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You searched for subject:(Nuclear reactor). Showing records 1 – 30 of 394 total matches.

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Texas A&M University

1. Kumar, Akansha. An Iterative Optimization Method Using Genetic Algorithms and Gaussian Process Based Regression in Nuclear Reactor Design Applications.

Degree: PhD, Nuclear Engineering, 2016, Texas A&M University

 The optimization of a complex system involves the determination of optimum values for a set of design parameters. The optimization search happens in order to… (more)

Subjects/Keywords: genetic algorithms; optimization; nuclear; reactor

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APA (6th Edition):

Kumar, A. (2016). An Iterative Optimization Method Using Genetic Algorithms and Gaussian Process Based Regression in Nuclear Reactor Design Applications. (Doctoral Dissertation). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/159121

Chicago Manual of Style (16th Edition):

Kumar, Akansha. “An Iterative Optimization Method Using Genetic Algorithms and Gaussian Process Based Regression in Nuclear Reactor Design Applications.” 2016. Doctoral Dissertation, Texas A&M University. Accessed April 19, 2019. http://hdl.handle.net/1969.1/159121.

MLA Handbook (7th Edition):

Kumar, Akansha. “An Iterative Optimization Method Using Genetic Algorithms and Gaussian Process Based Regression in Nuclear Reactor Design Applications.” 2016. Web. 19 Apr 2019.

Vancouver:

Kumar A. An Iterative Optimization Method Using Genetic Algorithms and Gaussian Process Based Regression in Nuclear Reactor Design Applications. [Internet] [Doctoral dissertation]. Texas A&M University; 2016. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/1969.1/159121.

Council of Science Editors:

Kumar A. An Iterative Optimization Method Using Genetic Algorithms and Gaussian Process Based Regression in Nuclear Reactor Design Applications. [Doctoral Dissertation]. Texas A&M University; 2016. Available from: http://hdl.handle.net/1969.1/159121


The Ohio State University

2. Kennedy, Ryanne Ariel. Quantifying Uncertainty in Reactor Flux/Power Distributions.

Degree: PhD, Nuclear Engineering, 2011, The Ohio State University

 The design and development of a conceptual system for power measurements in a reactor core using in-core sensors has been an ongoing focus of research… (more)

Subjects/Keywords: Nuclear Engineering; nuclear reactor; flux; uncertainty quantification; reactor physics

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APA (6th Edition):

Kennedy, R. A. (2011). Quantifying Uncertainty in Reactor Flux/Power Distributions. (Doctoral Dissertation). The Ohio State University. Retrieved from http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901

Chicago Manual of Style (16th Edition):

Kennedy, Ryanne Ariel. “Quantifying Uncertainty in Reactor Flux/Power Distributions.” 2011. Doctoral Dissertation, The Ohio State University. Accessed April 19, 2019. http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901.

MLA Handbook (7th Edition):

Kennedy, Ryanne Ariel. “Quantifying Uncertainty in Reactor Flux/Power Distributions.” 2011. Web. 19 Apr 2019.

Vancouver:

Kennedy RA. Quantifying Uncertainty in Reactor Flux/Power Distributions. [Internet] [Doctoral dissertation]. The Ohio State University; 2011. [cited 2019 Apr 19]. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901.

Council of Science Editors:

Kennedy RA. Quantifying Uncertainty in Reactor Flux/Power Distributions. [Doctoral Dissertation]. The Ohio State University; 2011. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1306360901


McMaster University

3. McDonald, Michael H. Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor.

Degree: MASc, 2011, McMaster University

The supercritical water cooled reactor (SCWR) is a Generation IV reactor concept that features light water coolant in a supercritical state. Canada is developing… (more)

Subjects/Keywords: Reactor physics; supercritical water cooled reactor; fuel; Nuclear Engineering; Nuclear Engineering

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APA (6th Edition):

McDonald, M. H. (2011). Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor. (Masters Thesis). McMaster University. Retrieved from http://hdl.handle.net/11375/11223

Chicago Manual of Style (16th Edition):

McDonald, Michael H. “Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor.” 2011. Masters Thesis, McMaster University. Accessed April 19, 2019. http://hdl.handle.net/11375/11223.

MLA Handbook (7th Edition):

McDonald, Michael H. “Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor.” 2011. Web. 19 Apr 2019.

Vancouver:

McDonald MH. Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor. [Internet] [Masters thesis]. McMaster University; 2011. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/11375/11223.

Council of Science Editors:

McDonald MH. Fuel and Core Physics Considerations for a Pressure Tube Supercritical Water Cooled Reactor. [Masters Thesis]. McMaster University; 2011. Available from: http://hdl.handle.net/11375/11223

4. Tondin, Julio Benedito Marin. Desenvolvimento de um programa computacional para gerenciamento de banco de dados de material nuclear.

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

Em instalações nucleares o controle do material nuclear é uma das atividades da maior importância. A Comissão Nacional de Energia Nuclear (CNEN) e a Agencia… (more)

Subjects/Keywords: material nuclear; nuclear material; nuclear reactor; reator nuclear; safeguard; salvaguarda

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APA (6th Edition):

Tondin, J. B. M. (2011). Desenvolvimento de um programa computacional para gerenciamento de banco de dados de material nuclear. (Masters Thesis). University of São Paulo. Retrieved from http://www.teses.usp.br/teses/disponiveis/85/85131/tde-03042012-110437/ ;

Chicago Manual of Style (16th Edition):

Tondin, Julio Benedito Marin. “Desenvolvimento de um programa computacional para gerenciamento de banco de dados de material nuclear.” 2011. Masters Thesis, University of São Paulo. Accessed April 19, 2019. http://www.teses.usp.br/teses/disponiveis/85/85131/tde-03042012-110437/ ;.

MLA Handbook (7th Edition):

Tondin, Julio Benedito Marin. “Desenvolvimento de um programa computacional para gerenciamento de banco de dados de material nuclear.” 2011. Web. 19 Apr 2019.

Vancouver:

Tondin JBM. Desenvolvimento de um programa computacional para gerenciamento de banco de dados de material nuclear. [Internet] [Masters thesis]. University of São Paulo; 2011. [cited 2019 Apr 19]. Available from: http://www.teses.usp.br/teses/disponiveis/85/85131/tde-03042012-110437/ ;.

Council of Science Editors:

Tondin JBM. Desenvolvimento de um programa computacional para gerenciamento de banco de dados de material nuclear. [Masters Thesis]. University of São Paulo; 2011. Available from: http://www.teses.usp.br/teses/disponiveis/85/85131/tde-03042012-110437/ ;

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

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

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)

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

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APA (6th 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 (16th Edition):

Faêda, Kelly Cristina Martins. “Caracterização do combustível para reatores nucleares produtores de hidrogênio.” 2011. Masters Thesis, Centro de Desenvolvimento da Tecnologia Nuclear. Accessed April 19, 2019. http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=139 ;.

MLA Handbook (7th Edition):

Faêda, Kelly Cristina Martins. “Caracterização do combustível para reatores nucleares produtores de hidrogênio.” 2011. Web. 19 Apr 2019.

Vancouver:

Faêda KCM. Caracterização do combustível para reatores nucleares produtores de hidrogênio. [Internet] [Masters thesis]. Centro de Desenvolvimento da Tecnologia Nuclear; 2011. [cited 2019 Apr 19]. 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 ;


University of Ontario Institute of Technology

6. Patel, Amin. Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices.

Degree: 2010, University of Ontario Institute of Technology

 Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several… (more)

Subjects/Keywords: Applied reactor physics; Transport theory; Diffusion theory; CANDU; Nuclear reactor

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APA (6th Edition):

Patel, A. (2010). Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices. (Thesis). University of Ontario Institute of Technology. Retrieved from http://hdl.handle.net/10155/87

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

Chicago Manual of Style (16th Edition):

Patel, Amin. “Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices.” 2010. Thesis, University of Ontario Institute of Technology. Accessed April 19, 2019. http://hdl.handle.net/10155/87.

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

MLA Handbook (7th Edition):

Patel, Amin. “Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices.” 2010. Web. 19 Apr 2019.

Vancouver:

Patel A. Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices. [Internet] [Thesis]. University of Ontario Institute of Technology; 2010. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/10155/87.

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

Council of Science Editors:

Patel A. Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices. [Thesis]. University of Ontario Institute of Technology; 2010. Available from: http://hdl.handle.net/10155/87

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


University of South Carolina

7. Gehr, Dennis Franklin. The Effect of Coating Parameters On Advanced TRISO Fuels With Zirconium Carbide.

Degree: MS, Nuclear Engineering, 2009, University of South Carolina

  Recent studies of TRISO fuel behavior have shown a number of problems with the conventional SiC TRISO coating system at very high temperature, not… (more)

Subjects/Keywords: Engineering; Nuclear Engineering; Helium Cooled Reactor; HTR; Pebble Bed Reactor; TRISO

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APA (6th Edition):

Gehr, D. F. (2009). The Effect of Coating Parameters On Advanced TRISO Fuels With Zirconium Carbide. (Masters Thesis). University of South Carolina. Retrieved from https://scholarcommons.sc.edu/etd/33

Chicago Manual of Style (16th Edition):

Gehr, Dennis Franklin. “The Effect of Coating Parameters On Advanced TRISO Fuels With Zirconium Carbide.” 2009. Masters Thesis, University of South Carolina. Accessed April 19, 2019. https://scholarcommons.sc.edu/etd/33.

MLA Handbook (7th Edition):

Gehr, Dennis Franklin. “The Effect of Coating Parameters On Advanced TRISO Fuels With Zirconium Carbide.” 2009. Web. 19 Apr 2019.

Vancouver:

Gehr DF. The Effect of Coating Parameters On Advanced TRISO Fuels With Zirconium Carbide. [Internet] [Masters thesis]. University of South Carolina; 2009. [cited 2019 Apr 19]. Available from: https://scholarcommons.sc.edu/etd/33.

Council of Science Editors:

Gehr DF. The Effect of Coating Parameters On Advanced TRISO Fuels With Zirconium Carbide. [Masters Thesis]. University of South Carolina; 2009. Available from: https://scholarcommons.sc.edu/etd/33


Texas A&M University

8. Fick, Lambert Hendrik. Direct Numerical Simulation of Incompressible Flows in Domains of Close Packed Spheres.

Degree: 2017, Texas A&M University

 This study aimed to investigate and quantify turbulent flow effects for incompressible, isothermal fluid flows in computational domains consisting of regularly packed spheres using high-fidelity… (more)

Subjects/Keywords: Direct Numerical Simulation; Pebble Bed Nuclear Reactor

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APA (6th Edition):

Fick, L. H. (2017). Direct Numerical Simulation of Incompressible Flows in Domains of Close Packed Spheres. (Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/161650

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

Chicago Manual of Style (16th Edition):

Fick, Lambert Hendrik. “Direct Numerical Simulation of Incompressible Flows in Domains of Close Packed Spheres.” 2017. Thesis, Texas A&M University. Accessed April 19, 2019. http://hdl.handle.net/1969.1/161650.

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

MLA Handbook (7th Edition):

Fick, Lambert Hendrik. “Direct Numerical Simulation of Incompressible Flows in Domains of Close Packed Spheres.” 2017. Web. 19 Apr 2019.

Vancouver:

Fick LH. Direct Numerical Simulation of Incompressible Flows in Domains of Close Packed Spheres. [Internet] [Thesis]. Texas A&M University; 2017. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/1969.1/161650.

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

Council of Science Editors:

Fick LH. Direct Numerical Simulation of Incompressible Flows in Domains of Close Packed Spheres. [Thesis]. Texas A&M University; 2017. Available from: http://hdl.handle.net/1969.1/161650

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


Texas A&M University

9. Rauch, Eric B. Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities.

Degree: 2010, Texas A&M University

 Small graphite-moderated and gas-cooled reactors have been around since the beginning of the atomic age. Though their existence in the past has been associated with… (more)

Subjects/Keywords: graphite moderated reactor; nuclear safeguards; DPRK

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APA (6th Edition):

Rauch, E. B. (2010). Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities. (Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2009-05-739

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

Chicago Manual of Style (16th Edition):

Rauch, Eric B. “Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities.” 2010. Thesis, Texas A&M University. Accessed April 19, 2019. http://hdl.handle.net/1969.1/ETD-TAMU-2009-05-739.

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

MLA Handbook (7th Edition):

Rauch, Eric B. “Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities.” 2010. Web. 19 Apr 2019.

Vancouver:

Rauch EB. Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities. [Internet] [Thesis]. Texas A&M University; 2010. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2009-05-739.

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

Council of Science Editors:

Rauch EB. Development of a Safeguards Approach for a Small Graphite Moderated Reactor and Associated Fuel Cycle Facilities. [Thesis]. Texas A&M University; 2010. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2009-05-739

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


University of Pretoria

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

Degree: Physics, 2011, University of Pretoria

 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)

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

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APA (6th 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 (16th Edition):

Hlatshwayo, Thulani Thokozani. “Diffusion of silver in 6H-SiC.” 2011. Doctoral Dissertation, University of Pretoria. Accessed April 19, 2019. http://hdl.handle.net/2263/25616.

MLA Handbook (7th Edition):

Hlatshwayo, Thulani Thokozani. “Diffusion of silver in 6H-SiC.” 2011. Web. 19 Apr 2019.

Vancouver:

Hlatshwayo TT. Diffusion of silver in 6H-SiC. [Internet] [Doctoral dissertation]. University of Pretoria; 2011. [cited 2019 Apr 19]. 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


Georgia Tech

11. Huning, Alexander Jared. A novel core analysis method for prismatic high temperature gas reactors.

Degree: PhD, Mechanical Engineering, 2016, Georgia Tech

 A new transient thermal hydraulic method for simulating prismatic HTGRs during a loss-of-forced-circulation (LOFC) accident is presented. This expands upon the steady state thermal hydraulic… (more)

Subjects/Keywords: HTGR; Thermal hydraulics; Reactor analysis; Nuclear safety

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APA (6th Edition):

Huning, A. J. (2016). A novel core analysis method for prismatic high temperature gas reactors. (Doctoral Dissertation). Georgia Tech. Retrieved from http://hdl.handle.net/1853/58615

Chicago Manual of Style (16th Edition):

Huning, Alexander Jared. “A novel core analysis method for prismatic high temperature gas reactors.” 2016. Doctoral Dissertation, Georgia Tech. Accessed April 19, 2019. http://hdl.handle.net/1853/58615.

MLA Handbook (7th Edition):

Huning, Alexander Jared. “A novel core analysis method for prismatic high temperature gas reactors.” 2016. Web. 19 Apr 2019.

Vancouver:

Huning AJ. A novel core analysis method for prismatic high temperature gas reactors. [Internet] [Doctoral dissertation]. Georgia Tech; 2016. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/1853/58615.

Council of Science Editors:

Huning AJ. A novel core analysis method for prismatic high temperature gas reactors. [Doctoral Dissertation]. Georgia Tech; 2016. Available from: http://hdl.handle.net/1853/58615


University of Pretoria

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

Degree: 2011, University of Pretoria

 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)

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

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APA (6th 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 (16th Edition):

Hlatshwayo, Thulani Thokozani. “Diffusion of silver in 6H-SiC .” 2011. Doctoral Dissertation, University of Pretoria. Accessed April 19, 2019. http://upetd.up.ac.za/thesis/available/etd-06182011-165556/.

MLA Handbook (7th Edition):

Hlatshwayo, Thulani Thokozani. “Diffusion of silver in 6H-SiC .” 2011. Web. 19 Apr 2019.

Vancouver:

Hlatshwayo TT. Diffusion of silver in 6H-SiC . [Internet] [Doctoral dissertation]. University of Pretoria; 2011. [cited 2019 Apr 19]. 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

13. Youssefnia, Mohammed Hossein. Pressurized water reactor (PWR) accident analysis using the EMERALD code.

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

 A detail study of a Loss of Coolant Accident (LOCA) in a Pressurized Water Reactor was conducted in order to estimate the consequences of the… (more)

Subjects/Keywords: Nuclear reactor accidents

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APA (6th Edition):

Youssefnia, M. H. (1980). Pressurized water reactor (PWR) accident analysis using the EMERALD code. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/42890

Chicago Manual of Style (16th Edition):

Youssefnia, Mohammed Hossein. “Pressurized water reactor (PWR) accident analysis using the EMERALD code.” 1980. Masters Thesis, Oregon State University. Accessed April 19, 2019. http://hdl.handle.net/1957/42890.

MLA Handbook (7th Edition):

Youssefnia, Mohammed Hossein. “Pressurized water reactor (PWR) accident analysis using the EMERALD code.” 1980. Web. 19 Apr 2019.

Vancouver:

Youssefnia MH. Pressurized water reactor (PWR) accident analysis using the EMERALD code. [Internet] [Masters thesis]. Oregon State University; 1980. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/1957/42890.

Council of Science Editors:

Youssefnia MH. Pressurized water reactor (PWR) accident analysis using the EMERALD code. [Masters Thesis]. Oregon State University; 1980. Available from: http://hdl.handle.net/1957/42890


University of Arizona

14. Szeligowski, John Joseph, 1943-. Numerical methods for solving the reactor kinetics equations .

Degree: 1966, University of Arizona

Subjects/Keywords: Nuclear reactor kinetics.

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APA (6th Edition):

Szeligowski, John Joseph, 1. (1966). Numerical methods for solving the reactor kinetics equations . (Masters Thesis). University of Arizona. Retrieved from http://hdl.handle.net/10150/551841

Chicago Manual of Style (16th Edition):

Szeligowski, John Joseph, 1943-. “Numerical methods for solving the reactor kinetics equations .” 1966. Masters Thesis, University of Arizona. Accessed April 19, 2019. http://hdl.handle.net/10150/551841.

MLA Handbook (7th Edition):

Szeligowski, John Joseph, 1943-. “Numerical methods for solving the reactor kinetics equations .” 1966. Web. 19 Apr 2019.

Vancouver:

Szeligowski, John Joseph 1. Numerical methods for solving the reactor kinetics equations . [Internet] [Masters thesis]. University of Arizona; 1966. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/10150/551841.

Council of Science Editors:

Szeligowski, John Joseph 1. Numerical methods for solving the reactor kinetics equations . [Masters Thesis]. University of Arizona; 1966. Available from: http://hdl.handle.net/10150/551841

15. Pedro, Miguel António de Morais. Viabilidade económica da implementação de um reactor nuclear para a produção de energia eléctrica em Portugal.

Degree: 2012, Repositório Científico do Instituto Politécnico de Lisboa

 O presente trabalho tem como objectivo avaliar economicamente e determinar a viabilidade da implementação de um reactor nuclear para produção de energia eléctrica. Faz-se uma… (more)

Subjects/Keywords: Avaliação económica; Energia nuclear; Reactor PWR

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APA (6th Edition):

Pedro, M. A. d. M. (2012). Viabilidade económica da implementação de um reactor nuclear para a produção de energia eléctrica em Portugal. (Thesis). Repositório Científico do Instituto Politécnico de Lisboa. Retrieved from http://www.rcaap.pt/detail.jsp?id=oai:repositorio.ipl.pt:10400.21/2167

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

Chicago Manual of Style (16th Edition):

Pedro, Miguel António de Morais. “Viabilidade económica da implementação de um reactor nuclear para a produção de energia eléctrica em Portugal.” 2012. Thesis, Repositório Científico do Instituto Politécnico de Lisboa. Accessed April 19, 2019. http://www.rcaap.pt/detail.jsp?id=oai:repositorio.ipl.pt:10400.21/2167.

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

MLA Handbook (7th Edition):

Pedro, Miguel António de Morais. “Viabilidade económica da implementação de um reactor nuclear para a produção de energia eléctrica em Portugal.” 2012. Web. 19 Apr 2019.

Vancouver:

Pedro MAdM. Viabilidade económica da implementação de um reactor nuclear para a produção de energia eléctrica em Portugal. [Internet] [Thesis]. Repositório Científico do Instituto Politécnico de Lisboa; 2012. [cited 2019 Apr 19]. Available from: http://www.rcaap.pt/detail.jsp?id=oai:repositorio.ipl.pt:10400.21/2167.

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

Council of Science Editors:

Pedro MAdM. Viabilidade económica da implementação de um reactor nuclear para a produção de energia eléctrica em Portugal. [Thesis]. Repositório Científico do Instituto Politécnico de Lisboa; 2012. Available from: http://www.rcaap.pt/detail.jsp?id=oai:repositorio.ipl.pt:10400.21/2167

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

16. Petersson, Jens. CFD-analysis of buoyancy-driven flow inside a cooling pipe system attached to a reactor pressure vessel.

Degree: The Institute of Technology, 2014, Linköping UniversityLinköping University

  In this work a cooling system connected to a reactor pressure vessel has been studied using the CFD method for the purpose of investigating… (more)

Subjects/Keywords: CFD; OpenFOAM; LES; nuclear reactor pressure vessel

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APA (6th Edition):

Petersson, J. (2014). CFD-analysis of buoyancy-driven flow inside a cooling pipe system attached to a reactor pressure vessel. (Thesis). Linköping UniversityLinköping University. Retrieved from http://urn.kb.se/resolve?urn=urn:nbn:se:liu:diva-112796

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

Chicago Manual of Style (16th Edition):

Petersson, Jens. “CFD-analysis of buoyancy-driven flow inside a cooling pipe system attached to a reactor pressure vessel.” 2014. Thesis, Linköping UniversityLinköping University. Accessed April 19, 2019. http://urn.kb.se/resolve?urn=urn:nbn:se:liu:diva-112796.

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

MLA Handbook (7th Edition):

Petersson, Jens. “CFD-analysis of buoyancy-driven flow inside a cooling pipe system attached to a reactor pressure vessel.” 2014. Web. 19 Apr 2019.

Vancouver:

Petersson J. CFD-analysis of buoyancy-driven flow inside a cooling pipe system attached to a reactor pressure vessel. [Internet] [Thesis]. Linköping UniversityLinköping University; 2014. [cited 2019 Apr 19]. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:liu:diva-112796.

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

Council of Science Editors:

Petersson J. CFD-analysis of buoyancy-driven flow inside a cooling pipe system attached to a reactor pressure vessel. [Thesis]. Linköping UniversityLinköping University; 2014. Available from: http://urn.kb.se/resolve?urn=urn:nbn:se:liu:diva-112796

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


MIT

17. Connaway, Heather M. (Heather Moira). Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II) ; Development of a core design optimization tool and analysis in support of the planned low enriched uranium conversion of the MIT Research Reactor (MITR-II) .

Degree: MS, Department of Nuclear Science and Engineering, 2012, MIT

 The MIT Research Reactor (MITR-II) is currently undergoing analysis for the planned conversion from high enriched uranium (HEU) to low enriched uranium (LEU), as part… (more)

Subjects/Keywords: Nuclear Science and Engineering.; M.I.T. Research Reactor

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APA (6th Edition):

Connaway, H. M. (. M. (2012). Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II) ; Development of a core design optimization tool and analysis in support of the planned low enriched uranium conversion of the MIT Research Reactor (MITR-II) . (Masters Thesis). MIT. Retrieved from http://hdl.handle.net/1721.1/79032

Chicago Manual of Style (16th Edition):

Connaway, Heather M (Heather Moira). “Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II) ; Development of a core design optimization tool and analysis in support of the planned low enriched uranium conversion of the MIT Research Reactor (MITR-II) .” 2012. Masters Thesis, MIT. Accessed April 19, 2019. http://hdl.handle.net/1721.1/79032.

MLA Handbook (7th Edition):

Connaway, Heather M (Heather Moira). “Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II) ; Development of a core design optimization tool and analysis in support of the planned low enriched uranium conversion of the MIT Research Reactor (MITR-II) .” 2012. Web. 19 Apr 2019.

Vancouver:

Connaway HM(M. Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II) ; Development of a core design optimization tool and analysis in support of the planned low enriched uranium conversion of the MIT Research Reactor (MITR-II) . [Internet] [Masters thesis]. MIT; 2012. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/1721.1/79032.

Council of Science Editors:

Connaway HM(M. Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II) ; Development of a core design optimization tool and analysis in support of the planned low enriched uranium conversion of the MIT Research Reactor (MITR-II) . [Masters Thesis]. MIT; 2012. Available from: http://hdl.handle.net/1721.1/79032


University of Texas – Austin

18. Bagdatlioglu, Cem. Fluence based neutron balance approach using spatial flux calculations.

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

 This thesis describes the addition of spatially dependent power sharing to a methodology for calculating the input and output isotopics and burnup of nuclear reactors… (more)

Subjects/Keywords: Nuclear reactor; Simulation; Fluence; Spatial; Flux

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APA (6th Edition):

Bagdatlioglu, C. (2015). Fluence based neutron balance approach using spatial flux calculations. (Thesis). University of Texas – Austin. Retrieved from http://hdl.handle.net/2152/31995

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

Chicago Manual of Style (16th Edition):

Bagdatlioglu, Cem. “Fluence based neutron balance approach using spatial flux calculations.” 2015. Thesis, University of Texas – Austin. Accessed April 19, 2019. http://hdl.handle.net/2152/31995.

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

MLA Handbook (7th Edition):

Bagdatlioglu, Cem. “Fluence based neutron balance approach using spatial flux calculations.” 2015. Web. 19 Apr 2019.

Vancouver:

Bagdatlioglu C. Fluence based neutron balance approach using spatial flux calculations. [Internet] [Thesis]. University of Texas – Austin; 2015. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/2152/31995.

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

Council of Science Editors:

Bagdatlioglu C. Fluence based neutron balance approach using spatial flux calculations. [Thesis]. University of Texas – Austin; 2015. Available from: http://hdl.handle.net/2152/31995

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


Texas A&M University

19. Allred, Jarrod Ryan. Experimental Analysis of Plutonium Product and Raffinate Waste Streams from a PUREX Process on A Low Burn-Up, Fast Neutron Irradiated DUO2 Pellet.

Degree: MS, Nuclear Engineering, 2016, Texas A&M University

 Experimental investigations of separating actinides (uranium and plutonium) from fission products (FP) were conducted using a modified Plutonium Uranium Recovery by Extraction (PUREX) process. The… (more)

Subjects/Keywords: Nuclear; Radiochemistry; PUREX; Fast Breeder Reactor

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APA (6th Edition):

Allred, J. R. (2016). Experimental Analysis of Plutonium Product and Raffinate Waste Streams from a PUREX Process on A Low Burn-Up, Fast Neutron Irradiated DUO2 Pellet. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/157740

Chicago Manual of Style (16th Edition):

Allred, Jarrod Ryan. “Experimental Analysis of Plutonium Product and Raffinate Waste Streams from a PUREX Process on A Low Burn-Up, Fast Neutron Irradiated DUO2 Pellet.” 2016. Masters Thesis, Texas A&M University. Accessed April 19, 2019. http://hdl.handle.net/1969.1/157740.

MLA Handbook (7th Edition):

Allred, Jarrod Ryan. “Experimental Analysis of Plutonium Product and Raffinate Waste Streams from a PUREX Process on A Low Burn-Up, Fast Neutron Irradiated DUO2 Pellet.” 2016. Web. 19 Apr 2019.

Vancouver:

Allred JR. Experimental Analysis of Plutonium Product and Raffinate Waste Streams from a PUREX Process on A Low Burn-Up, Fast Neutron Irradiated DUO2 Pellet. [Internet] [Masters thesis]. Texas A&M University; 2016. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/1969.1/157740.

Council of Science Editors:

Allred JR. Experimental Analysis of Plutonium Product and Raffinate Waste Streams from a PUREX Process on A Low Burn-Up, Fast Neutron Irradiated DUO2 Pellet. [Masters Thesis]. Texas A&M University; 2016. Available from: http://hdl.handle.net/1969.1/157740


University of Texas – Austin

20. -3703-6555. Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling.

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

 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)

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

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APA (6th Edition):

-3703-6555. (2015). Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling. (Thesis). 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
Not specified: Masters Thesis or Doctoral Dissertation

Chicago Manual of Style (16th Edition):

-3703-6555. “Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling.” 2015. Thesis, University of Texas – Austin. Accessed April 19, 2019. http://hdl.handle.net/2152/45763.

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

MLA Handbook (7th Edition):

-3703-6555. “Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling.” 2015. Web. 19 Apr 2019.

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] [Thesis]. University of Texas – Austin; 2015. [cited 2019 Apr 19]. 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
Not specified: Masters Thesis or Doctoral Dissertation

Council of Science Editors:

-3703-6555. Novel methods for generalizing nuclear fuel cycle design, and fuel burnup modeling. [Thesis]. 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
Not specified: Masters Thesis or Doctoral Dissertation


University of California – Irvine

21. Khubrani, Fahad Ali. Nuclear Power Safety in Saudi Arabia.

Degree: Electrical and Computer Engineering, 2016, University of California – Irvine

 Saudi Arabia is the world’s largest oil producer, producing about 13 % of the total globe oil production. Meanwhile, its domestic consumption of oil has… (more)

Subjects/Keywords: Engineering; Nuclear engineering; Nuclear Lessons learned; Nuclear Reactor; Nuclear Safety; Safety Culture

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APA (6th Edition):

Khubrani, F. A. (2016). Nuclear Power Safety in Saudi Arabia. (Thesis). University of California – Irvine. Retrieved from http://www.escholarship.org/uc/item/41f529ch

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

Chicago Manual of Style (16th Edition):

Khubrani, Fahad Ali. “Nuclear Power Safety in Saudi Arabia.” 2016. Thesis, University of California – Irvine. Accessed April 19, 2019. http://www.escholarship.org/uc/item/41f529ch.

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

MLA Handbook (7th Edition):

Khubrani, Fahad Ali. “Nuclear Power Safety in Saudi Arabia.” 2016. Web. 19 Apr 2019.

Vancouver:

Khubrani FA. Nuclear Power Safety in Saudi Arabia. [Internet] [Thesis]. University of California – Irvine; 2016. [cited 2019 Apr 19]. Available from: http://www.escholarship.org/uc/item/41f529ch.

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

Council of Science Editors:

Khubrani FA. Nuclear Power Safety in Saudi Arabia. [Thesis]. University of California – Irvine; 2016. Available from: http://www.escholarship.org/uc/item/41f529ch

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

22. Talley, Kemper Dyar. Beta-Delayed Neutron Data and Models for SCALE.

Degree: 2016, University of Tennessee – Knoxville

 Recent advancements in experimental and theoretical nuclear physics have yielded new data and models that more accurately describe the decay of fission products compared to… (more)

Subjects/Keywords: Neutron Emission; Beta Decay; Nuclear Physics; Reactor Physics; Nuclear Data; SCALE; Nuclear; Nuclear Engineering

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APA (6th Edition):

Talley, K. D. (2016). Beta-Delayed Neutron Data and Models for SCALE. (Doctoral Dissertation). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_graddiss/4170

Chicago Manual of Style (16th Edition):

Talley, Kemper Dyar. “Beta-Delayed Neutron Data and Models for SCALE.” 2016. Doctoral Dissertation, University of Tennessee – Knoxville. Accessed April 19, 2019. https://trace.tennessee.edu/utk_graddiss/4170.

MLA Handbook (7th Edition):

Talley, Kemper Dyar. “Beta-Delayed Neutron Data and Models for SCALE.” 2016. Web. 19 Apr 2019.

Vancouver:

Talley KD. Beta-Delayed Neutron Data and Models for SCALE. [Internet] [Doctoral dissertation]. University of Tennessee – Knoxville; 2016. [cited 2019 Apr 19]. Available from: https://trace.tennessee.edu/utk_graddiss/4170.

Council of Science Editors:

Talley KD. Beta-Delayed Neutron Data and Models for SCALE. [Doctoral Dissertation]. University of Tennessee – Knoxville; 2016. Available from: https://trace.tennessee.edu/utk_graddiss/4170


Texas A&M University

23. Parham, Neil A. Development of Real-Time Fuel Management Capability at the Texas A&M Nuclear Science Center.

Degree: 2010, Texas A&M University

 For the Texas A&M University Nuclear Science Center reactor a fuel depletion code was created to develop real-time fuel management capability. This code package links… (more)

Subjects/Keywords: MCNP; ORIGEN2; Nuclear Science Center; NSCR; nuclear research reactor

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APA (6th Edition):

Parham, N. A. (2010). Development of Real-Time Fuel Management Capability at the Texas A&M Nuclear Science Center. (Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7744

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

Chicago Manual of Style (16th Edition):

Parham, Neil A. “Development of Real-Time Fuel Management Capability at the Texas A&M Nuclear Science Center.” 2010. Thesis, Texas A&M University. Accessed April 19, 2019. http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7744.

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

MLA Handbook (7th Edition):

Parham, Neil A. “Development of Real-Time Fuel Management Capability at the Texas A&M Nuclear Science Center.” 2010. Web. 19 Apr 2019.

Vancouver:

Parham NA. Development of Real-Time Fuel Management Capability at the Texas A&M Nuclear Science Center. [Internet] [Thesis]. Texas A&M University; 2010. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7744.

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

Council of Science Editors:

Parham NA. Development of Real-Time Fuel Management Capability at the Texas A&M Nuclear Science Center. [Thesis]. Texas A&M University; 2010. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7744

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


University of California – Berkeley

24. Zhang, Guanheng. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization.

Degree: Nuclear Engineering, 2015, University of California – Berkeley

 This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast Reactor (SFR) to generate a significant fraction of the core power from… (more)

Subjects/Keywords: Engineering; Nuclear engineering; Advanced Nuclear Reactor; Economics; Resource Utilization; Thorium

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APA (6th Edition):

Zhang, G. (2015). Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization. (Thesis). University of California – Berkeley. Retrieved from http://www.escholarship.org/uc/item/27z410zp

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

Chicago Manual of Style (16th Edition):

Zhang, Guanheng. “Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization.” 2015. Thesis, University of California – Berkeley. Accessed April 19, 2019. http://www.escholarship.org/uc/item/27z410zp.

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

MLA Handbook (7th Edition):

Zhang, Guanheng. “Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization.” 2015. Web. 19 Apr 2019.

Vancouver:

Zhang G. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization. [Internet] [Thesis]. University of California – Berkeley; 2015. [cited 2019 Apr 19]. Available from: http://www.escholarship.org/uc/item/27z410zp.

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

Council of Science Editors:

Zhang G. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization. [Thesis]. University of California – Berkeley; 2015. Available from: http://www.escholarship.org/uc/item/27z410zp

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


University of California – Berkeley

25. Galvez, Cristhian. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors.

Degree: Nuclear Engineering, 2011, University of California – Berkeley

 The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for… (more)

Subjects/Keywords: Nuclear Engineering; Mechanical Engineering; nuclear; passive; pebble-bed; reactor; safety; transient

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APA (6th Edition):

Galvez, C. (2011). Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors. (Thesis). University of California – Berkeley. Retrieved from http://www.escholarship.org/uc/item/0g2353c7

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

Chicago Manual of Style (16th Edition):

Galvez, Cristhian. “Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors.” 2011. Thesis, University of California – Berkeley. Accessed April 19, 2019. http://www.escholarship.org/uc/item/0g2353c7.

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

MLA Handbook (7th Edition):

Galvez, Cristhian. “Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors.” 2011. Web. 19 Apr 2019.

Vancouver:

Galvez C. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors. [Internet] [Thesis]. University of California – Berkeley; 2011. [cited 2019 Apr 19]. Available from: http://www.escholarship.org/uc/item/0g2353c7.

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

Council of Science Editors:

Galvez C. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors. [Thesis]. University of California – Berkeley; 2011. Available from: http://www.escholarship.org/uc/item/0g2353c7

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


Virginia Commonwealth University

26. Britton, Kyle A. Neutronics Studies on the NIST Reactor Using the GA LEU fuel.

Degree: MS, Mechanical and Nuclear Engineering, 2018, Virginia Commonwealth University

  The National Bureau of Standards Reactor (NBSR) located on the National Institute of Standards and Technology (NIST) Gaithersburg campus, is currently underway of fuel… (more)

Subjects/Keywords: Neutronics; NBSR; TRIGA fuel; LEU fuel; Research Reactor; Nuclear; Nuclear Engineering

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APA (6th Edition):

Britton, K. A. (2018). Neutronics Studies on the NIST Reactor Using the GA LEU fuel. (Thesis). Virginia Commonwealth University. Retrieved from https://scholarscompass.vcu.edu/etd/5702

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

Chicago Manual of Style (16th Edition):

Britton, Kyle A. “Neutronics Studies on the NIST Reactor Using the GA LEU fuel.” 2018. Thesis, Virginia Commonwealth University. Accessed April 19, 2019. https://scholarscompass.vcu.edu/etd/5702.

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

MLA Handbook (7th Edition):

Britton, Kyle A. “Neutronics Studies on the NIST Reactor Using the GA LEU fuel.” 2018. Web. 19 Apr 2019.

Vancouver:

Britton KA. Neutronics Studies on the NIST Reactor Using the GA LEU fuel. [Internet] [Thesis]. Virginia Commonwealth University; 2018. [cited 2019 Apr 19]. Available from: https://scholarscompass.vcu.edu/etd/5702.

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

Council of Science Editors:

Britton KA. Neutronics Studies on the NIST Reactor Using the GA LEU fuel. [Thesis]. Virginia Commonwealth University; 2018. Available from: https://scholarscompass.vcu.edu/etd/5702

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


University of California – Berkeley

27. Wang, Xin. Coupled neutronics and thermal-hydraulics modeling for pebble-bed Fluoride-Salt-Cooled, High-Temperature Reactor (FHR).

Degree: Nuclear Engineering, 2018, University of California – Berkeley

 Advances in computer abilities, intense competition on the energy market and stringent regulatoryrequirements during the last decade have spurred the development of robust numericalmodels to… (more)

Subjects/Keywords: Nuclear engineering; diffusion; FHR; multiphysics; nuclear reactor; pebble bed; porous media

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APA (6th Edition):

Wang, X. (2018). Coupled neutronics and thermal-hydraulics modeling for pebble-bed Fluoride-Salt-Cooled, High-Temperature Reactor (FHR). (Thesis). University of California – Berkeley. Retrieved from http://www.escholarship.org/uc/item/40q3985m

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

Chicago Manual of Style (16th Edition):

Wang, Xin. “Coupled neutronics and thermal-hydraulics modeling for pebble-bed Fluoride-Salt-Cooled, High-Temperature Reactor (FHR).” 2018. Thesis, University of California – Berkeley. Accessed April 19, 2019. http://www.escholarship.org/uc/item/40q3985m.

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

MLA Handbook (7th Edition):

Wang, Xin. “Coupled neutronics and thermal-hydraulics modeling for pebble-bed Fluoride-Salt-Cooled, High-Temperature Reactor (FHR).” 2018. Web. 19 Apr 2019.

Vancouver:

Wang X. Coupled neutronics and thermal-hydraulics modeling for pebble-bed Fluoride-Salt-Cooled, High-Temperature Reactor (FHR). [Internet] [Thesis]. University of California – Berkeley; 2018. [cited 2019 Apr 19]. Available from: http://www.escholarship.org/uc/item/40q3985m.

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

Council of Science Editors:

Wang X. Coupled neutronics and thermal-hydraulics modeling for pebble-bed Fluoride-Salt-Cooled, High-Temperature Reactor (FHR). [Thesis]. University of California – Berkeley; 2018. Available from: http://www.escholarship.org/uc/item/40q3985m

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


University of New Mexico

28. Ball, Thomas A. The Inverse Kinetics Method and Its Application to the Annular Core Research Reactor.

Degree: Nuclear Engineering, 2017, University of New Mexico

  The inverse kinetics method, is a method to calculate a reactor’s reactivity profile from its power profile. In this thesis, the reactivity profile corresponding… (more)

Subjects/Keywords: Reactivity; Nuclear Reactor; Inverse Method; Inverse Kinetics; ACRR; Nuclear Engineering

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APA (6th Edition):

Ball, T. A. (2017). The Inverse Kinetics Method and Its Application to the Annular Core Research Reactor. (Masters Thesis). University of New Mexico. Retrieved from https://digitalrepository.unm.edu/ne_etds/66

Chicago Manual of Style (16th Edition):

Ball, Thomas A. “The Inverse Kinetics Method and Its Application to the Annular Core Research Reactor.” 2017. Masters Thesis, University of New Mexico. Accessed April 19, 2019. https://digitalrepository.unm.edu/ne_etds/66.

MLA Handbook (7th Edition):

Ball, Thomas A. “The Inverse Kinetics Method and Its Application to the Annular Core Research Reactor.” 2017. Web. 19 Apr 2019.

Vancouver:

Ball TA. The Inverse Kinetics Method and Its Application to the Annular Core Research Reactor. [Internet] [Masters thesis]. University of New Mexico; 2017. [cited 2019 Apr 19]. Available from: https://digitalrepository.unm.edu/ne_etds/66.

Council of Science Editors:

Ball TA. The Inverse Kinetics Method and Its Application to the Annular Core Research Reactor. [Masters Thesis]. University of New Mexico; 2017. Available from: https://digitalrepository.unm.edu/ne_etds/66


University of Arizona

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

Degree: 1989, University of Arizona

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

Subjects/Keywords: Nuclear fuels.; Nuclear reactor accidents

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

APA (6th Edition):

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

Chicago Manual of Style (16th Edition):

Smith, Adrienne Bobbette, 1960-. “Nuclear excursions in aqueous solutions of fissile materials .” 1989. Masters Thesis, University of Arizona. Accessed April 19, 2019. http://hdl.handle.net/10150/277150.

MLA Handbook (7th Edition):

Smith, Adrienne Bobbette, 1960-. “Nuclear excursions in aqueous solutions of fissile materials .” 1989. Web. 19 Apr 2019.

Vancouver:

Smith, Adrienne Bobbette 1. Nuclear excursions in aqueous solutions of fissile materials . [Internet] [Masters thesis]. University of Arizona; 1989. [cited 2019 Apr 19]. Available from: http://hdl.handle.net/10150/277150.

Council of Science Editors:

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


University of Tennessee – Knoxville

30. Kirkland, William Matthews. Improvements to NESTLE: Cross Section Interpolation and <i>N</i>-Group Extension.

Degree: MS, Nuclear Engineering, 2017, University of Tennessee – Knoxville

  The NESTLE program is a few-group neutron diffusion reactor core simulator code utilizing the nodal expansion method (NEM). This thesis presents two improvements made… (more)

Subjects/Keywords: nestle; neutron diffusion; interpolation; nuclear cross section; nuclear reactor; computational; Nuclear Engineering

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

APA (6th Edition):

Kirkland, W. M. (2017). Improvements to NESTLE: Cross Section Interpolation and <i>N</i>-Group Extension. (Thesis). University of Tennessee – Knoxville. Retrieved from https://trace.tennessee.edu/utk_gradthes/4951

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

Chicago Manual of Style (16th Edition):

Kirkland, William Matthews. “Improvements to NESTLE: Cross Section Interpolation and <i>N</i>-Group Extension.” 2017. Thesis, University of Tennessee – Knoxville. Accessed April 19, 2019. https://trace.tennessee.edu/utk_gradthes/4951.

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

MLA Handbook (7th Edition):

Kirkland, William Matthews. “Improvements to NESTLE: Cross Section Interpolation and <i>N</i>-Group Extension.” 2017. Web. 19 Apr 2019.

Vancouver:

Kirkland WM. Improvements to NESTLE: Cross Section Interpolation and <i>N</i>-Group Extension. [Internet] [Thesis]. University of Tennessee – Knoxville; 2017. [cited 2019 Apr 19]. Available from: https://trace.tennessee.edu/utk_gradthes/4951.

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

Council of Science Editors:

Kirkland WM. Improvements to NESTLE: Cross Section Interpolation and <i>N</i>-Group Extension. [Thesis]. University of Tennessee – Knoxville; 2017. Available from: https://trace.tennessee.edu/utk_gradthes/4951

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

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