reactor fuel behavior in the last 40 years. The US currently does not have a fully reliable… (more)

Subjects/Keywords: nuclear; fuel; diffusion

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

APA (6^th Edition):

Edelmann, P. (2013). Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel. (Doctoral Dissertation). University of New Mexico. Retrieved from http://hdl.handle.net/1928/23107

Chicago Manual of Style (16^th Edition):

Edelmann, Paul. “Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel.” 2013. Doctoral Dissertation, University of New Mexico. Accessed March 03, 2021. http://hdl.handle.net/1928/23107.

MLA Handbook (7^th Edition):

Edelmann, Paul. “Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel.” 2013. Web. 03 Mar 2021.

Vancouver:

Edelmann P. Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel. [Internet] [Doctoral dissertation]. University of New Mexico; 2013. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/1928/23107.

Council of Science Editors:

Edelmann P. Modeling and Analysis of Actinide Diffusion Behavior in Irradiated Metal Fuel. [Doctoral Dissertation]. University of New Mexico; 2013. Available from: http://hdl.handle.net/1928/23107

University of Manchester

3. Tucker, Kate Louise. Heavy Metal Extraction Using Advanced Liquid - Liquid Style Partitioning Systems.

Degree: 2015, University of Manchester

URL: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:274686

► Understanding the behaviour of heavy metals involved in the nuclear fuel cycle is of paramount importance to the reprocessing and storage of spent nuclear fuel.… (more)

▼ Understanding the behaviour of heavy metals involved in the nuclear fuel cycle is of paramount importance to the reprocessing and storage of spent nuclear fuel. These studies have attempted to obtain a greater understanding of the fundamental chemistry of these systems, by investigating extraction performance and speciation in current (PUREX) and proposed (GANEX) extraction processes. Various complexes have been shown to exist in the post-extracted organic fraction of the systems analysed. For Zr(IV), U(VI) and Np(VI) separated from aqueous nitric and hydrochloric using TBP, the complexes [Zr(NO3/Cl)4(TBP)4], [UO2(NO3/Cl)2(TBP)2] and [NpO2(NO3/Cl)2(TBP)2] formed, respectively. For Zr(IV) separated from aqueous mixtures of HNO3 and HCl at equal concentration, a preference was shown to [Zr(Cl)4(TBP)4] over the analogous nitrate complex. For U(VI) separated from aqueous mixtures of HNO3 and HCl, a preference was shown to [UO2(Cl)2(TBP)2], even at high aqueous nitrate concentrations. NMR data for Pu(IV) separated from aqueous HNO3, HCl and mixtures of both, using TBP were presented, where possible complexation was observed. It is thought that [Pu(NO3)4(TBP)4] or [PuCl4(TBP)4] species existed within the organic fraction for Pu(IV) separated from aqueous HNO3 and HCl, respectively. These systems showed high distribution ratios where an increase was observed with increasing aqueous acid concentration overall. Distribution ratio data were presented for the lanthanide series separated from aqueous nitric acid, using the proposed GANEX solvent system(s). The lanthanides analysed showed an increase in distribution ratio with increasing aqueous nitric acid concentration and with increasing TODGA concentration in the organic fraction. Heavier lanthanides were observed to give higher distribution ratios overall. The best distribution ratios were observed for lanthanides separated using 0.2 M TODGA with 1-octanol (5 % by volume) over the nitric acid concentration range analysed. For lanthanides separated using 0.5 M DMDOHEMA, an optimum distribution ratio was observed at around 6 M aqueous nitric acid concentration. The distribution ratio data for lanthanides separated from a range of DMDOHEMA concentrations, were observed to increase with increasing organic DMDOHEMA concentration. The distribution ratios observed for isotopes of Np, Am, Eu and Pu separated using 0.2 M TODGA, increased with increasing aqueous nitric acid concentration. The same trend was observed for the aforementioned isotopes separated using 0.5 M DMDOHEMA. However, pertechnetate separated using 0.2 M TODGA from aqueous nitric acid, showed a decrease in the distribution ratios observed over the acid concentration range analysed. This was contrary to pertechnetate separated from aqueous nitric acid using 0.5 M DMDOHEMA, where a small increase in distribution ratio was observed over the concentration range analysed. For Np(VI) separated from some proposed GANEX solvents, the 0.2 M TODGA/0.5 DMDOHEMA combination gave the best distribution of neptunium into the… Advisors/Committee Members: SHARRAD, CLINT CA, Heath, Sarah, Sharrad, Clint.

Subjects/Keywords: Nuclear fuel reprocessing

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

APA (6^th Edition):

Tucker, K. L. (2015). Heavy Metal Extraction Using Advanced Liquid - Liquid Style Partitioning Systems. (Doctoral Dissertation). University of Manchester. Retrieved from http://www.manchester.ac.uk/escholar/uk-ac-man-scw:274686

Chicago Manual of Style (16^th Edition):

Tucker, Kate Louise. “Heavy Metal Extraction Using Advanced Liquid - Liquid Style Partitioning Systems.” 2015. Doctoral Dissertation, University of Manchester. Accessed March 03, 2021. http://www.manchester.ac.uk/escholar/uk-ac-man-scw:274686.

MLA Handbook (7^th Edition):

Tucker, Kate Louise. “Heavy Metal Extraction Using Advanced Liquid - Liquid Style Partitioning Systems.” 2015. Web. 03 Mar 2021.

Vancouver:

Tucker KL. Heavy Metal Extraction Using Advanced Liquid - Liquid Style Partitioning Systems. [Internet] [Doctoral dissertation]. University of Manchester; 2015. [cited 2021 Mar 03]. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:274686.

Council of Science Editors:

Tucker KL. Heavy Metal Extraction Using Advanced Liquid - Liquid Style Partitioning Systems. [Doctoral Dissertation]. University of Manchester; 2015. Available from: http://www.manchester.ac.uk/escholar/uk-ac-man-scw:274686

University of Texas – Austin

4. -3703-6555. Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling.

Degree: PhD, Mechanical Engineering, 2015, University of Texas – Austin

URL: http://hdl.handle.net/2152/45763

► The large number of reactor designs and concepts in existence open up a vast array of nuclear fuel cycle strategies. u. These different reactor types… (more)

▼ The large number of reactor designs and concepts in existence open up a vast array of nuclear fuel cycle strategies. u. These different reactor types require unique supporting systems from raw material extraction and handling to waste management. Any system designed to model nuclear energy should therefore have methods that are capability of representing a large number of unique fuel cycles. This work examines a user interface designed to generalize the design of nuclear fuel cycles. This software, known as CycIC, allows users to interact graphically with a fuel cycle simulator (Cyclus). In this work, the capabilities of CycIC were improved through two rounds of rigorous user experience testing. These tests were used as a basis for implementing improvements to the software. Two views inside the software were improved to allow for users to interact with the software more intuitively, and features that provide help to the users were added to improve understanding of fuel cycles and Cyclus. Additionally, this work expands the capabilities of a reactor modeling software (known as Bright-lite) which uses the fluence based neutron balance approach to determine burnup, criticality, and transmutation matrixes for nuclear reactors to augment its modeling of the broadest range of fuel cycle strategies. Specifically, a multi-dimensional interpolation method was implemented to enable reactors to be characterized by sets of cross section libraries which potentially depend on a large number of reactor characteristics. The accuracy of this interpolation method is demonstrated for a number of parameters for light water reactors, and techniques for using this interpolation method to automatically generate reactor libraries for Bright-lite are demonstrated. This research also generalizes the ability of the Bright-lite to blend multiple streams of nuclear fuel while still maintaining constraints. This system is demonstrated for continuous recycle nuclear fuel cycles utilizing light water and fast spectrum reactors. The results show that Bright-lite is capable of blending fuel to reach several targets using up to three different input streams. Advisors/Committee Members: Schneider, Erich A. (advisor), Wilson, Paul (committee member), Livnat, Yarden (committee member), Landsberger, Sheldon (committee member), Biegalski, Steven (committee member).

Subjects/Keywords: Nuclear fuel cycle; Reactor modeling; Fuel blending

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

APA (6^th Edition):

-3703-6555. (2015). Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling. (Doctoral Dissertation). University of Texas – Austin. Retrieved from http://hdl.handle.net/2152/45763

Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete

Chicago Manual of Style (16^th Edition):

-3703-6555. “Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling.” 2015. Doctoral Dissertation, University of Texas – Austin. Accessed March 03, 2021. http://hdl.handle.net/2152/45763.

Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete

MLA Handbook (7^th Edition):

-3703-6555. “Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling.” 2015. Web. 03 Mar 2021.

Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete

Vancouver:

-3703-6555. Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling. [Internet] [Doctoral dissertation]. University of Texas – Austin; 2015. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/2152/45763.

Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete

Council of Science Editors:

-3703-6555. Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling. [Doctoral Dissertation]. University of Texas – Austin; 2015. Available from: http://hdl.handle.net/2152/45763

Note: this citation may be lacking information needed for this citation format:
Author name may be incomplete

University of Manchester

5. Tucker, Kate Louise. Heavy metal extraction using advanced liquid-liquid style partitioning systems.

Degree: PhD, 2015, University of Manchester

URL: https://www.research.manchester.ac.uk/portal/en/theses/heavy-metal-extraction-using-advanced-liquid – liquid-style-partitioning-systems(6a238cb4-94cf-4fa8-bffe-d1a1b70adaa6).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.677739

► Understanding the behaviour of heavy metals involved in the nuclear fuel cycle is of paramount importance to the reprocessing and storage of spent nuclear fuel.… (more)

▼ Understanding the behaviour of heavy metals involved in the nuclear fuel cycle is of paramount importance to the reprocessing and storage of spent nuclear fuel. These studies have attempted to obtain a greater understanding of the fundamental chemistry of these systems, by investigating extraction performance and speciation in current (PUREX) and proposed (GANEX) extraction processes. Various complexes have been shown to exist in the post-extracted organic fraction of the systems analysed. For Zr(IV), U(VI) and Np(VI) separated from aqueous nitric and hydrochloric using TBP, the complexes [Zr(NO3/Cl)4(TBP)4], [UO2(NO3/Cl)2(TBP)2] and [NpO2(NO3/Cl)2(TBP)2] formed, respectively. For Zr(IV) separated from aqueous mixtures of HNO3 and HCl at equal concentration, a preference was shown to [Zr(Cl)4(TBP)4] over the analogous nitrate complex. For U(VI) separated from aqueous mixtures of HNO3 and HCl, a preference was shown to [UO2(Cl)2(TBP)2], even at high aqueous nitrate concentrations. NMR data for Pu(IV) separated from aqueous HNO3, HCl and mixtures of both, using TBP were presented, where possible complexation was observed. It is thought that [Pu(NO3)4(TBP)4] or [PuCl4(TBP)4] species existed within the organic fraction for Pu(IV) separated from aqueous HNO3 and HCl, respectively. These systems showed high distribution ratios where an increase was observed with increasing aqueous acid concentration overall. Distribution ratio data were presented for the lanthanide series separated from aqueous nitric acid, using the proposed GANEX solvent system(s). The lanthanides analysed showed an increase in distribution ratio with increasing aqueous nitric acid concentration and with increasing TODGA concentration in the organic fraction. Heavier lanthanides were observed to give higher distribution ratios overall. The best distribution ratios were observed for lanthanides separated using 0.2 M TODGA with 1-octanol (5 % by volume) over the nitric acid concentration range analysed. For lanthanides separated using 0.5 M DMDOHEMA, an optimum distribution ratio was observed at around 6 M aqueous nitric acid concentration. The distribution ratio data for lanthanides separated from a range of DMDOHEMA concentrations, were observed to increase with increasing organic DMDOHEMA concentration. The distribution ratios observed for isotopes of Np, Am, Eu and Pu separated using 0.2 M TODGA, increased with increasing aqueous nitric acid concentration. The same trend was observed for the aforementioned isotopes separated using 0.5 M DMDOHEMA. However, pertechnetate separated using 0.2 M TODGA from aqueous nitric acid, showed a decrease in the distribution ratios observed over the acid concentration range analysed. This was contrary to pertechnetate separated from aqueous nitric acid using 0.5 M DMDOHEMA, where a small increase in distribution ratio was observed over the concentration range analysed. For Np(VI) separated from some proposed GANEX solvents, the 0.2 M TODGA/0.5 DMDOHEMA combination gave the best distribution of neptunium into the…

Subjects/Keywords: 553.4; Nuclear fuel reprocessing

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

APA (6^th Edition):

Tucker, K. L. (2015). Heavy metal extraction using advanced liquid-liquid style partitioning systems. (Doctoral Dissertation). University of Manchester. Retrieved from https://www.research.manchester.ac.uk/portal/en/theses/heavy-metal-extraction-using-advanced-liquid – liquid-style-partitioning-systems(6a238cb4-94cf-4fa8-bffe-d1a1b70adaa6).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.677739

Chicago Manual of Style (16^th Edition):

Tucker, Kate Louise. “Heavy metal extraction using advanced liquid-liquid style partitioning systems.” 2015. Doctoral Dissertation, University of Manchester. Accessed March 03, 2021. https://www.research.manchester.ac.uk/portal/en/theses/heavy-metal-extraction-using-advanced-liquid – liquid-style-partitioning-systems(6a238cb4-94cf-4fa8-bffe-d1a1b70adaa6).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.677739.

MLA Handbook (7^th Edition):

Tucker, Kate Louise. “Heavy metal extraction using advanced liquid-liquid style partitioning systems.” 2015. Web. 03 Mar 2021.

Vancouver:

Tucker KL. Heavy metal extraction using advanced liquid-liquid style partitioning systems. [Internet] [Doctoral dissertation]. University of Manchester; 2015. [cited 2021 Mar 03]. Available from: https://www.research.manchester.ac.uk/portal/en/theses/heavy-metal-extraction-using-advanced-liquid – liquid-style-partitioning-systems(6a238cb4-94cf-4fa8-bffe-d1a1b70adaa6).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.677739.

Council of Science Editors:

Tucker KL. Heavy metal extraction using advanced liquid-liquid style partitioning systems. [Doctoral Dissertation]. University of Manchester; 2015. Available from: https://www.research.manchester.ac.uk/portal/en/theses/heavy-metal-extraction-using-advanced-liquid – liquid-style-partitioning-systems(6a238cb4-94cf-4fa8-bffe-d1a1b70adaa6).html ; http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.677739

University of Illinois – Urbana-Champaign

6. Lartonoix, David. Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica.

Degree: MS, 5183, 2012, University of Illinois – Urbana-Champaign

URL: http://hdl.handle.net/2142/31990

► While nuclear reactors in the United States have produced economy-driving power for several decades, they have also left behind a significant amount of spent nuclear… (more)

▼ While nuclear reactors in the United States have produced economy-driving power for several decades, they have also left behind a significant amount of spent nuclear fuel. The federal government, ultimately responsible for this spent fuel, has a history just as long in attempting to effectively bury, dispose, reprocess, or otherwise deal with this waste. As no attempts to date have been entirely successful, work continues to find an effective waste management solution. To aid planners, policymakers, and scientists in this endeavor, tools are currently needed to model the radioisotope inventory of all spent nuclear requiring disposal or other forms of remediation to accurately frame the scope of the issue. This project describes a simple method of calculating radioisotope concentrations in spent fuel by utilizing a unique approach to solving the diffusion equation eigenvalue problem. Herein, the dissolved boron concentration, essentially a chemical shim, is adjusted over an operational time period to maintain criticality in the reactor, compensating for fuel burnup, burnable poison burnout, and actinide and fission product buildup. It is shown, as an example, that the fractional reduction in boron concentration after a month of reactor operation is 2.3%. The normalized neutron flux in the example scenario is calculated and confirmed to be relatively flat radially and vertically. Similarly, the normalized thermal energy production rate is also shown to be relatively flat, as expected. Radionuclides of interest are tracked and isotopic concentrations are shown at various vertical heights within the core. Upon further refining, these concentrations can be taken to represent the radioisotope inventory of spent nuclear fuel under various burnup scenarios. Ultimately, characterizing the spent fuel requiring disposal will aid in developing an efficient waste management strategy. Even while several shortfalls are noted and described, tools such as this computer code can play a useful role in addressing the nation's nuclear waste disposal dilemma. Advisors/Committee Members: Singer, Clifford E. (advisor).

Subjects/Keywords: spent nuclear fuel; radioisotope inventory

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

APA (6^th Edition):

Lartonoix, D. (2012). Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica. (Thesis). University of Illinois – Urbana-Champaign. Retrieved from http://hdl.handle.net/2142/31990

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):

Lartonoix, David. “Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica.” 2012. Thesis, University of Illinois – Urbana-Champaign. Accessed March 03, 2021. http://hdl.handle.net/2142/31990.

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

MLA Handbook (7^th Edition):

Lartonoix, David. “Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica.” 2012. Web. 03 Mar 2021.

Vancouver:

Lartonoix D. Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica. [Internet] [Thesis]. University of Illinois – Urbana-Champaign; 2012. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/2142/31990.

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

Council of Science Editors:

Lartonoix D. Radioisotope Inventory Of Spent Nuclear Fuel In Mathematica. [Thesis]. University of Illinois – Urbana-Champaign; 2012. Available from: http://hdl.handle.net/2142/31990

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

Virginia Tech

7. Faierson, Eric J. Structure-Property Relationships of Tantalum Carbide Foams and Synthesis of an Interpenetrating Phase Composite.

Degree: PhD, Materials Science and Engineering, 2011, Virginia Tech

URL: http://hdl.handle.net/10919/28688

► Ceramic and refractory metal foams have a potential for use in extreme environments, such as in fuel elements within nuclear reactors both in space and… (more)

▼ Ceramic and refractory metal foams have a potential for use in extreme environments, such as in fuel elements within nuclear reactors both in space and terrestrial applications. In addition, infiltrating an open-cell ceramic foam with a continuous second phase can create an interpenetrating phase composite (IPC), consisting of a three-dimensional reinforcement structure. One aspect of investigation within this study was the influence of foam pore/strut size, foam composition, and foam density on neutronic and mechanical properties. Neutron transmission through open-cell tantalum carbide foams was measured using experimental techniques and modeled with Monte Carlo N-Particle (MCNP) transport code. Neutron transmission decreased linearly within tantalum carbide (TaC)/reticulated vitreous carbon (RVC) foams as areal TaC density increased. All MCNP modeling runs predicted slightly higher neutron transmission than what was experimentally measured, potentially indicating that the foam structure had a small influence on neutron transmission. Compressive strength and Young's moduli of tantalum carbide foams were measured for foam specimens that were exposed to thermal cycling and thermal shock, as well as for baseline specimens. Extensive micro-cracking was observed in the foams after 18 thermal cycles to 2100°C. However, thermal shock in liquid nitrogen did not produce observable micro-cracking in the TaC foams. The average strengths of baseline TaC/RVC foams ranged from 1.97 MPa - 3.82 MPa. The baseline TaC/PyC/RVC foams exhibited strengths ranging from 4.57 MPa - 12.60 MPa. The compressive strength of thermally cycled foams tended to be 1/3-1/2 that of baseline specimens. Another aspect of this study investigated the infiltration of RVC foams with tungsten powder in an attempt to form a tungsten-ceramic foam interpenetrating phase composite (IPC). It was found that tungsten particle size influenced infiltrated densities more than foam pore size. Significantly lower infiltrated densities were obtained using sub-micron tungsten than with 5-10 micron tungsten as a result of particle agglomeration. Infiltrated 5-10 micron tungsten achieved densities ranging from 23-25% theoretical within RVC foams, whereas sub-micron tungsten densities ranged from 11-16% theoretical. Constrained densification was observed during sintering of tungsten-infiltrated foams. Advisors/Committee Members: Logan, Kathryn V. (committeechair), Clark, David E. (committee member), Dowling, Norman E. (committee member), Kiefer, Richard L. (committee member).

Subjects/Keywords: foam; composite; neutron; nuclear fuel

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

APA (6^th Edition):

Faierson, E. J. (2011). Structure-Property Relationships of Tantalum Carbide Foams and Synthesis of an Interpenetrating Phase Composite. (Doctoral Dissertation). Virginia Tech. Retrieved from http://hdl.handle.net/10919/28688

Chicago Manual of Style (16^th Edition):

Faierson, Eric J. “Structure-Property Relationships of Tantalum Carbide Foams and Synthesis of an Interpenetrating Phase Composite.” 2011. Doctoral Dissertation, Virginia Tech. Accessed March 03, 2021. http://hdl.handle.net/10919/28688.

MLA Handbook (7^th Edition):

Faierson, Eric J. “Structure-Property Relationships of Tantalum Carbide Foams and Synthesis of an Interpenetrating Phase Composite.” 2011. Web. 03 Mar 2021.

Vancouver:

Faierson EJ. Structure-Property Relationships of Tantalum Carbide Foams and Synthesis of an Interpenetrating Phase Composite. [Internet] [Doctoral dissertation]. Virginia Tech; 2011. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/10919/28688.

Council of Science Editors:

Faierson EJ. Structure-Property Relationships of Tantalum Carbide Foams and Synthesis of an Interpenetrating Phase Composite. [Doctoral Dissertation]. Virginia Tech; 2011. Available from: http://hdl.handle.net/10919/28688

University of South Carolina

8. Shalloo, Matthew. Characterization and Drying of Oxyhydroxides on Aluminum Clad Spent Nuclear Fuel.

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

URL: https://scholarcommons.sc.edu/etd/5509

► Research reactors such as the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) employ aluminum-clad fuel elements made up of many… (more)

▼ Research reactors such as the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR) employ aluminum-clad fuel elements made up of many thin plates with uranium dispersed within. In most engineering applications, aluminum is considered to have favorable corrosion characteristics. It forms a thin oxide layer [Al2O3] under atmospheric conditions that is impenetrable to oxygen thus stopping any further corrosion. However, both aluminum metal and Al2O3 react with water to form hydrous oxides which are less protective against further corrosion and form significantly thicker layers than oxidation in dry air. As a result, aluminum-clad spent nuclear fuel (ASNF) hosts chemisorbed bound water on the fuel surface. In addition, adsorbed or physiosorbed water contributes to the total water within the oxide layer. This is a challenge for sealed dry storage of ASNF because the physiosorbed water and water in the hydroxides could be released as free water at high temperature or decomposed by radiolysis leading to further corrosion and a buildup of pressure within the cannister. The goal of this research is to study the formation of lab-grown oxides on aluminum samples as surrogates for those on ASNF, characterize those oxide layers, and quantify the conditions necessary to remove bulk, physiosorbed, and chemisorbed water. This knowledge will be used to set parameters for full-scale drying studies of ASNF later on. Testing of aluminum oxide powder samples by Thermogravimetric Analysis and Differential Scanning Calorimetry (TGA/DSC) has been performed on commercially available oxyhydroxide powders to determine the dehydroxylation temperatures to be expected in bulk tests. Gibbsite was found to decompose at about 300°C while dehydroxylation for fine and coarse boehmite averaged around 520°C, and 440°C respectively. Aluminum coupons of Al-1100, Al-5052, and Al-6061 were immersed in distilled water at 20°C, 50°C, and 100°C to produce a hydrated oxide layer. Bulk drying tests conducted via Thermogravimetric Analysis (TGA) on these aluminum-cladding surrogate samples found dewatering for 20°C, 50°C, and 100°C samples to initiate at modest temperatures below 100°C. The amount of water removed depended on a combination of the heating period and maximum temperature. However, even in low temperature TGA runs, the total amount of water removed matched closely with higher temperature runs as long as the low temperature was maintained for a sufficiently long time. Imaging by Scanning Electron Microscope (SEM) and analysis by X-Ray Diffraction (XRD) took place throughout the research for a detailed understanding of the microstructure and crystal structure at each stage of the process. Based on the findings from this work it is believed that the current drying process of vacuuming the drying canister to 5Torr and heating to 220°C for 35 to 45 minutes in air cyclically is insufficient for removing the maximum chemically bound water. Instead, the drying process should involve heating the spent fuel elements continuously to… Advisors/Committee Members: Travis W. Knight.

Subjects/Keywords: Nuclear Engineering; nuclear fuel; fuel; Aluminum Clad; Oxyhydroxides

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

APA (6^th Edition):

Shalloo, M. (2019). Characterization and Drying of Oxyhydroxides on Aluminum Clad Spent Nuclear Fuel. (Thesis). University of South Carolina. Retrieved from https://scholarcommons.sc.edu/etd/5509

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):

Shalloo, Matthew. “Characterization and Drying of Oxyhydroxides on Aluminum Clad Spent Nuclear Fuel.” 2019. Thesis, University of South Carolina. Accessed March 03, 2021. https://scholarcommons.sc.edu/etd/5509.

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

MLA Handbook (7^th Edition):

Shalloo, Matthew. “Characterization and Drying of Oxyhydroxides on Aluminum Clad Spent Nuclear Fuel.” 2019. Web. 03 Mar 2021.

Vancouver:

Shalloo M. Characterization and Drying of Oxyhydroxides on Aluminum Clad Spent Nuclear Fuel. [Internet] [Thesis]. University of South Carolina; 2019. [cited 2021 Mar 03]. Available from: https://scholarcommons.sc.edu/etd/5509.

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

Council of Science Editors:

Shalloo M. Characterization and Drying of Oxyhydroxides on Aluminum Clad Spent Nuclear Fuel. [Thesis]. University of South Carolina; 2019. Available from: https://scholarcommons.sc.edu/etd/5509

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

Texas A&M University

9. Smith, Joshua 1987-. Enhanced Thermal Conductivity UO2-BeO Nuclear Fuel: Neutronic Performance Studies and Economic Analyses.

Degree: MS, Nuclear Engineering, 2012, Texas A&M University

URL: http://hdl.handle.net/1969.1/156356

► The objective of this work was to continue the evaluation of the high thermal conductivity UO2-BeO (UBO) nuclear fuel. Current ceramic UO2 fuel offers many… (more)

▼ The objective of this work was to continue the evaluation of the high thermal conductivity UO2-BeO (UBO) nuclear fuel. Current ceramic UO2 fuel offers many fuel performance benefits, but it has a low thermal conductivity. This results in high operating fuel temperatures, but this is a well-excepted performance compromise. Addition of Beryllium oxide to the fuel structure has been shown to increase the fuel thermal conductivity and provide positive neutronic benefits. Pellet heat conduction studies were performed at different linear heat generation rates (LHGR). At an average LHGR of 163.4 W/cm, UBO 10vol% fuel showed a decrease of 74 and 166 degrees C in the effective and centerline temperature, respectively. Similarly at a peak LHGR of 590 W/cm, UBO 10vol% fuel showed a decrease of 219 and 493 degrees C, respectively. A drawback to UBO fuel is the lower eutectic melting point. At 590 W/cm and beginning of cycle, the melting margin for UO2 and UBO 10vol% is 411 and 254 degrees C, respectively. Comparisons of fuel types were performed using 2D infinite lattice and 3D equilibrium core neutronic simulations. A 2D lattice analysis showed that an increased UBO fuel enrichment is necessary to maintain an equivalent cycle length as UO2 fuel. Using a mass equivalent 235U basis and the linear reactivity model for an 18-month cycle, UBO 5vol% and 10vol% fuel showed a cycle increase of 1.9 and 3.3 days, respectively. Similarly, the 3D core simulation showed a cycle increase of 2.2 and 3.3 days, respectively. However, the maximum 3D burnup was increased by 3707 and 7624 MWd/t, respectively, which may cause selective UBO placement. An economic analysis of UBO fuel compared the increased cycle length to the extra fuel costs associated with UBO fuel. The 18-month break-even fuel cost occurred at 4.03 and 8.15 days for the UBO 5vol% and 10vol% fuel, respectively. Since the computed cycle length was shorter than the break-even fuel cost, this resulted in a -12,365 and -25,712 $/reload-assembly penalty, respectively. Advisors/Committee Members: Ragusa, Jean C (advisor), McDeavitt, Sean M (advisor), Khatri, Sunil (committee member).

Subjects/Keywords: Levelized Fuel Cost BeO Nuclear Fuel; Nuclear Core Simulations; BeO Enhanced Conductivity Fuel; Beryllium Oxide Nuclear Fuel

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APA (6^th Edition):

Smith, J. 1. (2012). Enhanced Thermal Conductivity UO2-BeO Nuclear Fuel: Neutronic Performance Studies and Economic Analyses. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/156356

Chicago Manual of Style (16^th Edition):

Smith, Joshua 1987-. “Enhanced Thermal Conductivity UO2-BeO Nuclear Fuel: Neutronic Performance Studies and Economic Analyses.” 2012. Masters Thesis, Texas A&M University. Accessed March 03, 2021. http://hdl.handle.net/1969.1/156356.

MLA Handbook (7^th Edition):

Smith, Joshua 1987-. “Enhanced Thermal Conductivity UO2-BeO Nuclear Fuel: Neutronic Performance Studies and Economic Analyses.” 2012. Web. 03 Mar 2021.

Vancouver:

Smith J1. Enhanced Thermal Conductivity UO2-BeO Nuclear Fuel: Neutronic Performance Studies and Economic Analyses. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/1969.1/156356.

Council of Science Editors:

Smith J1. Enhanced Thermal Conductivity UO2-BeO Nuclear Fuel: Neutronic Performance Studies and Economic Analyses. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/156356

10. Kelly Cristina Martins Faêda. Caracterização do combustível para reatores nucleares produtores de hidrogênio.

Degree: Master, 2011, Centro de Desenvolvimento da Tecnologia Nuclear

URL: http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=139 ;

►

Reatores nucleares de 4 geração do tipo HTGR (reatores de alta temperatura refrigerados a gás) apresentam vantagens em relação a um reator a água pressurizada,… (more)

▼

Reatores nucleares de 4 geração do tipo HTGR (reatores de alta temperatura refrigerados a gás) apresentam vantagens em relação a um reator a água pressurizada, do tipo de Angra I e II, como maior eficiência térmica, possibilidade de atingir queimas do combustível dez vezes mais altas e de troca de combustível com o reator em marcha. Devido à alta temperatura do núcleo do reator, eles também são considerados para a produção de hidrogênio, além da produção de energia elétrica. A produção do hidrogênio significa a inserção em um novo mercado para as operadoras das centrais nucleares, com características diferentes do mercado de eletricidade. Esse fato requer um longo preparo das operadoras, porque a compatibilização desses dois mercados na operação das centrais nucleares certamente será uma tarefa complexa. No caso brasileiro, o fornecimento de hidrogênio para o refino do petróleo pode ser o nicho mais claro para a introdução dos reatores nucleares produtores de hidrogênio. No caso do processo de fabricação do combustível nuclear, as caracterizações são realizadas com o intuito de garantir a minimização dos efeitos danosos da queima e da temperatura, de tal forma a assegurar o confinamento dos produtos de fissão e manter o combustível funcionando durante o tempo de sua permanência no núcleo do reator. Contudo a questão metrológica não tem recebido atenção suficiente. Neste trabalho é apresentado o estado da arte do desenvolvimento relativo à produção de hidrogênio por reatores nucleares e uma abordagem para o caso do Brasil. Adicionalmente, foi feito um estudo das técnicas de caracterizações relacionadas com algumas das principais propriedades do combustível nuclear, que são as mais críticas para o seu desempenho. Foram feitos estudos visando à otimização de rotinas experimentais para determinação densidadade, porosidade aberta, difusividade térmica, condutividade térmica e calor específico de pastilhas de UO2. Os valores obtidos nas medições realizadas apresentaram diferenças em relação aos valores reportados na literatura. Uma causa para essa diferença pode ser devido à presença de uma fase com relação O/U maior que 2 nas amostras utilizadas. Embora a difração de raios X não tenha sido capaz de identificar outras fases nas amostras de UO2, a espectroscopia na região do infravermelho se mostrou bastante sensível à presença dessas fases. Sugere-se que esta técnica, devido à sua facilidade experimental, seja incluída nas rotinas de caracterização de UO2, de forma a completar as informações fornecidas pela termogravimetria e a difração de raios X.

HTGR nuclear reactors of the 4th generation have advantages in relation to a pressurized water reactor, like Angra I and II, as higher thermal efficiency, ability to reach burnups ten times higher, and fuel reloading with the reactor running at full power. Due to the high temperature of the reactor core, they are also considered for the production of hydrogen, besides electricity. This work presents a review of the state of the art of developments related to hydrogen…

Advisors/Committee Members: Wilmar Barbosa Ferraz, Fernando Soares Lameiras, Marcio Soares Dias.

Subjects/Keywords: COMBUSTIVEL NUCLEAR; Reator Nuclear Combustível Nuclear Analises química; Nuclear fuel Nuclear reactor Chemical analysis

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APA (6^th Edition):

Faêda, K. C. M. (2011). Caracterização do combustível para reatores nucleares produtores de hidrogênio. (Masters Thesis). Centro de Desenvolvimento da Tecnologia Nuclear. Retrieved from http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=139 ;

Chicago Manual of Style (16^th Edition):

Faêda, Kelly Cristina Martins. “Caracterização do combustível para reatores nucleares produtores de hidrogênio.” 2011. Masters Thesis, Centro de Desenvolvimento da Tecnologia Nuclear. Accessed March 03, 2021. http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=139 ;.

MLA Handbook (7^th Edition):

Faêda, Kelly Cristina Martins. “Caracterização do combustível para reatores nucleares produtores de hidrogênio.” 2011. Web. 03 Mar 2021.

Vancouver:

Faêda KCM. Caracterização do combustível para reatores nucleares produtores de hidrogênio. [Internet] [Masters thesis]. Centro de Desenvolvimento da Tecnologia Nuclear; 2011. [cited 2021 Mar 03]. Available from: http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=139 ;.

Council of Science Editors:

Faêda KCM. Caracterização do combustível para reatores nucleares produtores de hidrogênio. [Masters Thesis]. Centro de Desenvolvimento da Tecnologia Nuclear; 2011. Available from: http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=139 ;

Texas A&M University

11. Totemeier, Aaron Robert. Helium Ion Implantation in Zirconium: Bubble Formation & Growth.

Degree: PhD, Nuclear Engineering, 2015, Texas A&M University

URL: http://hdl.handle.net/1969.1/156473

► To evaluate the behavior of inert helium gas bubbles in zirconium three variants of the metal were implanted with 140 keV helium ions to a… (more)

Subjects/Keywords: Nuclear; Zirconium, Helium Bubbles; Ion Implantation; Nuclear Fuel; Nuclear Waste

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APA (6^th Edition):

Totemeier, A. R. (2015). Helium Ion Implantation in Zirconium: Bubble Formation & Growth. (Doctoral Dissertation). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/156473

Chicago Manual of Style (16^th Edition):

Totemeier, Aaron Robert. “Helium Ion Implantation in Zirconium: Bubble Formation & Growth.” 2015. Doctoral Dissertation, Texas A&M University. Accessed March 03, 2021. http://hdl.handle.net/1969.1/156473.

MLA Handbook (7^th Edition):

Totemeier, Aaron Robert. “Helium Ion Implantation in Zirconium: Bubble Formation & Growth.” 2015. Web. 03 Mar 2021.

Vancouver:

Totemeier AR. Helium Ion Implantation in Zirconium: Bubble Formation & Growth. [Internet] [Doctoral dissertation]. Texas A&M University; 2015. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/1969.1/156473.

Council of Science Editors:

Totemeier AR. Helium Ion Implantation in Zirconium: Bubble Formation & Growth. [Doctoral Dissertation]. Texas A&M University; 2015. Available from: http://hdl.handle.net/1969.1/156473

University of Tennessee – Knoxville

12. Petersen, Gordon Matthew. Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy.

Degree: 2016, University of Tennessee – Knoxville

URL: https://trace.tennessee.edu/utk_graddiss/4156

► Commercial nuclear power plants produce long-lasting nuclear waste, primarily in the form of spent nuclear fuel (SNF) assemblies. Spent fuel pools (SFP) and canisters or… (more)

Subjects/Keywords: Optimization; Spent Nuclear Fuel; Allocation Strategy; Fuel Removal; Used Nuclear Fuel; Tractable Validation Model; Nuclear Engineering

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APA (6^th Edition):

Petersen, G. M. (2016). Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy. (Doctoral Dissertation). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_graddiss/4156

Chicago Manual of Style (16^th Edition):

Petersen, Gordon Matthew. “Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy.” 2016. Doctoral Dissertation, University of Tennessee – Knoxville. Accessed March 03, 2021. https://trace.tennessee.edu/utk_graddiss/4156.

MLA Handbook (7^th Edition):

Petersen, Gordon Matthew. “Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy.” 2016. Web. 03 Mar 2021.

Vancouver:

Petersen GM. Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy. [Internet] [Doctoral dissertation]. University of Tennessee – Knoxville; 2016. [cited 2021 Mar 03]. Available from: https://trace.tennessee.edu/utk_graddiss/4156.

Council of Science Editors:

Petersen GM. Algorithms and Methods for Optimizing the Spent Nuclear Fuel Allocation Strategy. [Doctoral Dissertation]. University of Tennessee – Knoxville; 2016. Available from: https://trace.tennessee.edu/utk_graddiss/4156

McMaster University

13. Friedlander, Yonni. A SCOPING STUDY OF ADVANCED THORIUM FUEL CYCLES FOR CANDU REACTORS.

Degree: MASc, 2011, McMaster University

URL: http://hdl.handle.net/11375/11283

►

A study was conducted to scope the relative merits of various thorium fuel cycles in CANDU reactors. It was determined that, due to the… (more)

▼

A study was conducted to scope the relative merits of various thorium fuel cycles in CANDU reactors. It was determined that, due to the very large reprocessing demands of the self-sustaining equilibrium thorium fuel cycle, an additional fissile driver fuel is required for a practical thorium fuel cycle. The driver fuels considered were PWR- and CANDU- derived plutonium, PWR-derived MOX fuel, and low-enriched uranium. The addition a RU-fuelled CANDU reactor with possible americium, curium, and lanthanides recycling was also considered. The fuel cycles were evaluated for natural uranium consumption and reprocessing demands as well as spent fuel characteristics such as: thermal and gamma power, radioactivity, and ingestion and inhalation hazards. The two-dimensional multigroup code, WIMS-AECL, was used to calculate the burnup and some controllability properties of the CANDU reactors. ORIGEN, a depletion and decay module, was used to evaluate the spent fuel characteristics and the systems code, DESAE, was used to simulate the introduction of the thorium fuel cycle to a growing global reactor park. It was determined that ²³³U production in the thorium fuels is optimized at lower exit burnups. Therefore, less external fissile driver material is required for the operation of the thorium reactors and natural uranium savings of the overall fuel cycle are increased. Furthermore, it was determined that driving the thorium fuel cycle with low-enriched uranium is the most efficient way to minimize natural uranium consumption. Assuming that a 40 MWd/kg exit burnup was achieved in the CANDU reactors, the fuel cycle yielded an 82% savings of natural uranium, compared to a scenario in which all power came from PWRs, while a 20 MWd/kg exit burnup increased the savings to 94%. The savings ranged over those exit burnups from 55%-69% for the variant with PWR-derived plutonium, 60%-73% for PWR-derived MOX fuel, and 78%-87% for CANDU-derived plutonium. The thermal power, radioactivity, and health hazards of the spent fuel were the highest for the case with MOX fuel driver but americium recycling proved effective for decreasing the long term dangers. In a global reactor park, the introduction of the thorium fuel cycle was hampered by the availability of fissile resources and, for a PWR-derived driver, natural uranium consumption was only reduced by 22% over 100 years relative to the PWR only scenario.

Master of Applied Science (MASc)

Advisors/Committee Members: Luxat, John, Engineering Physics and Nuclear Engineering.

Subjects/Keywords: Nuclear; Advanced Fuel Cycles; Thorium; CANDU; Nuclear resources; Nuclear Engineering; Nuclear Engineering

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

APA (6^th Edition):

Friedlander, Y. (2011). A SCOPING STUDY OF ADVANCED THORIUM FUEL CYCLES FOR CANDU REACTORS. (Masters Thesis). McMaster University. Retrieved from http://hdl.handle.net/11375/11283

Chicago Manual of Style (16^th Edition):

Friedlander, Yonni. “A SCOPING STUDY OF ADVANCED THORIUM FUEL CYCLES FOR CANDU REACTORS.” 2011. Masters Thesis, McMaster University. Accessed March 03, 2021. http://hdl.handle.net/11375/11283.

MLA Handbook (7^th Edition):

Friedlander, Yonni. “A SCOPING STUDY OF ADVANCED THORIUM FUEL CYCLES FOR CANDU REACTORS.” 2011. Web. 03 Mar 2021.

Vancouver:

Friedlander Y. A SCOPING STUDY OF ADVANCED THORIUM FUEL CYCLES FOR CANDU REACTORS. [Internet] [Masters thesis]. McMaster University; 2011. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/11375/11283.

Council of Science Editors:

Friedlander Y. A SCOPING STUDY OF ADVANCED THORIUM FUEL CYCLES FOR CANDU REACTORS. [Masters Thesis]. McMaster University; 2011. Available from: http://hdl.handle.net/11375/11283

University of Pretoria

14. Hlatshwayo, Thulani Thokozani. Diffusion of silver in 6H-SiC .

Degree: 2011, University of Pretoria

URL: http://upetd.up.ac.za/thesis/available/etd-06182011-165556/

► SiC is used as the main diffusion barrier in the fuel spheres of the pebble bed modular reactor (PBMR). The PBMR is a modern high… (more)

▼ SiC is used as the main diffusion barrier in the fuel spheres of the pebble bed modular reactor (PBMR). The PBMR is a modern high temperature nuclear reactor. However, the release of silver from the fuel spheres has raised some doubts about the effectiveness of this barrier, which has led to many studies on the possible migration paths of silver. The reported results of these studies have shown largely differing results concerning the magnitude and temperature dependence of silver being transported through the fuel particle coatings. Results from earlier investigations could be interpreted as a diffusion process governed by an Arrhenius type temperature dependence. In this study, the silver diffusion in 6H-SiC was investigated using two methods. In the first method a thin silver layer was deposited on 6H-SiC by vapour deposition while in the second method silver was implanted in 6H-SiC at room temperature, 350°C and 600°C to a fluence of 2×1016 silver ions cm-2. Finally the effect of neutron irradiation on the diffusion of silver was investigated for the samples implanted at 350°C and 600°C. Silver depth profiles before and after annealing were determined by Rutherford backscattering (RBS). Both isothermal and isochronal annealing were used in this study. Diffusion coefficients as well as detection limits were extracted by comparing the silver depth profiles before and after annealing. The radiation damage after implantation and their recovery after isothermal and isochronal annealing were analysed by Rutherford backscattering spectroscopy combined with channelling. The results of in-diffusion of silver into 6H-SiC at temperatures below the melting point (960°C) using un-encapsulated 6H-SiC samples with 100 nm deposited silver indicated no in-diffusion of silver; however, disappearance of silver occurred at these temperatures. For the encapsulated samples, no in-diffusion of silver was observed at 800°C, 900°C and 1000°C but silver disappeared from the samples’ surface and was found on the walls of the quartz glass ampoule. This disappearance of silver was established to be due to the wetting problem that existed between silver and SiC. The room temperature implantation resulted in a completely amorphous surface layer of approximately 270 nm thick. Epitaxial re-growth from the bulk was already taking place during annealing at 700°C and the crystalline structure seemed to be fully recovered at 1600°C, for samples that were sequentially isochronally annealed from 700°C in steps of 100°C up to 1600°C. However, no silver signal was detected at this temperature, which left certain doubts regarding the crystalline structure of the samples at this temperature. This was speculated to be due to thermal etching of the top original amorphous layer while the deeper amorphous layer was epitaxial re-growth from the bulk. The decomposition of SiC, giving rise to a carbon peak in the RBS spectra due to evaporation of Si, was clearly observed on the same samples at 1600°C. Isothermal annealing at 1300°C for 10 h cycles up to 80h… Advisors/Committee Members: Malherbe, Johan B (advisor), Friedland, Erich Karl Helmuth (advisor).

Subjects/Keywords: Nuclear reactor; Fuel spheres; Diffusion barrier; UCTD

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APA (6^th Edition):

Hlatshwayo, T. T. (2011). Diffusion of silver in 6H-SiC . (Doctoral Dissertation). University of Pretoria. Retrieved from http://upetd.up.ac.za/thesis/available/etd-06182011-165556/

Chicago Manual of Style (16^th Edition):

Hlatshwayo, Thulani Thokozani. “Diffusion of silver in 6H-SiC .” 2011. Doctoral Dissertation, University of Pretoria. Accessed March 03, 2021. http://upetd.up.ac.za/thesis/available/etd-06182011-165556/.

MLA Handbook (7^th Edition):

Hlatshwayo, Thulani Thokozani. “Diffusion of silver in 6H-SiC .” 2011. Web. 03 Mar 2021.

Vancouver:

Hlatshwayo TT. Diffusion of silver in 6H-SiC . [Internet] [Doctoral dissertation]. University of Pretoria; 2011. [cited 2021 Mar 03]. Available from: http://upetd.up.ac.za/thesis/available/etd-06182011-165556/.

Council of Science Editors:

Hlatshwayo TT. Diffusion of silver in 6H-SiC . [Doctoral Dissertation]. University of Pretoria; 2011. Available from: http://upetd.up.ac.za/thesis/available/etd-06182011-165556/

Oregon State University

15. Vincent, Andrew M. Fuel-cladding mechanical interaction in simulated sphere pac fuel.

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

URL: http://hdl.handle.net/1957/41696

► A study has been made to investigate the fuel-cladding mechanical interaction (FCMI) in sphere pac fuels. An FCMI simulation has been constructed using the JVG… (more)

Subjects/Keywords: Nuclear fuel claddings

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APA (6^th Edition):

Vincent, A. M. (1981). Fuel-cladding mechanical interaction in simulated sphere pac fuel. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/41696

Chicago Manual of Style (16^th Edition):

Vincent, Andrew M. “Fuel-cladding mechanical interaction in simulated sphere pac fuel.” 1981. Masters Thesis, Oregon State University. Accessed March 03, 2021. http://hdl.handle.net/1957/41696.

MLA Handbook (7^th Edition):

Vincent, Andrew M. “Fuel-cladding mechanical interaction in simulated sphere pac fuel.” 1981. Web. 03 Mar 2021.

Vancouver:

Vincent AM. Fuel-cladding mechanical interaction in simulated sphere pac fuel. [Internet] [Masters thesis]. Oregon State University; 1981. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/1957/41696.

Council of Science Editors:

Vincent AM. Fuel-cladding mechanical interaction in simulated sphere pac fuel. [Masters Thesis]. Oregon State University; 1981. Available from: http://hdl.handle.net/1957/41696

Texas A&M University

16. Hogelin, Thomas Russell. Radioactive Flow Characterization for Real-Time Detection Systems in UREX+ Nuclear Fuel Reprocessing.

Degree: MS, Nuclear Engineering, 2011, Texas A&M University

URL: http://hdl.handle.net/1969.1/ETD-TAMU-2010-12-8638

► The reprocessing of used nuclear fuel requires the dissolution and separation of numerous radioisotopes that are present as fission products in the fuel. The leading… (more)

▼ The reprocessing of used nuclear fuel requires the dissolution and separation of numerous radioisotopes that are present as fission products in the fuel. The leading technology option in the U.S. for reprocessing is a sequence of processing methods known as UREX+ (Uranium Extraction ). However, an industrial scale facility implementing this separation procedure will require the establishment of safeguards and security systems to ensure the protection of the separated materials. A number of technologies have been developed for meeting the measurement demands for such a facility. This project focuses on the design of a gamma detection system for taking measurements of the flow streams of such a reprocessing facility. An experimental apparatus was constructed capable of pumping water spiked with soluble radioisotopes under various flow conditions through a stainless steel coil around a sodium iodide (NaI) detector system. Experiments were conducted to characterize the impact of flow rate, pipe air voids, geometry, and radioactivity dilution level on activity measurements and gamma energy spectra. Two coil geometries were used for these experiments, using 0.5 in stainless steel pipe wound into a coil with a 6 inch diameter; the first coil was 5.5 revolutions tall and the second coil was 9.5 revolutions tall. The isotopes dissolved in the flowing water were produced at the Texas A&M Nuclear Science Center via neutron activation of chromium, gold, cerium, and ytterbium nitrate salts. After activation, the salts were dissolved in distilled water and inserted into the radioactive flow assembly for quantitative measurements. Flow rate variations from 100 to 2000 ml/min were used and activity dilution levels for the experiments conducted were between 0.02 and 1.6 μCi/liter. Detection of system transients was observed to improve with decreasing flow rate. The detection limits observed for this system were 0.02 μCi/liter over background, 0.5% total activity change in a pre-spiked system, and a dilution change of 2% of the coil volume. MCNP (Monte Carlo N-Particle Transport) models were constructed to simulate the results and were used to extend the results to other geometries and piping materials as well as simulate actual UREX stream material in the system. The stainless steel piping for the flow around the detector was found to attenuate key identifying gamma peaks on the low end of the energy spectrum. For the proposed schedule 40 stainless steel pipe for an actual reprocessing facility, gamma rays below 100 keV in energy would be reduced to less than half their initial intensities. The exact ideal detection set up is largely activity and flow stream dependant. However, the characteristics best suited for flow stream detection are: 1) minimize volume around detector, 2) low flow rate for long count times, and 3) low attenuation piping material such as glass. Advisors/Committee Members: McDeavitt, Sean M. (advisor), Charlton, William (committee member), Radovic, Miladin (committee member).

Subjects/Keywords: nuclear; fuel; reprocessing; UREX; radioactive; HRGS

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APA (6^th Edition):

Hogelin, T. R. (2011). Radioactive Flow Characterization for Real-Time Detection Systems in UREX+ Nuclear Fuel Reprocessing. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2010-12-8638

Chicago Manual of Style (16^th Edition):

Hogelin, Thomas Russell. “Radioactive Flow Characterization for Real-Time Detection Systems in UREX+ Nuclear Fuel Reprocessing.” 2011. Masters Thesis, Texas A&M University. Accessed March 03, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2010-12-8638.

MLA Handbook (7^th Edition):

Hogelin, Thomas Russell. “Radioactive Flow Characterization for Real-Time Detection Systems in UREX+ Nuclear Fuel Reprocessing.” 2011. Web. 03 Mar 2021.

Vancouver:

Hogelin TR. Radioactive Flow Characterization for Real-Time Detection Systems in UREX+ Nuclear Fuel Reprocessing. [Internet] [Masters thesis]. Texas A&M University; 2011. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-12-8638.

Council of Science Editors:

Hogelin TR. Radioactive Flow Characterization for Real-Time Detection Systems in UREX+ Nuclear Fuel Reprocessing. [Masters Thesis]. Texas A&M University; 2011. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-12-8638

University of Saskatchewan

17. Oladimeji, Dotun J 1989-. Thermal Conductivity of Nuclear Fuel and its Degradation by Physical and Chemical Burnup.

Degree: 2017, University of Saskatchewan

URL: http://hdl.handle.net/10388/8024

► Nuclear fuel performance during reactor operation has been studied using both atomic scale simulation and experimental procedure in order to investigate how nuclear fission process… (more)

Subjects/Keywords: Nuclear Fuel; Molecular Dynamics; Thermal conductivity

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APA (6^th Edition):

Oladimeji, D. J. 1. (2017). Thermal Conductivity of Nuclear Fuel and its Degradation by Physical and Chemical Burnup. (Thesis). University of Saskatchewan. Retrieved from http://hdl.handle.net/10388/8024

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):

Oladimeji, Dotun J 1989-. “Thermal Conductivity of Nuclear Fuel and its Degradation by Physical and Chemical Burnup.” 2017. Thesis, University of Saskatchewan. Accessed March 03, 2021. http://hdl.handle.net/10388/8024.

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

MLA Handbook (7^th Edition):

Oladimeji, Dotun J 1989-. “Thermal Conductivity of Nuclear Fuel and its Degradation by Physical and Chemical Burnup.” 2017. Web. 03 Mar 2021.

Vancouver:

Oladimeji DJ1. Thermal Conductivity of Nuclear Fuel and its Degradation by Physical and Chemical Burnup. [Internet] [Thesis]. University of Saskatchewan; 2017. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/10388/8024.

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

Council of Science Editors:

Oladimeji DJ1. Thermal Conductivity of Nuclear Fuel and its Degradation by Physical and Chemical Burnup. [Thesis]. University of Saskatchewan; 2017. Available from: http://hdl.handle.net/10388/8024

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

University of Pretoria

18. Hlatshwayo, Thulani Thokozani. Diffusion of silver in 6H-SiC.

Degree: Physics, 2011, University of Pretoria

URL: http://hdl.handle.net/2263/25616

► SiC is used as the main diffusion barrier in the fuel spheres of the pebble bed modular reactor (PBMR). The PBMR is a modern high… (more)

▼ SiC is used as the main diffusion barrier in the fuel spheres of the pebble bed modular reactor (PBMR). The PBMR is a modern high temperature nuclear reactor. However, the release of silver from the fuel spheres has raised some doubts about the effectiveness of this barrier, which has led to many studies on the possible migration paths of silver. The reported results of these studies have shown largely differing results concerning the magnitude and temperature dependence of silver being transported through the fuel particle coatings. Results from earlier investigations could be interpreted as a diffusion process governed by an Arrhenius type temperature dependence. In this study, the silver diffusion in 6H-SiC was investigated using two methods. In the first method a thin silver layer was deposited on 6H-SiC by vapour deposition while in the second method silver was implanted in 6H-SiC at room temperature, 350°C and 600°C to a fluence of 2×1016 silver ions cm-2. Finally the effect of neutron irradiation on the diffusion of silver was investigated for the samples implanted at 350°C and 600°C. Silver depth profiles before and after annealing were determined by Rutherford backscattering (RBS). Both isothermal and isochronal annealing were used in this study. Diffusion coefficients as well as detection limits were extracted by comparing the silver depth profiles before and after annealing. The radiation damage after implantation and their recovery after isothermal and isochronal annealing were analysed by Rutherford backscattering spectroscopy combined with channelling. The results of in-diffusion of silver into 6H-SiC at temperatures below the melting point (960°C) using un-encapsulated 6H-SiC samples with 100 nm deposited silver indicated no in-diffusion of silver; however, disappearance of silver occurred at these temperatures. For the encapsulated samples, no in-diffusion of silver was observed at 800°C, 900°C and 1000°C but silver disappeared from the samples’ surface and was found on the walls of the quartz glass ampoule. This disappearance of silver was established to be due to the wetting problem that existed between silver and SiC. The room temperature implantation resulted in a completely amorphous surface layer of approximately 270 nm thick. Epitaxial re-growth from the bulk was already taking place during annealing at 700°C and the crystalline structure seemed to be fully recovered at 1600°C, for samples that were sequentially isochronally annealed from 700°C in steps of 100°C up to 1600°C. However, no silver signal was detected at this temperature, which left certain doubts regarding the crystalline structure of the samples at this temperature. This was speculated to be due to thermal etching of the top original amorphous layer while the deeper amorphous layer was epitaxial re-growth from the bulk. The decomposition of SiC, giving rise to a carbon peak in the RBS spectra due to evaporation of Si, was clearly observed on the same samples at 1600°C. Isothermal annealing at 1300°C for 10 h cycles up to 80h… Advisors/Committee Members: Malherbe, Johan B. (advisor), Friedland, Erich Karl Helmuth (advisor).

Subjects/Keywords: Nuclear reactor; Fuel spheres; Diffusion barrier; UCTD

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APA (6^th Edition):

Hlatshwayo, T. T. (2011). Diffusion of silver in 6H-SiC. (Doctoral Dissertation). University of Pretoria. Retrieved from http://hdl.handle.net/2263/25616

Chicago Manual of Style (16^th Edition):

Hlatshwayo, Thulani Thokozani. “Diffusion of silver in 6H-SiC.” 2011. Doctoral Dissertation, University of Pretoria. Accessed March 03, 2021. http://hdl.handle.net/2263/25616.

MLA Handbook (7^th Edition):

Hlatshwayo, Thulani Thokozani. “Diffusion of silver in 6H-SiC.” 2011. Web. 03 Mar 2021.

Vancouver:

Hlatshwayo TT. Diffusion of silver in 6H-SiC. [Internet] [Doctoral dissertation]. University of Pretoria; 2011. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/2263/25616.

Council of Science Editors:

Hlatshwayo TT. Diffusion of silver in 6H-SiC. [Doctoral Dissertation]. University of Pretoria; 2011. Available from: http://hdl.handle.net/2263/25616

Oregon State University

19. Clark, R. G. (Robert Gilder), 1917-2005. Monitoring surface temperature of irradiated fuel elements.

Degree: MS, Electrical Engineering, 1963, Oregon State University

URL: http://hdl.handle.net/1957/48787

Subjects/Keywords: Nuclear fuel elements

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APA (6^th Edition):

Clark, R. G. (Robert Gilder), 1. (1963). Monitoring surface temperature of irradiated fuel elements. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/48787

Chicago Manual of Style (16^th Edition):

Clark, R. G. (Robert Gilder), 1917-2005. “Monitoring surface temperature of irradiated fuel elements.” 1963. Masters Thesis, Oregon State University. Accessed March 03, 2021. http://hdl.handle.net/1957/48787.

MLA Handbook (7^th Edition):

Clark, R. G. (Robert Gilder), 1917-2005. “Monitoring surface temperature of irradiated fuel elements.” 1963. Web. 03 Mar 2021.

Vancouver:

Clark, R. G. (Robert Gilder) 1. Monitoring surface temperature of irradiated fuel elements. [Internet] [Masters thesis]. Oregon State University; 1963. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/1957/48787.

Council of Science Editors:

Clark, R. G. (Robert Gilder) 1. Monitoring surface temperature of irradiated fuel elements. [Masters Thesis]. Oregon State University; 1963. Available from: http://hdl.handle.net/1957/48787

Oregon State University

20. Latimer, Griffen D. Vibration Analysis of Advanced Test Reactor Miniplate Hydraulic Test in the OSU HMFTF.

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

URL: http://hdl.handle.net/1957/60099

► Current research on the topic of advanced reactor fuel types include the use of ultrahigh density Uranium-Molybdenum fuels, towards their use in high-performance research reactors.… (more)

Subjects/Keywords: Flow Induced Vibration; Nuclear fuel elements – Testing

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APA (6^th Edition):

Latimer, G. D. (2016). Vibration Analysis of Advanced Test Reactor Miniplate Hydraulic Test in the OSU HMFTF. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/60099

Chicago Manual of Style (16^th Edition):

Latimer, Griffen D. “Vibration Analysis of Advanced Test Reactor Miniplate Hydraulic Test in the OSU HMFTF.” 2016. Masters Thesis, Oregon State University. Accessed March 03, 2021. http://hdl.handle.net/1957/60099.

MLA Handbook (7^th Edition):

Latimer, Griffen D. “Vibration Analysis of Advanced Test Reactor Miniplate Hydraulic Test in the OSU HMFTF.” 2016. Web. 03 Mar 2021.

Vancouver:

Latimer GD. Vibration Analysis of Advanced Test Reactor Miniplate Hydraulic Test in the OSU HMFTF. [Internet] [Masters thesis]. Oregon State University; 2016. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/1957/60099.

Council of Science Editors:

Latimer GD. Vibration Analysis of Advanced Test Reactor Miniplate Hydraulic Test in the OSU HMFTF. [Masters Thesis]. Oregon State University; 2016. Available from: http://hdl.handle.net/1957/60099

University of Ontario Institute of Technology

21. Peiman, Wargha. Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs).

Degree: 2017, University of Ontario Institute of Technology

URL: http://hdl.handle.net/10155/827

► A group of countries has initiated an international collaboration to develop a next generation (i.e., Generation IV) of nuclear reactors. Chosen as one of the… (more)

▼ A group of countries has initiated an international collaboration to develop a next generation (i.e., Generation IV) of nuclear reactors. Chosen as one of the six Generation???IV nuclear-reactor concepts, the SCWRs are expected to have high thermal efficiencies within the range of 40 ??? 50% owing to reactor???s high outlet temperatures. The Canadian pressure-tube-type SCWR is featured with 3-batch refueling, 336 vertical fuel channels, a porous ceramic insulator inside the pressure tube, and stainless-steel cladding. The reactor operates at a pressure of 25 MPa with the coolant temperature rising from 350 to 625??C. Consequently, sheath and fuel centerline temperatures are significantly higher in SCWRs compared to those of the current water-cooled nuclear reactors. The main objective of this thesis is to conduct a study on specifics of the thermalhydraulics and neutronics of a pressure-tube SCWR based on an understanding of the supercritical water phenomena and their impacts on reactor design and operation. This thesis investigates the impact of several thermalhydraulic modeling parameters on fuel and cladding temperatures of a pressure-tube SCWR. The investigated thermalhydraulic modeling parameters are: 1) variable heat transfer coefficient, which is affected by thermophysical properties of supercritical water, axial heat flux, and three heat-transfer regimes: normal, improved and deteriorated; 2) thermophysical properties, which are affected by the bulk-fluid-temperature profile along the heated length and pressure drop along the fuel channel; 3) variable axial and radial heat-flux profiles of a fuel assembly (bundle string), which are affected by the neutron flux; 4) radial non-uniform heat generation inside the fuel; 5) axial and radial variable thermal conductivity of a fuel; 6) contact thermal resistance between the fuel and cladding; 7) heat loss from the coolant to the moderator, which is affected by the thermal conductivity of a ceramic insert; and 8) pressure drop of the coolant along the fuel channel. The main neutronic aspects, which have been incorporated in the neutronic model, include 1) variable coolant density along the heated length of the fuel channel, which affects neutronic properties of a lattice and, hence, the neutron flux and 2) number of energy groups, which affects the calculated channel powers. Advisors/Committee Members: Pioro, Igor, Gabriel, Kamiel.

Subjects/Keywords: SCWR; Thermalhydraulics; Neutronics; Generation IV; Nuclear fuel

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APA (6^th Edition):

Peiman, W. (2017). Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs). (Thesis). University of Ontario Institute of Technology. Retrieved from http://hdl.handle.net/10155/827

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):

Peiman, Wargha. “Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs).” 2017. Thesis, University of Ontario Institute of Technology. Accessed March 03, 2021. http://hdl.handle.net/10155/827.

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

MLA Handbook (7^th Edition):

Peiman, Wargha. “Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs).” 2017. Web. 03 Mar 2021.

Vancouver:

Peiman W. Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs). [Internet] [Thesis]. University of Ontario Institute of Technology; 2017. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/10155/827.

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

Council of Science Editors:

Peiman W. Study on specifics of thermalhydraulics and neutronics of pressure-channel supercritical water-cooled reactors (SCWRs). [Thesis]. University of Ontario Institute of Technology; 2017. Available from: http://hdl.handle.net/10155/827

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

University of Tennessee – Knoxville

22. Littell, Jennifer Lynn. Development and Application of Nuclear Fuel Cycle Simulators for Evaluating Potential Fuel Cycle Options.

Degree: MS, Nuclear Engineering, 2016, University of Tennessee – Knoxville

URL: https://trace.tennessee.edu/utk_gradthes/3783

► The Nuclear Fuel Cycle Evaluation and Screening Study was chartered by the DOE in order to weigh the relative benefits and challenges of potential… (more)

▼ The Nuclear Fuel Cycle Evaluation and Screening Study was chartered by the DOE in order to weigh the relative benefits and challenges of potential future fuel cycle options. In order to efficiently implement these alternative fuel cycles, the transition from the current once-through cycle to the most promising of these potential fuel cycles must also be analyzed. This analysis requires the use of fuel cycle simulators which have the capability to quickly calculate the mass flows between numerous facilities over hundreds of years. In this work, Cyclus and ORION have both been utilized to simulate transitions from the current once-through fuel cycle to one which involves fast reactors with continuous reprocessing of spent fuel. This transition was found to take approximately 140 years while staying within the constraints of maintaining the mass of excess plutonium in storage below 100 tonnes, introducing fast reactors gradually in the first years, and waiting until 2050 to begin reprocessing. Before completing this transition analysis, Cyclus was also used to create a handful of less sophisticated simulations in order to demonstrate its range of capabilities. In addition to using Cyclus to contribute to the Evaluation and Screening Study, this work contains the beginning of an ORIGEN-based repository of modules for use with Cyclus. This repository, called CyBORG, incorporates ORIGEN's isotopic depletion and decay calculations directly into Cyclus. The first module added to CyBORG is a reactor facility which uses ORIGEN to calculate its spent fuel isotopics based on reactor specifications from the user such as assembly type, fresh fuel recipe, and power capacity. By creating problem-specific cross section libraries for the depletion calculations, combined with ORIGEN's capability to track more than 2000 isotopes, accurate spent fuel isotopics can be created which will reflect how any changes to the system affect the availability of fissile material. Advisors/Committee Members: Steven E. Skutnik, Ivan G. Maldanado, Ondrej Chvala.

Subjects/Keywords: nuclear fuel cycle; ORIGEN; Cyclus; ORION

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APA (6^th Edition):

Littell, J. L. (2016). Development and Application of Nuclear Fuel Cycle Simulators for Evaluating Potential Fuel Cycle Options. (Thesis). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_gradthes/3783

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):

Littell, Jennifer Lynn. “Development and Application of Nuclear Fuel Cycle Simulators for Evaluating Potential Fuel Cycle Options.” 2016. Thesis, University of Tennessee – Knoxville. Accessed March 03, 2021. https://trace.tennessee.edu/utk_gradthes/3783.

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

MLA Handbook (7^th Edition):

Littell, Jennifer Lynn. “Development and Application of Nuclear Fuel Cycle Simulators for Evaluating Potential Fuel Cycle Options.” 2016. Web. 03 Mar 2021.

Vancouver:

Littell JL. Development and Application of Nuclear Fuel Cycle Simulators for Evaluating Potential Fuel Cycle Options. [Internet] [Thesis]. University of Tennessee – Knoxville; 2016. [cited 2021 Mar 03]. Available from: https://trace.tennessee.edu/utk_gradthes/3783.

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

Council of Science Editors:

Littell JL. Development and Application of Nuclear Fuel Cycle Simulators for Evaluating Potential Fuel Cycle Options. [Thesis]. University of Tennessee – Knoxville; 2016. Available from: https://trace.tennessee.edu/utk_gradthes/3783

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

University of Cambridge

23. Charles, Alan. Development of a Multi-Objective Optimization Capability for Heterogeneous Light Water Reactor Fuel Assemblies.

Degree: PhD, 2020, University of Cambridge

URL: https://www.repository.cam.ac.uk/handle/1810/315342

► As pressure grows on developed nations to move away from fossil fuel-based energy sources, so does the potential for nuclear energy to make its resurgence.… (more)

Subjects/Keywords: Nuclear; Optimization; Differential Evolution; Fuel Assembly Design

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APA (6^th Edition):

Charles, A. (2020). Development of a Multi-Objective Optimization Capability for Heterogeneous Light Water Reactor Fuel Assemblies. (Doctoral Dissertation). University of Cambridge. Retrieved from https://www.repository.cam.ac.uk/handle/1810/315342

Chicago Manual of Style (16^th Edition):

Charles, Alan. “Development of a Multi-Objective Optimization Capability for Heterogeneous Light Water Reactor Fuel Assemblies.” 2020. Doctoral Dissertation, University of Cambridge. Accessed March 03, 2021. https://www.repository.cam.ac.uk/handle/1810/315342.

MLA Handbook (7^th Edition):

Charles, Alan. “Development of a Multi-Objective Optimization Capability for Heterogeneous Light Water Reactor Fuel Assemblies.” 2020. Web. 03 Mar 2021.

Vancouver:

Charles A. Development of a Multi-Objective Optimization Capability for Heterogeneous Light Water Reactor Fuel Assemblies. [Internet] [Doctoral dissertation]. University of Cambridge; 2020. [cited 2021 Mar 03]. Available from: https://www.repository.cam.ac.uk/handle/1810/315342.

Council of Science Editors:

Charles A. Development of a Multi-Objective Optimization Capability for Heterogeneous Light Water Reactor Fuel Assemblies. [Doctoral Dissertation]. University of Cambridge; 2020. Available from: https://www.repository.cam.ac.uk/handle/1810/315342

24. Merwin, Augustus. Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide.

Degree: 2012, University of Nevada – Reno

URL: http://hdl.handle.net/11714/3744

► Reprocessing of spent nuclear fuel is an essential step in closing the nuclear fuel cycle. In order to consume current stockpiles, ceramic uranium dioxide spent… (more)

▼ Reprocessing of spent nuclear fuel is an essential step in closing the nuclear fuel cycle. In order to consume current stockpiles, ceramic uranium dioxide spent nuclear fuel will be subjected to an electrolytic reduction process. The current reduction process employs a platinum anode and a stainless steel alloy 316 cathode in a molten salt bath consisting of LiCl-2wt% Li2O and occurs at 700⁰C. A major shortcoming of the existing process is the degradation of the platinum anode under the severely oxidizing conditions encountered during electrolytic reduction. This work investigates alternative anode materials for the electrolytic reduction of uranium oxide.The high temperature and extreme oxidizing conditions encountered in these studies necessitated a unique set of design constraints on the system. Thus, a customized experimental apparatus was designed and constructed. The electrochemical experiments were performed in an electrochemical reactor placed inside a furnace. This entire setup was housed inside a glove box, in order to maintain an inert atmosphere.This study investigates alternative anode materials through accelerated corrosion testing. Surface morphology was studied using scanning electron microscopy. Surface chemistry was characterized using energy dispersive spectroscopy and Raman spectroscopy. Electrochemical behavior of candidate materials was evaluated using potentiodynamic polarization characteristics. After narrowing the number of candidate electrode materials, ferrous stainless steel alloy 316, nickel based Inconel 718 and elemental tungsten were chosen for further investigation. Of these materials only tungsten was found to be sufficiently stable at the anodic potential required for electrolysis of uranium dioxide in molten salt. The tungsten anode and stainless steel alloy 316 cathode electrode system was studied at the required reduction potential for UO2 with varying lithium oxide concentrations. Electrochemical impedance spectroscopy showed mixed (kinetic and diffusion) control and an overall low impedance due to extreme corrosion. It was observed that tungsten is sufficiently stable in LiCl - 2wt% Li2O at 700⁰C at the required anodic potential for the reduction of uranium oxide. This study identifies tungsten to be a superior anode material to platinum for the electrolytic reduction of uranium oxide, both in terms of superior corrosion behavior and reduced cost, and thus recommends that tungsten be further investigated as an alternative anode for the electrolytic reduction of uranium dioxide. Advisors/Committee Members: Chidambaram, Dev (advisor), Tsoulfanidis, Nicholas (committee member), Raja, Krishnan (committee member), Greiner, Miles (committee member).

Subjects/Keywords: Corrosion; molten salt; nuclear fuel cycle; pyroprocessing

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APA (6^th Edition):

Merwin, A. (2012). Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide. (Thesis). University of Nevada – Reno. Retrieved from http://hdl.handle.net/11714/3744

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):

Merwin, Augustus. “Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide.” 2012. Thesis, University of Nevada – Reno. Accessed March 03, 2021. http://hdl.handle.net/11714/3744.

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

MLA Handbook (7^th Edition):

Merwin, Augustus. “Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide.” 2012. Web. 03 Mar 2021.

Vancouver:

Merwin A. Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide. [Internet] [Thesis]. University of Nevada – Reno; 2012. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/11714/3744.

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

Council of Science Editors:

Merwin A. Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide. [Thesis]. University of Nevada – Reno; 2012. Available from: http://hdl.handle.net/11714/3744

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

University of Cambridge

25. Charles, Alan. Development of a multi-objective optimization capability for heterogeneous light water reactor fuel assemblies.

Degree: PhD, 2020, University of Cambridge

URL: https://doi.org/10.17863/CAM.62450 ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.821619

► As pressure grows on developed nations to move away from fossil fuel-based energy sources, so does the potential for nuclear energy to make its resurgence.… (more)

Subjects/Keywords: Nuclear; Optimization; Differential Evolution; Fuel Assembly Design

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APA (6^th Edition):

Charles, A. (2020). Development of a multi-objective optimization capability for heterogeneous light water reactor fuel assemblies. (Doctoral Dissertation). University of Cambridge. Retrieved from https://doi.org/10.17863/CAM.62450 ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.821619

Chicago Manual of Style (16^th Edition):

Charles, Alan. “Development of a multi-objective optimization capability for heterogeneous light water reactor fuel assemblies.” 2020. Doctoral Dissertation, University of Cambridge. Accessed March 03, 2021. https://doi.org/10.17863/CAM.62450 ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.821619.

MLA Handbook (7^th Edition):

Charles, Alan. “Development of a multi-objective optimization capability for heterogeneous light water reactor fuel assemblies.” 2020. Web. 03 Mar 2021.

Vancouver:

Charles A. Development of a multi-objective optimization capability for heterogeneous light water reactor fuel assemblies. [Internet] [Doctoral dissertation]. University of Cambridge; 2020. [cited 2021 Mar 03]. Available from: https://doi.org/10.17863/CAM.62450 ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.821619.

Council of Science Editors:

Charles A. Development of a multi-objective optimization capability for heterogeneous light water reactor fuel assemblies. [Doctoral Dissertation]. University of Cambridge; 2020. Available from: https://doi.org/10.17863/CAM.62450 ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.821619

26. Vieira, Edeval. Estudo paramétrico da deformação de placas combutíveis com núcleos de dispersão de U3Si2-AI.

Degree: Mestrado, Tecnologia Nuclear - Materiais, 2011, University of São Paulo

URL: http://www.teses.usp.br/teses/disponiveis/85/85134/tde-02032012-133617/ ;

►

O Instituto de Pesquisas Energéticas e Nucleares IPEN-CNEN/SP atualmente produz rotineiramente o combustível nuclear necessário para a operação de seu reator de pesquisas IEA-R1. Esse… (more)

▼

O Instituto de Pesquisas Energéticas e Nucleares IPEN-CNEN/SP atualmente produz rotineiramente o combustível nuclear necessário para a operação de seu reator de pesquisas IEA-R1. Esse combustível é formado por placas combustíveis contendo núcleos de compósitos U3Si2-Al, obtidas por laminação. O processo de laminação atualmente implantado foi desenvolvido com base em informações obtidas na literatura, as quais foram usadas como premissas para a definição dos atuais procedimentos de fabricação, segundo uma metodologia de caráter essencialmente empírico. Apesar do processo de laminação atual estar perfeitamente estável e reprodutível, ele não é totalmente conhecido. O objetivo deste trabalho é caracterizar o processo de laminação de placas combustíveis adotado pelo IPEN, especificamente no que se refere à evolução dos parâmetros dimensionais da placa combustível em função da sua deformação no processo de laminação. Estão apresentados resultados da evolução das espessuras do núcleo e revestimentos da placa combustível ao longo da sua deformação, assim como dos defeitos terminais, microestrutura e porosidade do núcleo.

The Nuclear and Energy Research Institute - IPEN-CNEN/SP produces routinely the nuclear fuel necessary for operating its research reactor, IEA-R1. This fuel consists of fuel plates containing U3Si2-Al composites as the meat, which are fabricated by rolling. The rolling process currently deployed was developed with base on information obtained from literature, which were used as premises for defining the current manufacturing procedures, according to a methodology with essentially empirical character. Despite the current rolling process to be perfectly stable and highly reproducible, it is not well characterized and therefore is not fully known. The objective of this work is to characterize the rolling process for producing fuel plates, specifically the evolution of dimensional parameters of the fuel plate as a function of its deformation in the rolling process. Results are presented in terms of the evolution of the thickness of the fuel meat and cladding of the fuel plate along the deformation, as well as the terminals defects, microstructure and porosity of the fuel meat.

Advisors/Committee Members: Durazzo, Michelangelo.

Subjects/Keywords: combustível nuclear; deformação; deformation; dispersão; dispersion; fuel plates; nuclear fuel; placas combustíveis

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

APA (6^th Edition):

Vieira, E. (2011). Estudo paramétrico da deformação de placas combutíveis com núcleos de dispersão de U3Si2-AI. (Masters Thesis). University of São Paulo. Retrieved from http://www.teses.usp.br/teses/disponiveis/85/85134/tde-02032012-133617/ ;

Chicago Manual of Style (16^th Edition):

Vieira, Edeval. “Estudo paramétrico da deformação de placas combutíveis com núcleos de dispersão de U3Si2-AI.” 2011. Masters Thesis, University of São Paulo. Accessed March 03, 2021. http://www.teses.usp.br/teses/disponiveis/85/85134/tde-02032012-133617/ ;.

MLA Handbook (7^th Edition):

Vieira, Edeval. “Estudo paramétrico da deformação de placas combutíveis com núcleos de dispersão de U3Si2-AI.” 2011. Web. 03 Mar 2021.

Vancouver:

Vieira E. Estudo paramétrico da deformação de placas combutíveis com núcleos de dispersão de U3Si2-AI. [Internet] [Masters thesis]. University of São Paulo; 2011. [cited 2021 Mar 03]. Available from: http://www.teses.usp.br/teses/disponiveis/85/85134/tde-02032012-133617/ ;.

Council of Science Editors:

Vieira E. Estudo paramétrico da deformação de placas combutíveis com núcleos de dispersão de U3Si2-AI. [Masters Thesis]. University of São Paulo; 2011. Available from: http://www.teses.usp.br/teses/disponiveis/85/85134/tde-02032012-133617/ ;

Brno University of Technology

27. Ježek, Martin. Palivový cyklus jaderné elektrárny Temelín: Nuclear Fuel Cycle of Temelin NPP.

Degree: 2019, Brno University of Technology

URL: http://hdl.handle.net/11012/11531

► This bachelor's thesis deals with the central part of a fuel cycle in the power plant of Temelín. It describes generally the whole fuel cycle,… (more)

Subjects/Keywords: palivový cyklus; jaderné palivo; Temelín; VVER; Nuclear fuel cycle; nuclear fuel; Temelin; VVER

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

APA (6^th Edition):

Ježek, M. (2019). Palivový cyklus jaderné elektrárny Temelín: Nuclear Fuel Cycle of Temelin NPP. (Thesis). Brno University of Technology. Retrieved from http://hdl.handle.net/11012/11531

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):

Ježek, Martin. “Palivový cyklus jaderné elektrárny Temelín: Nuclear Fuel Cycle of Temelin NPP.” 2019. Thesis, Brno University of Technology. Accessed March 03, 2021. http://hdl.handle.net/11012/11531.

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

MLA Handbook (7^th Edition):

Ježek, Martin. “Palivový cyklus jaderné elektrárny Temelín: Nuclear Fuel Cycle of Temelin NPP.” 2019. Web. 03 Mar 2021.

Vancouver:

Ježek M. Palivový cyklus jaderné elektrárny Temelín: Nuclear Fuel Cycle of Temelin NPP. [Internet] [Thesis]. Brno University of Technology; 2019. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/11012/11531.

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

Council of Science Editors:

Ježek M. Palivový cyklus jaderné elektrárny Temelín: Nuclear Fuel Cycle of Temelin NPP. [Thesis]. Brno University of Technology; 2019. Available from: http://hdl.handle.net/11012/11531

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

Penn State University

28. Bratton, Ryan Nathaniel. Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure.

Degree: 2015, Penn State University

URL: https://submit-etda.libraries.psu.edu/catalog/27289

► The bias and uncertainty in fuel isotopic calculations for a well-defined radiochemical assay benchmark are investigated with Sampler, the new sampling-based uncertainty quantification tool in… (more)

▼ The bias and uncertainty in fuel isotopic calculations for a well-defined radiochemical assay benchmark are investigated with Sampler, the new sampling-based uncertainty quantification tool in the SCALE code system. Isotopic predictions are compared to measurements of fuel rod MKP109 of assembly D047 from the Calvert Cliffs Unit 1 core at three axial locations, representing a range of discharged fuel burnups. A methodology is developed which quantifies the significance of input parameter uncertainties and modeling decisions on isotopic prediction by comparing to isotopic measurement uncertainties. The SCALE Sampler model of the D047 assembly incorporates input parameter uncertainties for key input data such as multigroup cross sections, decay constants, fission product yields, the cladding thickness, and the power history for fuel rod MKP109. The effects of each set of input parameter uncertainty on the uncertainty of isotopic predictions have been quantified. In this work, isotopic prediction biases are identified and an investigation into their sources is proposed; namely, biases have been identified for certain plutonium, europium, and gadolinium isotopes for all three axial locations. Moreover, isotopic prediction uncertainty resulting from only nuclear data is found to be greatest for Eu-154, Gd-154, and Gd-160. The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed which tracks inter-cycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each considered WBN1 fuel rod. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layers is derived and applied to FRAPCON output data to quantify the RIP and CHS for these fuel rods. SCALE/Polaris is used to quantify fuel rod-specific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel blankets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd/MTU is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be… Advisors/Committee Members: Kostadin Nikolov Ivanov, Dissertation Advisor/Co-Advisor, Igor Jovanovic, Committee Chair/Co-Chair, Monique Yvonne Yaari, Committee Member, Maria Nikolova Avramova, Committee Member, Matthew A Jessee, Special Member, William A Wieselquist, Special Member.

Subjects/Keywords: SCALE; FRAPCON; rod internal pressure; fuel isotopics; spent nuclear fuel

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

APA (6^th Edition):

Bratton, R. N. (2015). Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure. (Thesis). Penn State University. Retrieved from https://submit-etda.libraries.psu.edu/catalog/27289

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):

Bratton, Ryan Nathaniel. “Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure.” 2015. Thesis, Penn State University. Accessed March 03, 2021. https://submit-etda.libraries.psu.edu/catalog/27289.

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

MLA Handbook (7^th Edition):

Bratton, Ryan Nathaniel. “Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure.” 2015. Web. 03 Mar 2021.

Vancouver:

Bratton RN. Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure. [Internet] [Thesis]. Penn State University; 2015. [cited 2021 Mar 03]. Available from: https://submit-etda.libraries.psu.edu/catalog/27289.

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

Council of Science Editors:

Bratton RN. Uncertainty Analysis of Spent Nuclear Fuel Isotopics and Rod Internal Pressure. [Thesis]. Penn State University; 2015. Available from: https://submit-etda.libraries.psu.edu/catalog/27289

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

Georgia Tech

29. Kingsbury, Christopher W. Fuel cycle cost and fabrication model for fluoride-salt high-temperature reactor (FHR) "Plank" fuel design optimization.

Degree: MS, Mechanical Engineering, 2015, Georgia Tech

URL: http://hdl.handle.net/1853/54337

► The fluoride-salt-cooled high-temperature reactor (FHR) is a novel reactor design benefitting from passive safety features, high operating temperatures with corresponding high conversion efficiency, to name… (more)

Subjects/Keywords: Nuclear fuel design; Fuel cycle cost; Advanced high temperature reactors

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

APA (6^th Edition):

Kingsbury, C. W. (2015). Fuel cycle cost and fabrication model for fluoride-salt high-temperature reactor (FHR) "Plank" fuel design optimization. (Masters Thesis). Georgia Tech. Retrieved from http://hdl.handle.net/1853/54337

Chicago Manual of Style (16^th Edition):

Kingsbury, Christopher W. “Fuel cycle cost and fabrication model for fluoride-salt high-temperature reactor (FHR) "Plank" fuel design optimization.” 2015. Masters Thesis, Georgia Tech. Accessed March 03, 2021. http://hdl.handle.net/1853/54337.

MLA Handbook (7^th Edition):

Kingsbury, Christopher W. “Fuel cycle cost and fabrication model for fluoride-salt high-temperature reactor (FHR) "Plank" fuel design optimization.” 2015. Web. 03 Mar 2021.

Vancouver:

Kingsbury CW. Fuel cycle cost and fabrication model for fluoride-salt high-temperature reactor (FHR) "Plank" fuel design optimization. [Internet] [Masters thesis]. Georgia Tech; 2015. [cited 2021 Mar 03]. Available from: http://hdl.handle.net/1853/54337.

Council of Science Editors:

Kingsbury CW. Fuel cycle cost and fabrication model for fluoride-salt high-temperature reactor (FHR) "Plank" fuel design optimization. [Masters Thesis]. Georgia Tech; 2015. Available from: http://hdl.handle.net/1853/54337

University of South Carolina

30. Metzger, Kathryn Elizabeth. Fabrication and Characterization of Surrogate Fuel Particles Using the Spark Erosion Method.

Degree: MS, Nuclear Engineering, 2013, University of South Carolina

URL: https://scholarcommons.sc.edu/etd/2368

► In light of the disaster at the Fukushima Daiichi Nuclear Plant, the Department of Energy's Advanced Fuels Program has shifted its interest from enhanced… (more)

▼ In light of the disaster at the Fukushima Daiichi Nuclear Plant, the Department of Energy's Advanced Fuels Program has shifted its interest from enhanced performance fuels to enhanced accident tolerance fuels. Dispersion fuels possess higher thermal conductivities than traditional light water reactor fuel and as a result, offer improved safety margins. The benefits of a dispersion fuel are due to the presence of the secondary non-fissile phase (matrix), which serves as a barrier to fission products and improves the overall thermal performance of the fuel. However, the presence of a matrix material reduces the fuel volume, which lowers the fissile content of dispersion. This issue can be remedied through the development of higher density fuel phases or through an optimization of fuel particle size and volume loading. The latter requirement necessitates the development of fabrication methods to produce small, micron-order fuel particles. This research examines the capabilities of the spark erosion process to fabricate particles on the order of 10 μm. A custom-built spark erosion device by CT Electromechanica was used to produce stainless steel surrogate fuel particles in a deionized water dielectric. Three arc intensities were evaluated to determine the effect on particle size. Particles were filtered from the dielectric using a polycarbonate membrane filter and vacuum filtration system. Fabricated particles were characterized via field emission scanning electron microscopy (FESEM), laser light particle size analysis, energy-dispersive spectroscopy (EDS), X-ray diffraction analysis (XRD), and gas pycnometry. FESEM images reveal that the spark erosion process produces highly spherical particles on the order of 10 microns. These findings are substantiated by the results of particle size analysis. Additionally, EDS and XRD results indicate the presence of oxide phases, which suggests the dielectric reacted with the molten debris during particle formation. Advisors/Committee Members: Travis W Knight.

Subjects/Keywords: Engineering; Nuclear Engineering; Dispersion Fuel; Fabrication; Fuel Particles; Spark Erosion

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

APA (6^th Edition):

Metzger, K. E. (2013). Fabrication and Characterization of Surrogate Fuel Particles Using the Spark Erosion Method. (Masters Thesis). University of South Carolina. Retrieved from https://scholarcommons.sc.edu/etd/2368

Chicago Manual of Style (16^th Edition):

Metzger, Kathryn Elizabeth. “Fabrication and Characterization of Surrogate Fuel Particles Using the Spark Erosion Method.” 2013. Masters Thesis, University of South Carolina. Accessed March 03, 2021. https://scholarcommons.sc.edu/etd/2368.

MLA Handbook (7^th Edition):

Metzger, Kathryn Elizabeth. “Fabrication and Characterization of Surrogate Fuel Particles Using the Spark Erosion Method.” 2013. Web. 03 Mar 2021.

Vancouver:

Metzger KE. Fabrication and Characterization of Surrogate Fuel Particles Using the Spark Erosion Method. [Internet] [Masters thesis]. University of South Carolina; 2013. [cited 2021 Mar 03]. Available from: https://scholarcommons.sc.edu/etd/2368.

Council of Science Editors:

Metzger KE. Fabrication and Characterization of Surrogate Fuel Particles Using the Spark Erosion Method. [Masters Thesis]. University of South Carolina; 2013. Available from: https://scholarcommons.sc.edu/etd/2368

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