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You searched for subject:(VHTR). Showing records 1 – 30 of 40 total matches.

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Texas A&M University

1. Alwafi, Anas Mohammed. Investigation of the Flow of the Upper Plenum of a Scaled Very High Temperature Reactor during a Depressurized Cooldown Conduction Accident.

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

 Very High Temperature Reactors (VHTRs) are the future of nuclear reactors. A 1/16th scaled upper plenum of a VHTR was designed and assembled at Texas… (more)

Subjects/Keywords: PTV; VHTR; DCC; LOCA

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APA (6th Edition):

Alwafi, A. M. (2015). Investigation of the Flow of the Upper Plenum of a Scaled Very High Temperature Reactor during a Depressurized Cooldown Conduction Accident. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/187463

Chicago Manual of Style (16th Edition):

Alwafi, Anas Mohammed. “Investigation of the Flow of the Upper Plenum of a Scaled Very High Temperature Reactor during a Depressurized Cooldown Conduction Accident.” 2015. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/187463.

MLA Handbook (7th Edition):

Alwafi, Anas Mohammed. “Investigation of the Flow of the Upper Plenum of a Scaled Very High Temperature Reactor during a Depressurized Cooldown Conduction Accident.” 2015. Web. 24 Jan 2021.

Vancouver:

Alwafi AM. Investigation of the Flow of the Upper Plenum of a Scaled Very High Temperature Reactor during a Depressurized Cooldown Conduction Accident. [Internet] [Masters thesis]. Texas A&M University; 2015. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/187463.

Council of Science Editors:

Alwafi AM. Investigation of the Flow of the Upper Plenum of a Scaled Very High Temperature Reactor during a Depressurized Cooldown Conduction Accident. [Masters Thesis]. Texas A&M University; 2015. Available from: http://hdl.handle.net/1969.1/187463


Texas A&M University

2. Kanjanakijkasem, Worasit 1975-. Preliminary Study of Bypass Flow in Prismatic Core of Very High Temperature Reactor Using Small-Scale Model.

Degree: PhD, Mechanical Engineering, 2012, Texas A&M University

 Very high temperature reactor (VHTR) is one of the candidates for Generation IV reactor. It can be continuously operated with average core outlet temperature between… (more)

Subjects/Keywords: Prismatic core; VHTR; Bypass flow

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APA (6th Edition):

Kanjanakijkasem, W. 1. (2012). Preliminary Study of Bypass Flow in Prismatic Core of Very High Temperature Reactor Using Small-Scale Model. (Doctoral Dissertation). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/148222

Chicago Manual of Style (16th Edition):

Kanjanakijkasem, Worasit 1975-. “Preliminary Study of Bypass Flow in Prismatic Core of Very High Temperature Reactor Using Small-Scale Model.” 2012. Doctoral Dissertation, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/148222.

MLA Handbook (7th Edition):

Kanjanakijkasem, Worasit 1975-. “Preliminary Study of Bypass Flow in Prismatic Core of Very High Temperature Reactor Using Small-Scale Model.” 2012. Web. 24 Jan 2021.

Vancouver:

Kanjanakijkasem W1. Preliminary Study of Bypass Flow in Prismatic Core of Very High Temperature Reactor Using Small-Scale Model. [Internet] [Doctoral dissertation]. Texas A&M University; 2012. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/148222.

Council of Science Editors:

Kanjanakijkasem W1. Preliminary Study of Bypass Flow in Prismatic Core of Very High Temperature Reactor Using Small-Scale Model. [Doctoral Dissertation]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/148222


Texas A&M University

3. Wang, Huhu 1985-. CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor.

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

 Very High Temperature Rector (VHTR) had been designated as one of those promising reactors for the Next Generation (IV) Nuclear Plant (NGNP). For a prismatic… (more)

Subjects/Keywords: Crossflow; Bypass flow; Prismatic VHTR

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APA (6th Edition):

Wang, H. 1. (2012). CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/148310

Chicago Manual of Style (16th Edition):

Wang, Huhu 1985-. “CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor.” 2012. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/148310.

MLA Handbook (7th Edition):

Wang, Huhu 1985-. “CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor.” 2012. Web. 24 Jan 2021.

Vancouver:

Wang H1. CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/148310.

Council of Science Editors:

Wang H1. CFD Analysis of Core Bypass Flow and Crossflow in the Prismatic Very High Temperature Gas-cooled Nuclear Reactor. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/148310


Texas A&M University

4. Corson, James. Development of MELCOR Input Techniques for High Temperature Gas-Cooled Reactor Analysis.

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

 High Temperature Gas-cooled Reactors (HTGRs) can provide clean electricity,as well as process heat that can be used to produce hydrogen for transportation and other sectors.… (more)

Subjects/Keywords: nuclear; htgr; vhtr; pbmr; httf; melcor

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APA (6th Edition):

Corson, J. (2011). Development of MELCOR Input Techniques for High Temperature Gas-Cooled Reactor Analysis. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7940

Chicago Manual of Style (16th Edition):

Corson, James. “Development of MELCOR Input Techniques for High Temperature Gas-Cooled Reactor Analysis.” 2011. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7940.

MLA Handbook (7th Edition):

Corson, James. “Development of MELCOR Input Techniques for High Temperature Gas-Cooled Reactor Analysis.” 2011. Web. 24 Jan 2021.

Vancouver:

Corson J. Development of MELCOR Input Techniques for High Temperature Gas-Cooled Reactor Analysis. [Internet] [Masters thesis]. Texas A&M University; 2011. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7940.

Council of Science Editors:

Corson J. Development of MELCOR Input Techniques for High Temperature Gas-Cooled Reactor Analysis. [Masters Thesis]. Texas A&M University; 2011. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7940


Texas A&M University

5. Frisani, Angelo. Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools.

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

 The design of passive heat removal systems is one of the main concerns for the modular Very High Temperature Gas-Cooled Reactors (VHTR) vessel cavity. The… (more)

Subjects/Keywords: Turbulence; buoyancy; convection; RCCS; CFD; VHTR

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APA (6th Edition):

Frisani, A. (2011). Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7767

Chicago Manual of Style (16th Edition):

Frisani, Angelo. “Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools.” 2011. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7767.

MLA Handbook (7th Edition):

Frisani, Angelo. “Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools.” 2011. Web. 24 Jan 2021.

Vancouver:

Frisani A. Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools. [Internet] [Masters thesis]. Texas A&M University; 2011. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7767.

Council of Science Editors:

Frisani A. Analysis of the Reactor Cavity Cooling System for Very High Temperature Gas-cooled Reactors Using Computational Fluid Dynamics Tools. [Masters Thesis]. Texas A&M University; 2011. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7767


North-West University

6. Sambureni, Privilege. Thermal fluid network model for a prismatic block in a gas-cooled reactor using FLOWNEX / Privilege Sambureni .

Degree: 2015, North-West University

 Very High Temperature Reactors are complex reactors and various system codes have been developed to design different aspects such as neutronics, thermal hydraulics etc. Flownex… (more)

Subjects/Keywords: VHTR; HTTR; Flownex; CFD; STAR-CCM+

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APA (6th Edition):

Sambureni, P. (2015). Thermal fluid network model for a prismatic block in a gas-cooled reactor using FLOWNEX / Privilege Sambureni . (Thesis). North-West University. Retrieved from http://hdl.handle.net/10394/15535

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):

Sambureni, Privilege. “Thermal fluid network model for a prismatic block in a gas-cooled reactor using FLOWNEX / Privilege Sambureni .” 2015. Thesis, North-West University. Accessed January 24, 2021. http://hdl.handle.net/10394/15535.

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

MLA Handbook (7th Edition):

Sambureni, Privilege. “Thermal fluid network model for a prismatic block in a gas-cooled reactor using FLOWNEX / Privilege Sambureni .” 2015. Web. 24 Jan 2021.

Vancouver:

Sambureni P. Thermal fluid network model for a prismatic block in a gas-cooled reactor using FLOWNEX / Privilege Sambureni . [Internet] [Thesis]. North-West University; 2015. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/10394/15535.

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

Council of Science Editors:

Sambureni P. Thermal fluid network model for a prismatic block in a gas-cooled reactor using FLOWNEX / Privilege Sambureni . [Thesis]. North-West University; 2015. Available from: http://hdl.handle.net/10394/15535

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


North-West University

7. Sehoana, Kabelo Albert. Simulation of natural circulation in an air-cooled Reactor Cavity Cooling System using Flownex / Kabelo Albert Sehoana .

Degree: 2014, North-West University

 Nuclear reactors with improved safety concepts are currently being studied within the nuclear engineering community, with a focus on passive safety features. One of these… (more)

Subjects/Keywords: VHTR; RCCS; Passive safety; Simulation; Flownex®

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APA (6th Edition):

Sehoana, K. A. (2014). Simulation of natural circulation in an air-cooled Reactor Cavity Cooling System using Flownex / Kabelo Albert Sehoana . (Thesis). North-West University. Retrieved from http://hdl.handle.net/10394/15542

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):

Sehoana, Kabelo Albert. “Simulation of natural circulation in an air-cooled Reactor Cavity Cooling System using Flownex / Kabelo Albert Sehoana .” 2014. Thesis, North-West University. Accessed January 24, 2021. http://hdl.handle.net/10394/15542.

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

MLA Handbook (7th Edition):

Sehoana, Kabelo Albert. “Simulation of natural circulation in an air-cooled Reactor Cavity Cooling System using Flownex / Kabelo Albert Sehoana .” 2014. Web. 24 Jan 2021.

Vancouver:

Sehoana KA. Simulation of natural circulation in an air-cooled Reactor Cavity Cooling System using Flownex / Kabelo Albert Sehoana . [Internet] [Thesis]. North-West University; 2014. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/10394/15542.

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

Council of Science Editors:

Sehoana KA. Simulation of natural circulation in an air-cooled Reactor Cavity Cooling System using Flownex / Kabelo Albert Sehoana . [Thesis]. North-West University; 2014. Available from: http://hdl.handle.net/10394/15542

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


The Ohio State University

8. Arcilesi, David J., Jr. Experimental Verification of the Initial Stages of an HTGR Double-ended Guillotine Break.

Degree: PhD, Nuclear Engineering, 2018, The Ohio State University

 A critical event in the safety analysis of a High Temperature Gas-cooled Reactor (HTGR) is a depressurized loss-of-forced circulation (D-LOFC) accident followed by air ingress.… (more)

Subjects/Keywords: Nuclear Engineering; HTGR, VHTR, air-ingress accident

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APA (6th Edition):

Arcilesi, David J., J. (2018). Experimental Verification of the Initial Stages of an HTGR Double-ended Guillotine Break. (Doctoral Dissertation). The Ohio State University. Retrieved from http://rave.ohiolink.edu/etdc/view?acc_num=osu1533356728728114

Chicago Manual of Style (16th Edition):

Arcilesi, David J., Jr. “Experimental Verification of the Initial Stages of an HTGR Double-ended Guillotine Break.” 2018. Doctoral Dissertation, The Ohio State University. Accessed January 24, 2021. http://rave.ohiolink.edu/etdc/view?acc_num=osu1533356728728114.

MLA Handbook (7th Edition):

Arcilesi, David J., Jr. “Experimental Verification of the Initial Stages of an HTGR Double-ended Guillotine Break.” 2018. Web. 24 Jan 2021.

Vancouver:

Arcilesi, David J. J. Experimental Verification of the Initial Stages of an HTGR Double-ended Guillotine Break. [Internet] [Doctoral dissertation]. The Ohio State University; 2018. [cited 2021 Jan 24]. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1533356728728114.

Council of Science Editors:

Arcilesi, David J. J. Experimental Verification of the Initial Stages of an HTGR Double-ended Guillotine Break. [Doctoral Dissertation]. The Ohio State University; 2018. Available from: http://rave.ohiolink.edu/etdc/view?acc_num=osu1533356728728114

9. GARCÍA, Jesús Alberto Rosales. Modelagem detalhada de sistemas nucleares avançados do tipo leito de bolas com combustível encapsulado .

Degree: 2015, Universidade Federal de Pernambuco

 A sustentabilidade da energia nuclear dependerá, entre outros fatores, da capacidade de redução dos inventários dos resíduos nucleares de vida longa. Com esse objetivo, desenvolveu-se… (more)

Subjects/Keywords: Engenharia de Energia Nuclear; VHTR; ADS; transmutação; HTR-10

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APA (6th Edition):

GARCÍA, J. A. R. (2015). Modelagem detalhada de sistemas nucleares avançados do tipo leito de bolas com combustível encapsulado . (Thesis). Universidade Federal de Pernambuco. Retrieved from http://repositorio.ufpe.br/handle/123456789/15170

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):

GARCÍA, Jesús Alberto Rosales. “Modelagem detalhada de sistemas nucleares avançados do tipo leito de bolas com combustível encapsulado .” 2015. Thesis, Universidade Federal de Pernambuco. Accessed January 24, 2021. http://repositorio.ufpe.br/handle/123456789/15170.

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

MLA Handbook (7th Edition):

GARCÍA, Jesús Alberto Rosales. “Modelagem detalhada de sistemas nucleares avançados do tipo leito de bolas com combustível encapsulado .” 2015. Web. 24 Jan 2021.

Vancouver:

GARCÍA JAR. Modelagem detalhada de sistemas nucleares avançados do tipo leito de bolas com combustível encapsulado . [Internet] [Thesis]. Universidade Federal de Pernambuco; 2015. [cited 2021 Jan 24]. Available from: http://repositorio.ufpe.br/handle/123456789/15170.

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

Council of Science Editors:

GARCÍA JAR. Modelagem detalhada de sistemas nucleares avançados do tipo leito de bolas com combustível encapsulado . [Thesis]. Universidade Federal de Pernambuco; 2015. Available from: http://repositorio.ufpe.br/handle/123456789/15170

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


Texas A&M University

10. Ames, David E. High-Fidelity Nuclear Energy System Optimization towards an Environmentally Benign, Sustainable, and Secure Energy Source.

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

 A new high-fidelity integrated system method and analysis approach was developed and implemented for consistent and comprehensive evaluations of advanced fuel cycles leading to minimized… (more)

Subjects/Keywords: Nuclear; Optimization; VHTR; AP100; Recycle; Environment; Sensitivity; TRU

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APA (6th Edition):

Ames, D. E. (2011). High-Fidelity Nuclear Energy System Optimization towards an Environmentally Benign, Sustainable, and Secure Energy Source. (Doctoral Dissertation). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2010-08-8549

Chicago Manual of Style (16th Edition):

Ames, David E. “High-Fidelity Nuclear Energy System Optimization towards an Environmentally Benign, Sustainable, and Secure Energy Source.” 2011. Doctoral Dissertation, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2010-08-8549.

MLA Handbook (7th Edition):

Ames, David E. “High-Fidelity Nuclear Energy System Optimization towards an Environmentally Benign, Sustainable, and Secure Energy Source.” 2011. Web. 24 Jan 2021.

Vancouver:

Ames DE. High-Fidelity Nuclear Energy System Optimization towards an Environmentally Benign, Sustainable, and Secure Energy Source. [Internet] [Doctoral dissertation]. Texas A&M University; 2011. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-08-8549.

Council of Science Editors:

Ames DE. High-Fidelity Nuclear Energy System Optimization towards an Environmentally Benign, Sustainable, and Secure Energy Source. [Doctoral Dissertation]. Texas A&M University; 2011. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2010-08-8549


Texas A&M University

11. Marcantel, Grace Ann. Analysis of Neutron Environments in Advanced Reactors Vs. Triga Reactor for In-Core Behavior Studies.

Degree: MS, Nuclear Engineering, 2018, Texas A&M University

 The Fluoride High-Temperature Reactor (FHR) and the Very High-Temperature Reactor (VHTR) are advanced reactor designs in the generation IV class. Advanced reactors operate in more… (more)

Subjects/Keywords: TRIGA; FHR; VHTR; energy spectrum; MCNP; similarity factors

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APA (6th Edition):

Marcantel, G. A. (2018). Analysis of Neutron Environments in Advanced Reactors Vs. Triga Reactor for In-Core Behavior Studies. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/173485

Chicago Manual of Style (16th Edition):

Marcantel, Grace Ann. “Analysis of Neutron Environments in Advanced Reactors Vs. Triga Reactor for In-Core Behavior Studies.” 2018. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/173485.

MLA Handbook (7th Edition):

Marcantel, Grace Ann. “Analysis of Neutron Environments in Advanced Reactors Vs. Triga Reactor for In-Core Behavior Studies.” 2018. Web. 24 Jan 2021.

Vancouver:

Marcantel GA. Analysis of Neutron Environments in Advanced Reactors Vs. Triga Reactor for In-Core Behavior Studies. [Internet] [Masters thesis]. Texas A&M University; 2018. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/173485.

Council of Science Editors:

Marcantel GA. Analysis of Neutron Environments in Advanced Reactors Vs. Triga Reactor for In-Core Behavior Studies. [Masters Thesis]. Texas A&M University; 2018. Available from: http://hdl.handle.net/1969.1/173485


Texas A&M University

12. Gorman, Michael Joseph. Characterization of the Thermal Hydraulic Behavior of an Experimental Reactor Cavity Cooling System with Water.

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

 An existing experimental Reactor Cavity Cooling System using water as the coolant received extensive instrumentation and control upgrades to allow for a thorough investigation into… (more)

Subjects/Keywords: RCCS; natural convection; VHTR; geysering; multiphase flow; UVP; ultrasonic velocity profiling

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APA (6th Edition):

Gorman, M. J. (2015). Characterization of the Thermal Hydraulic Behavior of an Experimental Reactor Cavity Cooling System with Water. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/155387

Chicago Manual of Style (16th Edition):

Gorman, Michael Joseph. “Characterization of the Thermal Hydraulic Behavior of an Experimental Reactor Cavity Cooling System with Water.” 2015. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/155387.

MLA Handbook (7th Edition):

Gorman, Michael Joseph. “Characterization of the Thermal Hydraulic Behavior of an Experimental Reactor Cavity Cooling System with Water.” 2015. Web. 24 Jan 2021.

Vancouver:

Gorman MJ. Characterization of the Thermal Hydraulic Behavior of an Experimental Reactor Cavity Cooling System with Water. [Internet] [Masters thesis]. Texas A&M University; 2015. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/155387.

Council of Science Editors:

Gorman MJ. Characterization of the Thermal Hydraulic Behavior of an Experimental Reactor Cavity Cooling System with Water. [Masters Thesis]. Texas A&M University; 2015. Available from: http://hdl.handle.net/1969.1/155387


Texas A&M University

13. Anderson, Nolan Alan. Coupling RELAP5-3D and Fluent to analyze a Very High Temperature Reactor (VHTR) outlet plenum.

Degree: MS, Nuclear Engineering, 2006, Texas A&M University

 The Very High Temperature Reactor (VHTR) system behavior should be predicted during normal operating conditions and during transient conditions. To predict the VHTR system behavior… (more)

Subjects/Keywords: VHTR; RELAP5-3D; Fluent

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APA (6th Edition):

Anderson, N. A. (2006). Coupling RELAP5-3D and Fluent to analyze a Very High Temperature Reactor (VHTR) outlet plenum. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/4160

Chicago Manual of Style (16th Edition):

Anderson, Nolan Alan. “Coupling RELAP5-3D and Fluent to analyze a Very High Temperature Reactor (VHTR) outlet plenum.” 2006. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/4160.

MLA Handbook (7th Edition):

Anderson, Nolan Alan. “Coupling RELAP5-3D and Fluent to analyze a Very High Temperature Reactor (VHTR) outlet plenum.” 2006. Web. 24 Jan 2021.

Vancouver:

Anderson NA. Coupling RELAP5-3D and Fluent to analyze a Very High Temperature Reactor (VHTR) outlet plenum. [Internet] [Masters thesis]. Texas A&M University; 2006. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/4160.

Council of Science Editors:

Anderson NA. Coupling RELAP5-3D and Fluent to analyze a Very High Temperature Reactor (VHTR) outlet plenum. [Masters Thesis]. Texas A&M University; 2006. Available from: http://hdl.handle.net/1969.1/4160

14. ROJAS MAZAIRA, Leorlen Yunier. Desenvolvimento de um modelo geométrico detalhado para a modelagem termoidráulica de sistemas nucleares, do tipo leito de bolas .

Degree: 2016, Universidade Federal de Pernambuco

 A tecnologia VHTR (do inglês Very High Temperature Reactor, Reator de Temperatura Muito Elevada) representa o próximo estágio na evolução dos reatores HTGR (do inglês… (more)

Subjects/Keywords: Energia Nuclear; VHTR; CFD; Termoidráulica nuclear; HTR-10

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APA (6th Edition):

ROJAS MAZAIRA, L. Y. (2016). Desenvolvimento de um modelo geométrico detalhado para a modelagem termoidráulica de sistemas nucleares, do tipo leito de bolas . (Doctoral Dissertation). Universidade Federal de Pernambuco. Retrieved from https://repositorio.ufpe.br/handle/123456789/23485

Chicago Manual of Style (16th Edition):

ROJAS MAZAIRA, Leorlen Yunier. “Desenvolvimento de um modelo geométrico detalhado para a modelagem termoidráulica de sistemas nucleares, do tipo leito de bolas .” 2016. Doctoral Dissertation, Universidade Federal de Pernambuco. Accessed January 24, 2021. https://repositorio.ufpe.br/handle/123456789/23485.

MLA Handbook (7th Edition):

ROJAS MAZAIRA, Leorlen Yunier. “Desenvolvimento de um modelo geométrico detalhado para a modelagem termoidráulica de sistemas nucleares, do tipo leito de bolas .” 2016. Web. 24 Jan 2021.

Vancouver:

ROJAS MAZAIRA LY. Desenvolvimento de um modelo geométrico detalhado para a modelagem termoidráulica de sistemas nucleares, do tipo leito de bolas . [Internet] [Doctoral dissertation]. Universidade Federal de Pernambuco; 2016. [cited 2021 Jan 24]. Available from: https://repositorio.ufpe.br/handle/123456789/23485.

Council of Science Editors:

ROJAS MAZAIRA LY. Desenvolvimento de um modelo geométrico detalhado para a modelagem termoidráulica de sistemas nucleares, do tipo leito de bolas . [Doctoral Dissertation]. Universidade Federal de Pernambuco; 2016. Available from: https://repositorio.ufpe.br/handle/123456789/23485

15. PAIVA, Pedro Paulo Dantas de Souza. Análise CFD do núcleo prismático do VHTR com distintos modelos de turbulência e alteração de parâmetros da geometria .

Degree: 2017, Universidade Federal de Pernambuco

 O VHTR é um reator nuclear térmico, moderado a grafite e refrigerado por hélio. Para seu desenvolvimento, há a necessidade de utilização de ferramentas computacionais… (more)

Subjects/Keywords: Fluidodinâmica computacional; VHTR; Análise paramétrica; Acoplamento 1D-3D; Turbulência

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APA (6th Edition):

PAIVA, P. P. D. d. S. (2017). Análise CFD do núcleo prismático do VHTR com distintos modelos de turbulência e alteração de parâmetros da geometria . (Masters Thesis). Universidade Federal de Pernambuco. Retrieved from https://repositorio.ufpe.br/handle/123456789/25465

Chicago Manual of Style (16th Edition):

PAIVA, Pedro Paulo Dantas de Souza. “Análise CFD do núcleo prismático do VHTR com distintos modelos de turbulência e alteração de parâmetros da geometria .” 2017. Masters Thesis, Universidade Federal de Pernambuco. Accessed January 24, 2021. https://repositorio.ufpe.br/handle/123456789/25465.

MLA Handbook (7th Edition):

PAIVA, Pedro Paulo Dantas de Souza. “Análise CFD do núcleo prismático do VHTR com distintos modelos de turbulência e alteração de parâmetros da geometria .” 2017. Web. 24 Jan 2021.

Vancouver:

PAIVA PPDdS. Análise CFD do núcleo prismático do VHTR com distintos modelos de turbulência e alteração de parâmetros da geometria . [Internet] [Masters thesis]. Universidade Federal de Pernambuco; 2017. [cited 2021 Jan 24]. Available from: https://repositorio.ufpe.br/handle/123456789/25465.

Council of Science Editors:

PAIVA PPDdS. Análise CFD do núcleo prismático do VHTR com distintos modelos de turbulência e alteração de parâmetros da geometria . [Masters Thesis]. Universidade Federal de Pernambuco; 2017. Available from: https://repositorio.ufpe.br/handle/123456789/25465

16. GÁMEZ RODRÍGUEZ, Abel. Uma metodologia termo-fluido-dinâmica computacional para avaliação de reatores que operam a altíssimas temperaturas com leitos de combustíveis esféricos .

Degree: 2019, Universidade Federal de Pernambuco

 O crescimento da população mundial, a dependência dos combustíveis fósseis, a crescente demanda de energia por parte dos países em desenvolvimento, e os problemas associados… (more)

Subjects/Keywords: Engenharia nuclear; VHTR; Termoidráulica nuclear; HTR-10; CFD; ANSYS CFX

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APA (6th Edition):

GÁMEZ RODRÍGUEZ, A. (2019). Uma metodologia termo-fluido-dinâmica computacional para avaliação de reatores que operam a altíssimas temperaturas com leitos de combustíveis esféricos . (Doctoral Dissertation). Universidade Federal de Pernambuco. Retrieved from https://repositorio.ufpe.br/handle/123456789/36154

Chicago Manual of Style (16th Edition):

GÁMEZ RODRÍGUEZ, Abel. “Uma metodologia termo-fluido-dinâmica computacional para avaliação de reatores que operam a altíssimas temperaturas com leitos de combustíveis esféricos .” 2019. Doctoral Dissertation, Universidade Federal de Pernambuco. Accessed January 24, 2021. https://repositorio.ufpe.br/handle/123456789/36154.

MLA Handbook (7th Edition):

GÁMEZ RODRÍGUEZ, Abel. “Uma metodologia termo-fluido-dinâmica computacional para avaliação de reatores que operam a altíssimas temperaturas com leitos de combustíveis esféricos .” 2019. Web. 24 Jan 2021.

Vancouver:

GÁMEZ RODRÍGUEZ A. Uma metodologia termo-fluido-dinâmica computacional para avaliação de reatores que operam a altíssimas temperaturas com leitos de combustíveis esféricos . [Internet] [Doctoral dissertation]. Universidade Federal de Pernambuco; 2019. [cited 2021 Jan 24]. Available from: https://repositorio.ufpe.br/handle/123456789/36154.

Council of Science Editors:

GÁMEZ RODRÍGUEZ A. Uma metodologia termo-fluido-dinâmica computacional para avaliação de reatores que operam a altíssimas temperaturas com leitos de combustíveis esféricos . [Doctoral Dissertation]. Universidade Federal de Pernambuco; 2019. Available from: https://repositorio.ufpe.br/handle/123456789/36154

17. Huning, Alexander. A steady state thermal hydraulic analysis method for prismatic gas reactors.

Degree: MS, Mechanical Engineering, 2014, Georgia Tech

 A new methodology for the accurate and efficient determination of steady state thermal hydraulic parameters for prismatic high temperature gas reactors is developed. Two conceptual… (more)

Subjects/Keywords: Thermal hydraulics; VHTR; MHTGR

…Temperature Reactor (VHTR) Background HTGRs are gas reactor systems with coolant outlet… …temperatures up to 850°C. The VHTR is distinct from HTGRs as its coolant outlet temperature ranges… …C, the terms VHTR and HTGR are often used interchangeably. Higher outlet temperatures… …becomes a large concern. The need for the VHTR is driven by goals set forth by the Generation IV… …International Forum (GIF) (U.S. DOE, 2002). These goals that the VHTR must meet… 

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APA (6th Edition):

Huning, A. (2014). A steady state thermal hydraulic analysis method for prismatic gas reactors. (Masters Thesis). Georgia Tech. Retrieved from http://hdl.handle.net/1853/52196

Chicago Manual of Style (16th Edition):

Huning, Alexander. “A steady state thermal hydraulic analysis method for prismatic gas reactors.” 2014. Masters Thesis, Georgia Tech. Accessed January 24, 2021. http://hdl.handle.net/1853/52196.

MLA Handbook (7th Edition):

Huning, Alexander. “A steady state thermal hydraulic analysis method for prismatic gas reactors.” 2014. Web. 24 Jan 2021.

Vancouver:

Huning A. A steady state thermal hydraulic analysis method for prismatic gas reactors. [Internet] [Masters thesis]. Georgia Tech; 2014. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1853/52196.

Council of Science Editors:

Huning A. A steady state thermal hydraulic analysis method for prismatic gas reactors. [Masters Thesis]. Georgia Tech; 2014. Available from: http://hdl.handle.net/1853/52196

18. Lewis, Tom Goslee. Analysis of tru-fueled vhtr prismatic core performance domains.

Degree: MS, Nuclear Engineering, 2009, Texas A&M University

 The current waste management strategy for spent nuclear fuel (SNF) mandated by the U.S. Congress is the disposal of high-level waste (HLW) in a geological… (more)

Subjects/Keywords: VHTR; TRU; Operation Domains

…19 20 23 24 III. VHTR PRISMATIC CORE MODEL… …25 III.A 3D Whole-Core Model of a Power-Size VHTR .................................. III.A… …48 IV. PERFORMANCE ANALYSIS OF TRU-FUELED VHTR SYSTEMS OPERATING IN A SINGLE BATCH MODE… …fuel-ring VHTR configurations ......................................... 12 2 TRISO-coated… …18 5 CSAS25 sequence for double heterogeneous VHTR model ...................... 22 6… 

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APA (6th Edition):

Lewis, T. G. (2009). Analysis of tru-fueled vhtr prismatic core performance domains. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2557

Chicago Manual of Style (16th Edition):

Lewis, Tom Goslee. “Analysis of tru-fueled vhtr prismatic core performance domains.” 2009. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2557.

MLA Handbook (7th Edition):

Lewis, Tom Goslee. “Analysis of tru-fueled vhtr prismatic core performance domains.” 2009. Web. 24 Jan 2021.

Vancouver:

Lewis TG. Analysis of tru-fueled vhtr prismatic core performance domains. [Internet] [Masters thesis]. Texas A&M University; 2009. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2557.

Council of Science Editors:

Lewis TG. Analysis of tru-fueled vhtr prismatic core performance domains. [Masters Thesis]. Texas A&M University; 2009. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2557

19. Pritchard, Megan Leigh. Neutronic analysis of pebble-bed cores with transuranics.

Degree: MS, Nuclear Engineering, 2009, Texas A&M University

 At the brink of nuclear waste repository crises, viable alternatives for the long term radiotoxic wastes are seriously being considered worldwide. Minor actinides serve as… (more)

Subjects/Keywords: pebble-bed; VHTR

…IV.B Prototype Pebble-Bed VHTR Configuration ........................... IV.C. Neutron… …Minor Actinides as a Fuel Component for Ultra-Long Life VHTR Configurations: Designs… …VHTR CONCEPT The Next Generation Nuclear Plant (NGNP) concept envisions an advanced… …collaborators is the Very High Temperature Reactor (VHTR) design. This concept has promise… …interest of the United States Department of Energy (U.S. DOE) in the VHTR concept stems… 

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APA (6th Edition):

Pritchard, M. L. (2009). Neutronic analysis of pebble-bed cores with transuranics. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2559

Chicago Manual of Style (16th Edition):

Pritchard, Megan Leigh. “Neutronic analysis of pebble-bed cores with transuranics.” 2009. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2559.

MLA Handbook (7th Edition):

Pritchard, Megan Leigh. “Neutronic analysis of pebble-bed cores with transuranics.” 2009. Web. 24 Jan 2021.

Vancouver:

Pritchard ML. Neutronic analysis of pebble-bed cores with transuranics. [Internet] [Masters thesis]. Texas A&M University; 2009. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2559.

Council of Science Editors:

Pritchard ML. Neutronic analysis of pebble-bed cores with transuranics. [Masters Thesis]. Texas A&M University; 2009. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2559


Brno University of Technology

20. Hovorka, Martin. Materiály pro reaktory IV. generace: Materials for IV. generation power plants.

Degree: 2018, Brno University of Technology

 The aim of this thesis is to describe material needs of chosen Generation IV nuclear reactors – VHTR, SCWR, and MSR with the emphasis on… (more)

Subjects/Keywords: Jaderný; reaktor; IV. generace; VHTR; materiál; SCWR; MSR; jaderná elektrárna; Monel; Inconel; Hastelloy; Nuclear; reactor; IV generation; VHTR; material; SCWR; MSR; power plant; Monel; Inconel. Hastelloy

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APA (6th Edition):

Hovorka, M. (2018). Materiály pro reaktory IV. generace: Materials for IV. generation power plants. (Thesis). Brno University of Technology. Retrieved from http://hdl.handle.net/11012/32439

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):

Hovorka, Martin. “Materiály pro reaktory IV. generace: Materials for IV. generation power plants.” 2018. Thesis, Brno University of Technology. Accessed January 24, 2021. http://hdl.handle.net/11012/32439.

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

MLA Handbook (7th Edition):

Hovorka, Martin. “Materiály pro reaktory IV. generace: Materials for IV. generation power plants.” 2018. Web. 24 Jan 2021.

Vancouver:

Hovorka M. Materiály pro reaktory IV. generace: Materials for IV. generation power plants. [Internet] [Thesis]. Brno University of Technology; 2018. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/11012/32439.

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

Council of Science Editors:

Hovorka M. Materiály pro reaktory IV. generace: Materials for IV. generation power plants. [Thesis]. Brno University of Technology; 2018. Available from: http://hdl.handle.net/11012/32439

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


Texas A&M University

21. Park, Jae Hyung. Natural Circulation in the Upper Plenum of a Scaled Model of a Very High Temperature Reactor in the Event of Loss-of-Coolant Accident.

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

 The very high temperature reactor (VHTR) is one of the most promising next generation reactors which will be commercialized in 2030. A loss-of-coolant accident (LOCA)… (more)

Subjects/Keywords: PIV; VHTR; LOCA; PCC; natural circulation; turbulent jets; plumes; self-similarity; Q-criterion

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APA (6th Edition):

Park, J. H. (2016). Natural Circulation in the Upper Plenum of a Scaled Model of a Very High Temperature Reactor in the Event of Loss-of-Coolant Accident. (Doctoral Dissertation). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/157998

Chicago Manual of Style (16th Edition):

Park, Jae Hyung. “Natural Circulation in the Upper Plenum of a Scaled Model of a Very High Temperature Reactor in the Event of Loss-of-Coolant Accident.” 2016. Doctoral Dissertation, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/157998.

MLA Handbook (7th Edition):

Park, Jae Hyung. “Natural Circulation in the Upper Plenum of a Scaled Model of a Very High Temperature Reactor in the Event of Loss-of-Coolant Accident.” 2016. Web. 24 Jan 2021.

Vancouver:

Park JH. Natural Circulation in the Upper Plenum of a Scaled Model of a Very High Temperature Reactor in the Event of Loss-of-Coolant Accident. [Internet] [Doctoral dissertation]. Texas A&M University; 2016. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/157998.

Council of Science Editors:

Park JH. Natural Circulation in the Upper Plenum of a Scaled Model of a Very High Temperature Reactor in the Event of Loss-of-Coolant Accident. [Doctoral Dissertation]. Texas A&M University; 2016. Available from: http://hdl.handle.net/1969.1/157998


Texas A&M University

22. Alhashimi, Tariq Yaqoob Sayed. Measurement of Temperature Profile in the Reactor Cavity Cooling System.

Degree: MS, Nuclear Engineering, 2014, Texas A&M University

 The Reactor Cavity Cooling System (RCCS) is an important passive cooling safety system used to cool the cavity of generation IV Very High Temperature Reactors… (more)

Subjects/Keywords: Reactor Cavity Cooling System (RCCS); VHTR; Reverse Flow; Preferential Flow; Temperature Profile Reconstruction

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APA (6th Edition):

Alhashimi, T. Y. S. (2014). Measurement of Temperature Profile in the Reactor Cavity Cooling System. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/154093

Chicago Manual of Style (16th Edition):

Alhashimi, Tariq Yaqoob Sayed. “Measurement of Temperature Profile in the Reactor Cavity Cooling System.” 2014. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/154093.

MLA Handbook (7th Edition):

Alhashimi, Tariq Yaqoob Sayed. “Measurement of Temperature Profile in the Reactor Cavity Cooling System.” 2014. Web. 24 Jan 2021.

Vancouver:

Alhashimi TYS. Measurement of Temperature Profile in the Reactor Cavity Cooling System. [Internet] [Masters thesis]. Texas A&M University; 2014. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/154093.

Council of Science Editors:

Alhashimi TYS. Measurement of Temperature Profile in the Reactor Cavity Cooling System. [Masters Thesis]. Texas A&M University; 2014. Available from: http://hdl.handle.net/1969.1/154093


University of New Mexico

23. Travis, Boyce. An Effective Methodology for Thermal-Hydraulics Analysis of a VHTR Core and Fuel Elements.

Degree: Nuclear Engineering, 2013, University of New Mexico

 The Very High Temperature Reactor (VHTR) is a Generation-IV design in the conceptual pre-licensing phase for potential construction by 2030-2050. It is graphite moderated, helium… (more)

Subjects/Keywords: thermal-hydraulics; bypass flow; VHTR; Nusselt number correlation; helium coolant; prismatic fuel element

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APA (6th Edition):

Travis, B. (2013). An Effective Methodology for Thermal-Hydraulics Analysis of a VHTR Core and Fuel Elements. (Masters Thesis). University of New Mexico. Retrieved from http://hdl.handle.net/1928/23215

Chicago Manual of Style (16th Edition):

Travis, Boyce. “An Effective Methodology for Thermal-Hydraulics Analysis of a VHTR Core and Fuel Elements.” 2013. Masters Thesis, University of New Mexico. Accessed January 24, 2021. http://hdl.handle.net/1928/23215.

MLA Handbook (7th Edition):

Travis, Boyce. “An Effective Methodology for Thermal-Hydraulics Analysis of a VHTR Core and Fuel Elements.” 2013. Web. 24 Jan 2021.

Vancouver:

Travis B. An Effective Methodology for Thermal-Hydraulics Analysis of a VHTR Core and Fuel Elements. [Internet] [Masters thesis]. University of New Mexico; 2013. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1928/23215.

Council of Science Editors:

Travis B. An Effective Methodology for Thermal-Hydraulics Analysis of a VHTR Core and Fuel Elements. [Masters Thesis]. University of New Mexico; 2013. Available from: http://hdl.handle.net/1928/23215


Georgia Tech

24. Zhang, Zhan. Neutron energy spectrum reconstruction method based for htr reactor calculations.

Degree: MS, Mechanical Engineering, 2011, Georgia Tech

 In the deep burn research of Very High Temperature Reactor (VHTR), it is desired to make an accurate estimation of absorption cross sections and absorption… (more)

Subjects/Keywords: Energy spectrum; Burnable poison; VHTR; Cross section; Fuel burnup (Nuclear engineering); Neutrons Spectra

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APA (6th Edition):

Zhang, Z. (2011). Neutron energy spectrum reconstruction method based for htr reactor calculations. (Masters Thesis). Georgia Tech. Retrieved from http://hdl.handle.net/1853/41195

Chicago Manual of Style (16th Edition):

Zhang, Zhan. “Neutron energy spectrum reconstruction method based for htr reactor calculations.” 2011. Masters Thesis, Georgia Tech. Accessed January 24, 2021. http://hdl.handle.net/1853/41195.

MLA Handbook (7th Edition):

Zhang, Zhan. “Neutron energy spectrum reconstruction method based for htr reactor calculations.” 2011. Web. 24 Jan 2021.

Vancouver:

Zhang Z. Neutron energy spectrum reconstruction method based for htr reactor calculations. [Internet] [Masters thesis]. Georgia Tech; 2011. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1853/41195.

Council of Science Editors:

Zhang Z. Neutron energy spectrum reconstruction method based for htr reactor calculations. [Masters Thesis]. Georgia Tech; 2011. Available from: http://hdl.handle.net/1853/41195

25. Nelson, Benjamin L. Scaling analysis for the pebble bed of the very high temperature gas-cooled reactor thermal hydraulic test facility.

Degree: MS, Nuclear Engineering, 2009, Oregon State University

 The Very High Temperature Reactor (VHTR) has two possible core configurations, a hexagonal prismatic and a pebble bed. It is essential that an experimental facility… (more)

Subjects/Keywords: VHTR; Pebble bed reactors  – Mathematical models

…Thermal Hydraulic Test Facility 1 INTRODUCTION The Very High Temperature Reactor (VHTR… …experimental data to validate safety and accident analysis codes. The VHTR was selected by the… …hydraulic scaling investigation in support of the NRC’s VHTR licensing program. The goals of this… …literature review of current and previous work with the VHTR and packed bed systems. The VHTR… …systems. The key phenomena of interest for the VHTR involve the pressure drop across the packed… 

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APA (6th Edition):

Nelson, B. L. (2009). Scaling analysis for the pebble bed of the very high temperature gas-cooled reactor thermal hydraulic test facility. (Masters Thesis). Oregon State University. Retrieved from http://hdl.handle.net/1957/11999

Chicago Manual of Style (16th Edition):

Nelson, Benjamin L. “Scaling analysis for the pebble bed of the very high temperature gas-cooled reactor thermal hydraulic test facility.” 2009. Masters Thesis, Oregon State University. Accessed January 24, 2021. http://hdl.handle.net/1957/11999.

MLA Handbook (7th Edition):

Nelson, Benjamin L. “Scaling analysis for the pebble bed of the very high temperature gas-cooled reactor thermal hydraulic test facility.” 2009. Web. 24 Jan 2021.

Vancouver:

Nelson BL. Scaling analysis for the pebble bed of the very high temperature gas-cooled reactor thermal hydraulic test facility. [Internet] [Masters thesis]. Oregon State University; 2009. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1957/11999.

Council of Science Editors:

Nelson BL. Scaling analysis for the pebble bed of the very high temperature gas-cooled reactor thermal hydraulic test facility. [Masters Thesis]. Oregon State University; 2009. Available from: http://hdl.handle.net/1957/11999


Texas A&M University

26. Cuvelier, Marie-Hermine. Advanced Fuel Cycle Scenarios with AP1000 PWRs and VHTRs and Fission Spectrum Uncertainties.

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

 Minimization of HLW inventories and U consumption are key elements guaranteeing nuclear energy expansion. The integration of complex nuclear systems into a viable cycle yet… (more)

Subjects/Keywords: TRU transmutation; TRU elimination; fission spectrum discrepancies; fission spectrum uncertainties impact on fuel cycle parameters; AP1000-VHTR fuel cycle; Thorium; ThO2

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APA (6th Edition):

Cuvelier, M. (2012). Advanced Fuel Cycle Scenarios with AP1000 PWRs and VHTRs and Fission Spectrum Uncertainties. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2012-05-10984

Chicago Manual of Style (16th Edition):

Cuvelier, Marie-Hermine. “Advanced Fuel Cycle Scenarios with AP1000 PWRs and VHTRs and Fission Spectrum Uncertainties.” 2012. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2012-05-10984.

MLA Handbook (7th Edition):

Cuvelier, Marie-Hermine. “Advanced Fuel Cycle Scenarios with AP1000 PWRs and VHTRs and Fission Spectrum Uncertainties.” 2012. Web. 24 Jan 2021.

Vancouver:

Cuvelier M. Advanced Fuel Cycle Scenarios with AP1000 PWRs and VHTRs and Fission Spectrum Uncertainties. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2012-05-10984.

Council of Science Editors:

Cuvelier M. Advanced Fuel Cycle Scenarios with AP1000 PWRs and VHTRs and Fission Spectrum Uncertainties. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2012-05-10984


Texas A&M University

27. Mcvay, Kyle. Experimental Design and Flow Visualization for the Upper Plenum of a Very High Temperature Gas Cooled for Computer Fluid Dynamics Validation.

Degree: MS, Mechanical Engineering, 2014, Texas A&M University

 The Very High Temperature Reactor (VHTR) is a Generation IV nuclear reactor that is currently under design. It modifies the current high temperature gas reactor… (more)

Subjects/Keywords: VHTR; PIV; Experimental Modeling; CFD; CFD Validation; Natural Ciculation; Jets; Upper Plenum; Pressurized Conduction Cooldown; PCC

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APA (6th Edition):

Mcvay, K. (2014). Experimental Design and Flow Visualization for the Upper Plenum of a Very High Temperature Gas Cooled for Computer Fluid Dynamics Validation. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/153485

Chicago Manual of Style (16th Edition):

Mcvay, Kyle. “Experimental Design and Flow Visualization for the Upper Plenum of a Very High Temperature Gas Cooled for Computer Fluid Dynamics Validation.” 2014. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/153485.

MLA Handbook (7th Edition):

Mcvay, Kyle. “Experimental Design and Flow Visualization for the Upper Plenum of a Very High Temperature Gas Cooled for Computer Fluid Dynamics Validation.” 2014. Web. 24 Jan 2021.

Vancouver:

Mcvay K. Experimental Design and Flow Visualization for the Upper Plenum of a Very High Temperature Gas Cooled for Computer Fluid Dynamics Validation. [Internet] [Masters thesis]. Texas A&M University; 2014. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/153485.

Council of Science Editors:

Mcvay K. Experimental Design and Flow Visualization for the Upper Plenum of a Very High Temperature Gas Cooled for Computer Fluid Dynamics Validation. [Masters Thesis]. Texas A&M University; 2014. Available from: http://hdl.handle.net/1969.1/153485


Texas A&M University

28. Ames, David E, II. Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels.

Degree: MS, Nuclear Engineering, 2006, Texas A&M University

 Minor actinides represent the long-term radiotoxicity of nuclear wastes. As one of their potential incineration options, partitioning and transmutation in fission reactors are seriously considered… (more)

Subjects/Keywords: VHTR; Minor Actinides; Dancoff Factor; Double Heterogeneity

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APA (6th Edition):

Ames, David E, I. (2006). Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/4382

Chicago Manual of Style (16th Edition):

Ames, David E, II. “Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels.” 2006. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/4382.

MLA Handbook (7th Edition):

Ames, David E, II. “Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels.” 2006. Web. 24 Jan 2021.

Vancouver:

Ames, David E I. Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels. [Internet] [Masters thesis]. Texas A&M University; 2006. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/4382.

Council of Science Editors:

Ames, David E I. Configuration adjustment potential of the Very High Temperature Reactor prismatic cores with advanced actinide fuels. [Masters Thesis]. Texas A&M University; 2006. Available from: http://hdl.handle.net/1969.1/4382

29. Alajo, Ayodeji Babatunde. Impact of PWR spent fuel variations on TRU-fueled VHTRS.

Degree: MS, Nuclear Engineering, 2009, Texas A&M University

 Several alternative strategies are being considered as spent nuclear fuel (SNF) management options. Transuranic nuclides (TRU) are responsible for the SNF long-term radiotoxicity beyond the… (more)

Subjects/Keywords: VHTR; Spent nuclear fuel; TRU-fuel

…Control rod block dimensions . 63 25 3-D whole-core VHTR model with… …spent fuel … 50 XVI Selected TRU vectors from PWR spent nuclear fuel for VHTR… …analysis… 51 XVII VHTR core specifications …. 54 XVIII Fuel assembly block… …for the TRU-fueled VHTR analysis … 73 XXVI Beginning-of-life multiplication factors… …Cooled Reactor (SCWR). Very High Temperature Reactor (VHTR). I.A.5… 

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APA (6th Edition):

Alajo, A. B. (2009). Impact of PWR spent fuel variations on TRU-fueled VHTRS. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-2556

Chicago Manual of Style (16th Edition):

Alajo, Ayodeji Babatunde. “Impact of PWR spent fuel variations on TRU-fueled VHTRS.” 2009. Masters Thesis, Texas A&M University. Accessed January 24, 2021. http://hdl.handle.net/1969.1/ETD-TAMU-2556.

MLA Handbook (7th Edition):

Alajo, Ayodeji Babatunde. “Impact of PWR spent fuel variations on TRU-fueled VHTRS.” 2009. Web. 24 Jan 2021.

Vancouver:

Alajo AB. Impact of PWR spent fuel variations on TRU-fueled VHTRS. [Internet] [Masters thesis]. Texas A&M University; 2009. [cited 2021 Jan 24]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2556.

Council of Science Editors:

Alajo AB. Impact of PWR spent fuel variations on TRU-fueled VHTRS. [Masters Thesis]. Texas A&M University; 2009. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-2556


Penn State University

30. Sarangi, Suchismita. Scaled Experiment on Gravity Driven Exchange Flow for the Very High Temperature Reactor.

Degree: 2010, Penn State University

 The process of lock-exchange and gravity driven exchange flow for fluids of differing densities is of particular interest in the postulated Depressurized Loss of Forced… (more)

Subjects/Keywords: gravity driven exchange; vhtr; air ingress; buoyancy driven exchange; water brine

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APA (6th Edition):

Sarangi, S. (2010). Scaled Experiment on Gravity Driven Exchange Flow for the Very High Temperature Reactor. (Thesis). Penn State University. Retrieved from https://submit-etda.libraries.psu.edu/catalog/10843

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):

Sarangi, Suchismita. “Scaled Experiment on Gravity Driven Exchange Flow for the Very High Temperature Reactor.” 2010. Thesis, Penn State University. Accessed January 24, 2021. https://submit-etda.libraries.psu.edu/catalog/10843.

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

MLA Handbook (7th Edition):

Sarangi, Suchismita. “Scaled Experiment on Gravity Driven Exchange Flow for the Very High Temperature Reactor.” 2010. Web. 24 Jan 2021.

Vancouver:

Sarangi S. Scaled Experiment on Gravity Driven Exchange Flow for the Very High Temperature Reactor. [Internet] [Thesis]. Penn State University; 2010. [cited 2021 Jan 24]. Available from: https://submit-etda.libraries.psu.edu/catalog/10843.

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

Council of Science Editors:

Sarangi S. Scaled Experiment on Gravity Driven Exchange Flow for the Very High Temperature Reactor. [Thesis]. Penn State University; 2010. Available from: https://submit-etda.libraries.psu.edu/catalog/10843

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

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