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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
URL: http://hdl.handle.net/1969.1/187463 
► 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)
▼ 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 A&M University (TAMU) in order to study the behavior of flow in the upper plenum of a
VHTR. Flow was investigated under one major accident scenario, the Depressurized Conduction Cooldown (DCC); this occurs due to loss of force when operation is interrupted by loss of power. In this case, the fluid will have a natural convection, forcing it to flow to the upper plenum. Particle Tracking Velocimetry (PTV) was used to illustrate flow, using water as the working fluid. A PTV code was used to track the particles, and this was then averaged over all vectors after filtering out those that failed. All flow velocity compounds, such as the velocity magnitude, y-velocity, x-velocity, standard deviation, and flow streamlines were visualized. A sensitivity analysis was performed in order to confirm that the number of frames used was sufficient to reach a steady state. In addition, repeatability analysis was applied to the output data. Turbulent intensity, Reynolds stress, and error occurring with these tests were all estimated. Finally, experimental data was validated using benchmark data.
Advisors/Committee Members: Hassan, Yassin A. (advisor), Marlow, William H. (committee member), King, Maria D. (committee member).
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
URL: http://hdl.handle.net/1969.1/148222
► 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)
▼ 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 900Β°C and 950Β°C, so the core temperature is one of the key features in the design of
VHTR. Bypass flow in the prismatic core of
VHTR is not a designed feature but it is inevitable due to the combination of several causes and considerably affects the core temperature. Although bypass flow has been studied extensively, the current status of research on thermal/hydraulic core flow of
VHTR is far from completion. Present study is the starting of bypass flow characteristic investigation using small-scale model that will fulfill understandings of bypass flow in the prismatic core of
VHTR.
Bypass flow experiments are conducted by using three small-scale models of prismatic blocks. They are stacked in a test section to form bypass gaps of single-layer blocks as exist in prismatic core of
VHTR. Three bypass gap widths set in air and water flow experiments are 6.1, 4.4 and 2.7 mm. Experimental data shows that bypass flow fraction depends on bypass gap width and downstream condition of prismatic blocks, while pressure drop of flow through bypass gaps depends on bypass gap width only.
Bypass flow simulations are performed by using STAR-CCM+ software after meshing parameters were determined from simulation exercises and grid independent study. Three turbulence models are employed in all bypass flow simulations which are stopped at physical time of 100 seconds marching by implicit unsteady scheme. Bypass flow fraction, coolant channel Reynolds number and bypass gap Reynolds number from air flow and water flow simulations with 6.1-mm bypass gap width are very close to experimental data. This is because bypass flow fractions from experiments at this bypass gap width are matched in construction of the simulation models. Discrepancies between results from simulations and experiments for remaining gaps increase when bypass gap width becomes smaller. Finally, guidelines for bypass flow experiments and simulations are drawn from the data in present study to improve bypass flow study in the future.
Advisors/Committee Members: Hassan, Yassin A (advisor), Lau, Sai C (committee member), Annamalai, Kalyan (committee member), Marlow, William H (committee member).
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
URL: http://hdl.handle.net/1969.1/148310
► 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)
▼ 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 core
VHTR, one of the most crucial design considerations is the bypass flow and crossflow effect. The bypass flow occurs when the coolant flow into gaps between fuel blocks. These gaps are formed as a result of carbon expansion and shrinkage induced by radiations and manufacturing and installation errors. Hot spots may appear in the core if the large portion of the coolant flows into bypass gaps instead of coolant channels in which the cooling efficiency is much higher.
A preliminary three dimensional steady-state CFD analysis was performed with commercial code STARCCM+ 6.04 to investigate the bypass flow and crossflow phenomenon in the prismatic
VHTR core. The k-Ξ΅ turbulence model was selected because of its robustness and low computational cost with respect to a decent accuracy for varied flow patterns. The wall treatment used in the present work is two-layer all y+ wall treatment to blend the wall laws to estimate the shear stress. Uniform mass flow rate was chose as the inlet condition and the outlet condition was zero gauge pressure outlet.
Grid independence study was performed and the results indicated that the discrepancy of the solution due to the mesh density was within 2% of the bypass flow fraction. The computational results showed that the bypass flow fraction was around 12%. Furthermore, the presence of the crossflow gap resulted in a up to 28% reduction of the coolant in the bypass flow gap while mass flow rate of coolant in coolant channels increased by around 5%. The pressure drop at the inlet due to the sudden contraction in area could be around 1kpa while the value was about 180 Pa around the crossflow gap region. The error analysis was also performed to evaluate the accumulated errors from the process of discretization and iteration. It was found that the total error was around 4% and the variation for the bypass flow fraction was within 1%.
Advisors/Committee Members: Hassan, Yassin A (advisor), Marlow, William H (committee member), Chen, Hamn-Ching (committee member).
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
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7940
► 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)
▼ 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. A prototypic HTGR, the Next Generation Nuclear Plant (NGNP),will be built at Idaho National Laboratory.The need for HTGR analysis tools and methods has led to the addition of gas-cooled reactor (GCR) capabilities to the light water reactor code MELCOR. MELCOR will be used by the Nuclear Regulatory Commission licensing of the NGNP and other HTGRs. In the present study, new input techniques have been developed
for MELCOR HTGR analysis. These new techniques include methods for modeling radiation heat transfer between solid surfaces in an HTGR, calculating fuel and
cladding geometric parameters for pebble bed and prismatic block-type HTGRs, and selecting appropriate input parameters for the reflector component in MELCOR.
The above methods have been applied to input decks for a water-cooled reactor cavity cooling system (RCCS); the 400 MW Pebble Bed Modular Reactor (PBMR), the input for which is based on a code-to-code benchmark activity; and the High Temperature Test Facility (HTTF), which is currently in the design phase at Oregon State University. RCCS results show that MELCOR accurately predicts radiation heat transfer rates from the vessel but may overpredict convective heat transfer rates and RCCS coolant flow rates. PBMR results show that thermal striping from hot jets in the lower plenum during steady-state operations, and in the upper plenum during a pressurized loss of forced cooling accident, may be a major design concern. Hot jets could potentially melt control rod drive mechanisms or cause thermal stresses in
plenum structures.
For the HTTF, results will provide data to validate MELCOR for HTGR analyses. Validation will be accomplished by comparing results from the MELCOR representation of the HTTF to experimental results from the facility. The validation process can be automated using a modular code written in Python, which is described here.
Advisors/Committee Members: Vierow, Karen (advisor), Tsvetkov, Pavel (committee member), Ranjan, Devesh (committee member).
Subjects/Keywords: nuclear; htgr; vhtr; pbmr; httf; melcor
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APA ·
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MLA ·
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CSE |
Export
to Zotero / EndNote / Reference
Manager
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
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2010-05-7767
► 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)
▼ 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 Reactor Cavity Cooling System (RCCS) is an important heat removal system in case of accidents. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The commercial Computational Fluid Dynamics (CFD) STAR-CCM+/ V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS.
Two models were developed to analyze heat exchange in the RCCS. Both models incorporate a 180 degree section resembling the
VHTR RCCS bench table test facility performed at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations.
Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls temperature below design limits.
Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions.
To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The models considered included the first-moment closure one equation Spalart-Allmaras model, the first-moment closure two-equation k-e and k-w models and the second-moment closure Reynolds Stress Transport (RST) model. For the near wall treatments, the low y+ and the all y+ wall treatments were considered. The two-layer model was also used to investigate the effect of near-wall treatment.
The comparison of the experimental data with the simulations showed a satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. The tested turbulence models demonstrated that the Realizable k-e model with two-layer all y+ wall treatment performs better than the other k-e models for such a complicated geometry and flow conditions. Results are in satisfactory agreement with the RST simulations and experimental data available.
A scaling analysis was developed to address the distortion introduced by the experimental facility and CFD model in simulating the physics inside the RCCS system with respect to the real plant configuration. The scaling analysis demonstrated that both the experimental facility and CFD model give a satisfactory reproduction of the main flow characteristics inside the RCCS cavity region, with convection and radiation heat exchange phenomena being properly scaled from the real plant to the model analyzed.
Advisors/Committee Members: Hassan, Yassin A. (advisor), Chen, Hamn-Ching (committee member), Tsvetkov, Pavel V. (committee member), Ugaz, Victor M. (committee member).
Subjects/Keywords: Turbulence; buoyancy; convection; RCCS; CFD; VHTR
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APA ·
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MLA ·
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CSE |
Export
to Zotero / EndNote / Reference
Manager
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
URL: http://hdl.handle.net/10394/15535
► 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)
▼ 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 is one of the system codes and it has been used to model the flow and heat transfer for a pebble fuel element and pebble-bed reactor. Although Flownex has been used to model the High Temperature Test Reactor, the prismatic block was modelled in a simplified manner. The aim of this study was to develop a more integrated model for a single block. A one sixth block was modelled in Flownex and the results were validated by comparing the results with results obtained using the Computational Fluid Dynamics (CFD) code STAR-CCM+.
The conduction heat transfer through the prismatic blocks containing the fuel elements in a Very High Temperature Reactor is of crucial importance for the proper operation of the reactor under normal operating conditions and upset conditions. In this study, a model developed in a system code, Flownex is discussed. The model comprised of a collection of 1-D solid conduction heat transfer, convection heat transfer and pipe elements that were arranged in such a manner to represent the heat transfer and fluid flow in the prismatic block using a network approach. The validity of the model was investigated by comparing the heat transfer and temperature distribution in the block for various scenarios with the corresponding values obtained using a detailed CFD model of one twelfth of a prismatic block.
Cubical and triangular block verification cases were conducted in Flownex and the results were validated by STAR-CCM+. The results were very comparable; however one issue has to be addressed. The one sixth integrated prismatic block was then modelled for a steady state and the results were also comparable. The outlet helium temperatures predicted by the STAR-CCM+ model was 542.94 C, at the same time the Flownex model predicted 542.98 C. Although the Flownex model did not provide the same detail as the STAR-CCM+ model the agreement between the results obtained with the two codes was satisfactory. Based on these findings it was concluded that Flownex could be used to build a representative integrated network model for a prismatic block reactor.
Subjects/Keywords: VHTR;
HTTR;
Flownex;
CFD;
STAR-CCM+
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APA ·
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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
URL: http://hdl.handle.net/10394/15542
► 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)
▼ 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 reactor concepts is the Very High Temperature gas-cooled Reactor (VHTR) of which the Reactor Cavity Cooling Systems (RCCS) is seen as an integral and crucial part of the passive safety concept. Considerable validation and development of the necessary software tools is required to perform analysis and designs of these future reactor concepts.
The primary objective of this study is to establish a methodology for the creation of an integrated system level process model of a typical air-cooled RCCS in FlownexΒ, and to illustrate its applicability by simulating different scenarios that illustrate the operational characteristics of such a system. For this purpose, the existing RCCS conceptual design that is being studied by the KAERI was used as the case study.
As a start, selected case studies were performed to verify that the FlownexΒ models were set up correctly to perform natural circulation flows, both in steady and transient conditions, and with radiation, convection and conduction taking part. These are the major typical physical phenomena in the RCCS. The models were compared with EES (Engineering Equation Solver) models of the same geometries and specifications. There was a good agreement between FlownexΒ and EES model results.
After this verification, a simulation model of the integrated RCCS system was developed. The FlownexΒ models were applied to model selected possible operational scenarios. The major observations from the results are that:
- The RCCS carries with it enough heat to the ambient such that the concrete wall temperature is maintained below the benchmark value of 65Β°C for the different boundary conditions imposed.
- The RCCS maintains its functionality even with three quarters of the risers blocked or in the event that there is a break in one of the chimney pipes.
Subjects/Keywords: VHTR;
RCCS;
Passive safety;
Simulation;
FlownexΒ
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Record Details
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APA ·
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MLA ·
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CSE |
Export
to Zotero / EndNote / Reference
Manager
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
URL: http://rave.ohiolink.edu/etdc/view?acc_num=osu1533356728728114
► 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)
▼ 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. This accident is initiated, in its worst case scenario, by
a double-ended guillotine break of its cross vessel. In an HTGR,
the reactor vessel is located within a reactor cavity that is
filled with air during normal operating conditions. During a D-LOFC
event followed by air ingress, an air-helium mixture may enter the
reactor vessel following a reactor vessel depressurization. Since
air chemically reacts with high-temperature graphite, this could
lead to damage of core-bottom and in-core graphite structures as
well as core heat-up, toxic gas release, and failure of the
structural integrity of the system unless mitigating action is
taken.Early studies postulated that the dominant mechanism of air
ingress is molecular diffusion, which is a slow process. However,
recent studies show that the air-ingress process could be initially
controlled by density-driven stratified flow of hot helium and a
relatively cooler air-helium mixture in the hot duct. If
density-driven stratified flow initially dominates, earlier onset
of natural circulation within the core would occur. This would lead
to an earlier onset of oxidation of internal graphite structures
and, most likely, at a more rapid rate. Thus, it is important to
understand both of these air-ingress mechanisms in a HTGR. These
mechanisms may be important at different times for different
scenarios, specifically breaks of varying size, orientation, shape,
and location. Also, understanding which ingress mechanism dominates
informs the type of mitigating measures that need to be considered
for HTGR designs.The principal question of this dissertation is to
determine whether density-driven stratified flow is the dominant
ingress mechanism during the initial stages of an air-ingress
accident in the unlikely event of a double-ended guillotine break
(DEGB). As a corollary, this dissertation also examines which
ingress mechanism dominates for a smaller break, or more
specically, an axial break (1/64th area of DEGB) in the hot duct
that is accompanied by failure of the cold duct enabling air
ingress from the containment. The results of this dissertation
demonstrate that during the initial time frame of the double-ended
guillotine and axial break density-driven stratified flow is the
dominant ingress mechanism.The rate at which ingress flow occurs is
different for each case. There is data to substantiate that the
rate of ingress is proportional to the cross-sectional flow area of
the break. Specifically, the ingress rate for the double-ended
guillotine break is much larger than the axial break oriented in
the top position (ABT). In the DEGB 400 C case, experimental
results show that the time required to reach an average oxygen
concentration of 10% in the plenum is approximately 45 s. Whereas
in the ABT 300 C case, experimental results show that the time
required to reach an average oxygen concentration of 10% in…
Advisors/Committee Members: Christensen, Richard N. (Advisor), Sun, Xiaodong (Advisor).
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.
GARCΓA, JesΓΊs Alberto Rosales.
Modelagem detalhada de sistemas nucleares avanΓ§ados do tipo leito de bolas com combustΓvel encapsulado
.
Degree: 2015, Universidade Federal de Pernambuco
URL: http://repositorio.ufpe.br/handle/123456789/15170
► A sustentabilidade da energia nuclear dependerΓ‘, entre outros fatores, da capacidade de redução dos inventΓ‘rios dos resΓduos nucleares de vida longa. Com esse objetivo, desenvolveu-se…
(more)
▼ A sustentabilidade da energia nuclear dependerΓ‘, entre outros fatores, da
capacidade de redução dos inventΓ‘rios dos resΓduos nucleares de vida longa. Com esse
objetivo, desenvolveu-se a nova geração de reatores nucleares, com seis protótipos que
se destacam por sua seguranΓ§a, resistΓͺncia Γ proliferação e a gestΓ£o dos resΓduos. Dentro
dessa nova geração de reatores, encontram-se os reatores de temperatura muito alta
(
VHTR), pela capacidade de produzir energia e a obtenção de altas temperaturas na saΓda
do refrigerante, para seu uso em aplicaçáes de alta temperatura como a produção de
hidrogΓͺnio. Os ADS (Accelerator Driven Systems), que podem ser projetados como
VHTR,
são sistemas projetados para a redução dos elementos transurÒnicos provenientes dos
LWRs (Light Water Reactors).
O TADSEA (Transmutation Advanced Device for sustainable Energy Applications)
Γ© um ADS do tipo leito de bolas, projetado para atingir uma queima profunda dos
elementos transurÒnicos, a produção colateral de energia e a obtenção de altas
temperaturas para produzir hidrogΓͺnio. O presente trabalho tΓͺm como objetivo realizar
melhoras no projeto conceitual do TADSEA, através da simulação geométrica detalhada
do combustΓvel, para o qual foi desenvolvida e avaliada uma metodologia para a
modelagem computacional detalhada da dupla heterogeneidade do combustΓvel em um
leito de bolas, usando o cΓ³digo probabilista MCNPX. Foram incluΓdos novos elementos no
projeto como a blindagem, as barras absorvedoras para garantir a segurança do sistema,
e foi avaliado o desempenho na redução dos resΓduos e sua radiotoxicidade associada,
assim como a produção de energia.
Advisors/Committee Members: LIRA, Carlos Alberto Brayner de Oliveira (advisor).
Subjects/Keywords: Engenharia de Energia Nuclear;
VHTR;
ADS;
transmutação;
HTR-10
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Manager
APA (6th Edition):
GARCΓA, J. A. R. (2015). Modelagem detalhada de sistemas nucleares avanΓ§ados do tipo leito de bolas com combustΓvel 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):
GARCΓA, JesΓΊs Alberto Rosales. “Modelagem detalhada de sistemas nucleares avanΓ§ados do tipo leito de bolas com combustΓvel 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):
GARCΓA, JesΓΊs Alberto Rosales. “Modelagem detalhada de sistemas nucleares avanΓ§ados do tipo leito de bolas com combustΓvel encapsulado
.” 2015. Web. 24 Jan 2021.
Vancouver:
GARCΓA JAR. Modelagem detalhada de sistemas nucleares avanΓ§ados do tipo leito de bolas com combustΓvel 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:
GARCΓA JAR. Modelagem detalhada de sistemas nucleares avanΓ§ados do tipo leito de bolas com combustΓvel 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
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2010-08-8549
► 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)
▼ 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 Transuranic (TRU) inventories. The method has been
implemented in a developed code system integrating capabilities of MCNPX for highfidelity
fuel cycle component simulations.
The impact associated with energy generation and utilization is immeasurable due
to the immense, widespread, and myriad effects it has on the world and its inhabitants.
The polar extremes are demonstrated on the one hand, by the high quality of life enjoyed
by individuals with access to abundant reliable energy sources, and on the other hand by
the global-scale environmental degradation attributed to the affects of energy production
and use. Thus, nations strive to increase their energy generation, but are faced with the
challenge of doing so with a minimal impact on the environment and in a manner that is
self-reliant. Consequently, a revival of interest in nuclear energy has followed with much
focus placed on technologies for transmuting nuclear spent fuel.
In this dissertation, a Nuclear Energy System (NES) configuration was developed
to take advantage of used fuel recycling and transmutation capabilities in waste
management scenarios leading to minimized TRU waste inventories, long-term activities,
and radiotoxicities. The reactor systems and fuel cycle components that make up the
NES were selected for their ability to perform in tandem to produce clean, safe, and
dependable energy in an environmentally conscious manner. The reactor systems include
the AP1000,
VHTR, and HEST. The diversity in performance and spectral
characteristics for each was used to enhance TRU waste elimination while efficiently
utilizing uranium resources and providing an abundant energy source.
The High Level Waste (HLW) stream produced by typical nuclear systems was
characterized according to the radionuclides that are key contributors to long-term waste
management issues. The TRU component of the waste stream becomes the main
radiological concern for time periods greater than 300 years. A TRU isotopic assessment
was developed and implemented to produce a priority ranking system for the TRU
nuclides as related to long-term waste management and their expected characteristics
under irradiation in the different reactor systems of the NES.
Detailed 3D whole-core models were developed for analysis of the individual
reactor systems of the NES. As an inherent part of the process, the models were
validated and verified by performing experiment-to-code and/or code-to-code
benchmarking procedures, which provided substantiation for obtained data and results.
Reactor core physics and material depletion calculations were performed and analyzed.
A computational modeling approach was developed for integrating the individual
models of the NES. A general approach was utilized allowing for the Integrated System
Model (ISM) to be modified in order to provide simulation for other systems with similar…
Advisors/Committee Members: Tsvetkov, Pavel V. (advisor), Peddicord, Kenneth L. (committee member), McDeavitt, Sean M. (committee member), Allen, Graham D. (committee member).
Subjects/Keywords: Nuclear; Optimization; VHTR; AP100; Recycle; Environment; Sensitivity; TRU
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Manager
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
URL: http://hdl.handle.net/1969.1/173485
► 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)
▼ 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 strenuous conditions than older generation II designs which can be detrimental to data collecting equipment; therefore, equipment must be tested under comparable conditions to simulate how it would respond and operate. This research focuses on computationally altering Texas A&Mβs TRIGA reactor neutronics using MCNP to compare against advanced reactor neutronics.
Several energy spectrums of interest were collected using MCNP for the FHR and
VHTR over the active fuel region, coolant, graphite (in the active core), and graphite reflector. The energy spectrum of the TRIGAβs in-core irradiation location was collected with a thermal peak centerline energy (the energy at which the peak is located) considerably lower than the FHR and
VHTR in all locations. The in-core irradiation location was altered by incorporating various moderators, temperatures, and neutron absorbers and then compared, using the thermal peak centerline energy and the general spectrum shape, to determine the likeness of the altered spectrum to the advanced reactor spectrums.
Based on the collected data, the FHR and
VHTR core characteristics were best represented with high-temperature, 900 K, graphite in the TRIGAβs irradiation location D1. In addition, all regions of interest where spectrums were found (active core, coolant, graphite, and reflector) can be represented in the same locations. Once the peaks were adequately matched, reactor similarity factors were found which can be used to convert experimental data into predicted FHR or
VHTR results in the case of an ensuing experiment.
Advisors/Committee Members: Tsvetkov, Pavel V. (advisor), Kirkland, Karen V. (committee member), Pate, Michael B. (committee member).
Subjects/Keywords: TRIGA; FHR; VHTR; energy spectrum; MCNP; similarity factors
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APA ·
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MLA ·
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CSE |
Export
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Manager
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
URL: http://hdl.handle.net/1969.1/155387
► 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)
▼ 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 the single-phase flow behavior of the system under a variety of experimental conditions. Base level conditions used a uniform heat flux at a power level appropriately scaled from a benchmark computer simulation of the Gas Turbine Modular Helium Reactor (GT-MHR) using scaling relationships derived by Argonne National Laboratory. Experiments were setup to gauge the effects of flow throttling, non-uniform heat flux profiles, alternate power levels and alternate coolant inventory levels on the flow distribution in the Cooling Panel, and to investigate the relationships between system variables of applied power, the temperature difference across the Cooling Panel (ΞT) and flowrate. In addition, a single scoping experiment was executed to observe system performance with coolant at the saturation temperature.
The system variables proved to have highly linear relationships amongst each other under all experimental conditions. Flow instabilities were observed in the form of counter-phase sinusoidal oscillations of flowrate and ΞT, the frequency thereof showed a roughly linear relationship with power. Ultrasonic Velocity Profiling (UVP) was used to determine the flow distribution, which increased at the outlet side of the panel with either increased system flowrate or higher heat flux applied to the outlet side, and vice-versa. The effect caused by flowrate changes was the same whether due to a change in power level or throttling, indicating the fluidβs momentum is the driving factor.
The phenomenon of sudden, high velocity, short duration flow excursions, called geysering, was observed as the system coolant was brought to saturation. This was caused by the trapping of non-condensable gases in the top horizontal section of the flow loop, which in turn brought the flowrate down considerably, increasing residence time and temperature of the coolant in the Cooling Panel. Subsequent rise of saturated coolant to a higher elevation in the hot leg resulted in flashing of the coolant to steam, whose sudden expansion drove the flow excursion.
Advisors/Committee Members: Hassan, Yassin (advisor), Marlow, William (committee member), King, Maria (committee member).
Subjects/Keywords: RCCS; natural convection; VHTR; geysering; multiphase flow; UVP; ultrasonic velocity profiling
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APA ·
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MLA ·
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CSE |
Export
to Zotero / EndNote / Reference
Manager
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
URL: http://hdl.handle.net/1969.1/4160
► 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)
▼ 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 there is an urgent need for development, testing and
validation of design tools to demonstrate the feasibility of the design concepts and guide
the improvement of the plant components. One of the identified design issues for the
gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet
plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream
components. This analysis was performed by coupling a RELAP5-3DΓΒΓΒ©
VHTR model to
a Fluent outlet plenum model. The RELAP5
VHTR model outlet conditions provide the
inlet boundary conditions to the Fluent outlet plenum model. By coupling the two codes
in this manner, the important three-dimensional flow effects in the outlet plenum are
well modeled without having to model the entire reactor with a computationally
expensive code such as Fluent. The two codes were successfully coupled. The values of
pressure, mass flow rate and temperature across the coupled boundary showed only
slight differences. The coupling tool used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the
domain.
Advisors/Committee Members: Hassan, Yassin A. (advisor), Annamalai, Kalyan (committee member), Marlow, William H. (committee member).
Subjects/Keywords: VHTR; RELAP5-3D; Fluent
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APA ·
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Export
to Zotero / EndNote / Reference
Manager
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 geomΓ©trico detalhado para a modelagem termoidrΓ‘ulica de sistemas nucleares, do tipo leito de bolas
.
Degree: 2016, Universidade Federal de Pernambuco
URL: https://repositorio.ufpe.br/handle/123456789/23485
► A tecnologia VHTR (do inglΓͺs Very High Temperature Reactor, Reator de Temperatura Muito Elevada) representa o prΓ³ximo estΓ‘gio na evolução dos reatores HTGR (do inglΓͺs…
(more)
▼ A tecnologia
VHTR (do inglΓͺs Very High Temperature Reactor, Reator de Temperatura Muito
Elevada) representa o prΓ³ximo estΓ‘gio na evolução dos reatores HTGR (do inglΓͺs High Temperature
Gas-Cooled Reactor, Reator de Alta Temperatura Refrigerado a GΓ‘s). Moderados a grafite e
refrigerados a hélio, os sistemas VHTRs podem ser usados para a cogeração de calor e de
eletricidade com temperaturas de saΓda entre 700 e 950 ΒΊC, e potencialmente com mais de 1.000 ΒΊC
no futuro. A temperatura do combustΓvel durante toda a operação do reator Γ© um aspecto muito
importante para a segurança dos reatores nucleares, no projeto deseja-se que seja menor que um
valor limite para garantir a integridade dos materiais do elemento combustΓvel evitando a liberação
de produtos de fissΓ£o. O TADSEA (Transmutation Advanced Device for Sustainable Energy
Applications) Γ© um
VHTR do tipo leito de bolas, projetado para atingir uma queima profunda dos
elementos transurÒnicos, a produção colateral de energia e a obtenção de altas temperaturas para
produzir hidrogΓͺnio. O presente trabalho tem como objetivo o desenvolvimento de uma metodologia
para a anΓ‘lises termoidrΓ‘ulica do nΓΊcleo de reatores do tipo leito de bolas de muito alta temperatura,
baseada no uso de uma abordagem realΓstica com um cΓ³digo de DinΓ’mica dos Fluidos
Computacional (CFD). Inicialmente, usando o modelo realΓstico da coluna com altura inteira do
reator HTR-10 com cΓ©lulas FCC e BCC, foram comparados os resultados obtidos com dados
experimentais e de simulação para a primeira tarefa de referΓͺncia do HTR-10 disponibilizados pela
IAEA (2013) para validação do modelo. No reator TADSEA, foram comparados resultados dos
projetos inicial e atual do nΓΊcleo com uma coluna com a altura completa do reator na regiΓ£o de
maior potΓͺncia. A partir dos resultados o projeto inicial nΓ£o tem margem de seguranΓ§a suficiente
para casos de perda de refrigerante. Nas simulaçáes do projeto atual do TADSEA as temperaturas
mΓ‘ximas atingidas foram muito inferiores ao limite. E os resultados de casos de perda de refrigerante
mostram que com 45% do fluxo mΓ‘ssico Γ© atingida uma temperatura apenas 30 K abaixo do limite.
Advisors/Committee Members: LIRA, Carlos Alberto Brayner de Oliveira (advisor), http://lattes.cnpq.br/3035514390746549 (advisor).
Subjects/Keywords: Energia Nuclear;
VHTR;
CFD;
TermoidrΓ‘ulica nuclear;
HTR-10
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Manager
APA (6th Edition):
ROJAS MAZAIRA, L. Y. (2016). Desenvolvimento de um modelo geomΓ©trico detalhado para a modelagem termoidrΓ‘ulica 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 geomΓ©trico detalhado para a modelagem termoidrΓ‘ulica 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 geomΓ©trico detalhado para a modelagem termoidrΓ‘ulica de sistemas nucleares, do tipo leito de bolas
.” 2016. Web. 24 Jan 2021.
Vancouver:
ROJAS MAZAIRA LY. Desenvolvimento de um modelo geomΓ©trico detalhado para a modelagem termoidrΓ‘ulica 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 geomΓ©trico detalhado para a modelagem termoidrΓ‘ulica 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.
AnΓ‘lise CFD do nΓΊcleo prismΓ‘tico do VHTR com distintos modelos de turbulΓͺncia e alteração de parΓ’metros da geometria
.
Degree: 2017, Universidade Federal de Pernambuco
URL: https://repositorio.ufpe.br/handle/123456789/25465
► O VHTR Γ© um reator nuclear tΓ©rmico, moderado a grafite e refrigerado por hΓ©lio. Para seu desenvolvimento, hΓ‘ a necessidade de utilização de ferramentas computacionais…
(more)
▼ O
VHTR é um reator nuclear térmico, moderado a grafite e refrigerado por hélio. Para seu desenvolvimento, hÑ a necessidade de utilização de ferramentas computacionais eficientes para a anÑlise de aspectos de modelagem, operação e segurança. A proposta deste trabalho é estudar o comportamento do
VHTR por meio de anΓ‘lise paramΓ©trica, alterando-se modelo de turbulΓͺncia, perfil de geração de energia nos blocos combustΓveis e a influΓͺncia de modificaçáes na prΓ³pria geometria. Busca-se tambΓ©m avaliar a implementação de uma metodologia simplificada que reduza o esforΓ§o computacional e a duração de uma simulação. Procedeu-se Γ anΓ‘lise do escoamento do fluido refrigerante atravΓ©s dos canais refrigerantes e canais by-pass em uma seção de 1/12 de uma coluna de blocos combustΓveis, utilizando-se diferentes modelos de turbulΓͺncia. Os resultados obtidos com essas simulaçáes foram comparados Γ queles obtidos por meio de correlaçáes do nΓΊmero de Nusselt descritos na literatura. Observou-se que a simulação na qual se utiliza o modelo π-Ξ΅ possibilita a obtenção de resultados que convergem bem com aqueles fornecidos pelas correlaçáes, para ambos os tipos de canais. O modelo π-Ο proporciona bons resultados para os canais refrigerantes e, o SSG, para o canal by-pass. Utilizou-se geometria contendo canais by-pass de diferentes dimensΓ΅es, alΓ©m de uma que possuΓa apenas os canais refrigerantes, sem canal by-pass. Verificou-se que a existΓͺncia de um escoamento by-pass induz a um aumento no gradiente de temperatura no bloco combustΓvel. Realizaram-se estudos comparativos entre os resultados obtidos em simulaçáes realizadas com diferentes perfis de geração de energia tΓ©rmica (uniforme e senoidal) nos canais combustΓveis. Verificou-se que, quando hΓ‘ a mesma geração de energia tΓ©rmica total no bloco combustΓvel, a mΓ‘xima temperatura constatada em cada um dos materiais Γ© menor para o caso da geração de energia com perfil senoidal. Quando utilizado, no perfil senoidal, um fator radial de pico (1,25), hΓ‘ um aumento considerΓ‘vel na temperatura de todos os materiais, possibilitando a ocorrΓͺncia de regiΓ΅es em que a temperatura pode ultrapassar o limite usualmente aceito para o combustΓvel do reator (1250Β°C) em operação normal. O canal refrigerante localizado no centro do bloco combustΓvel tem diΓ’metro inferior aos demais canais existentes nesse bloco. Para verificar a hipΓ³tese de que a existΓͺncia de um gradiente de temperatura no bloco combustΓvel, com a temperatura mais elevada ao centro e a temperatura mais baixa estando na periferia desse bloco, deve-se fortemente Γ menor dimensΓ£o desse canal central, realizaram-se simulaçáes computacionais utilizando-se uma geometria com canal central de diΓ’metro igual ao dos demais. A condição de entrada escolhida para essa nova estrutura foi, primeiramente, o mesmo fluxo mΓ‘ssico total e, depois, a mesma diferenΓ§a de pressΓ£o entre entrada e saΓda verificados na simulação da geometria padrΓ£o. Os resultados obtidos confirmam a hipΓ³tese aventada. Realizou-se simulação utilizando uma metodologia simplificada, que…
Advisors/Committee Members: LIRA, Carlos Alberto Brayner de Oliveira (advisor), http://lattes.cnpq.br/3035514390746549 (advisor).
Subjects/Keywords: FluidodinΓ’mica computacional;
VHTR;
AnΓ‘lise paramΓ©trica;
Acoplamento 1D-3D;
TurbulΓͺncia
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APA (6th Edition):
PAIVA, P. P. D. d. S. (2017). AnΓ‘lise CFD do nΓΊcleo prismΓ‘tico do VHTR com distintos modelos de turbulΓͺncia e alteração de parΓ’metros 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. “AnΓ‘lise CFD do nΓΊcleo prismΓ‘tico do VHTR com distintos modelos de turbulΓͺncia e alteração de parΓ’metros 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. “AnΓ‘lise CFD do nΓΊcleo prismΓ‘tico do VHTR com distintos modelos de turbulΓͺncia e alteração de parΓ’metros da geometria
.” 2017. Web. 24 Jan 2021.
Vancouver:
PAIVA PPDdS. AnΓ‘lise CFD do nΓΊcleo prismΓ‘tico do VHTR com distintos modelos de turbulΓͺncia e alteração de parΓ’metros 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. AnΓ‘lise CFD do nΓΊcleo prismΓ‘tico do VHTR com distintos modelos de turbulΓͺncia e alteração de parΓ’metros da geometria
. [Masters Thesis]. Universidade Federal de Pernambuco; 2017. Available from: https://repositorio.ufpe.br/handle/123456789/25465
16.
GΓMEZ RODRΓGUEZ, Abel.
Uma metodologia termo-fluido-dinΓ’mica computacional para avaliação de reatores que operam a altΓssimas temperaturas com leitos de combustΓveis esfΓ©ricos
.
Degree: 2019, Universidade Federal de Pernambuco
URL: https://repositorio.ufpe.br/handle/123456789/36154
► O crescimento da população mundial, a dependΓͺncia dos combustΓveis fΓ³sseis, a crescente demanda de energia por parte dos paΓses em desenvolvimento, e os problemas associados…
(more)
▼ O crescimento da população mundial, a dependΓͺncia dos combustΓveis fΓ³sseis, a crescente demanda de energia por parte dos paΓses em desenvolvimento, e os problemas associados Γ emissΓ£o de gases de efeito estufa, sΓ£o algumas das razΓ΅es pelas quais a sociedade procura melhorar as tecnologias de produção de energia existentes. A energia nuclear com a nova Geração IV de reatores nucleares oferece uma solução para enfrentar o problema do crescimento da demanda mundial de energia. Esta tecnologia nΓ£o emite gases de efeito estufa e permite construir usinas de potΓͺncia elevada ou de pequenas potΓͺncias, um aspecto que supera a maioria das fontes renovΓ‘veis de energia. Outrossim, contribui satisfatoriamente com os avanΓ§os na sustentabilidade, seguranΓ§a, confiabilidade e resistΓͺncia Γ proliferação de armas nucleares. O Reator de Temperatura Muito Alta (
VHTR) Γ© um dos candidatos da prΓ³xima geração de reatores nucleares, de acordo com a IAEA. Prever o desempenho termoidrΓ‘ulico de reatores de temperatura alta Γ© uma contribuição importante para o desenvolvimento da tecnologia. A avaliação do comportamento termoidrΓ‘ulico de estados estacionΓ‘rios e transitΓ³rios do reator de teste de temperatura alta de leito de bolas refrigerado a gΓ‘s HTR-10, foi um desafio proposto Γ comunidade cientΓfica internacional pela IAEA. Este trabalho propΓ΅e uma metodologia para o estudo termoidrΓ‘ulico de estados estacionΓ‘rios e transitΓ³rios de reatores nucleares de temperatura muito alta de leito de bolas refrigerados a gΓ‘s, a partir de modelagem termoidrΓ‘ulica computacional tridimensional em escala real. AnΓ‘lises dos principais parΓ’metros termoidrΓ‘ulicos: temperatura dos elementos combustΓveis, do refrigerante, dos elementos estruturais, velocidades e pressΓ΅es foram realizadas. Estas anΓ‘lises foram realizadas a partir de estudos comparativos com dados experimentais e com dados obtidos por outros cΓ³digos computacionais. Foi comprovada a capacidade de predição dos principais parΓ’metros termoidrΓ‘ulicos a partir de dois modelos computacionais, um βmodelo simplificadoβ, com menor utilização de recursos computacionais que permite obter uma descrição aceitΓ‘vel da termoidrΓ‘ulica do reator HTR-10 e um segundo modelo mais abrangente, nomeado βmodelo integralβ que permite a determinação dos principais parΓ’metros termoidrΓ‘ulicos com uma maior exatidΓ£o a custo de maior utilização dos recursos computacionais. TambΓ©m foram avaliados os principais parΓ’metros termoidrΓ‘ulicos do reator HTR-10 durante o acidente postulado de falha do circulador de hΓ©lio (ATWS). Com a metodologia e o uso do modelo integral foram capturados os efeitos transitΓ³rios, de acordo com os experimentos, que demostram a seguranΓ§a passiva que dispΓ΅e este reator de temperatura alta de leito de bolas refrigerado a gΓ‘s.
Advisors/Committee Members: LIRA, Carlos Alberto Brayner de Oliveira (advisor), http://lattes.cnpq.br/3035514390746549 (advisor).
Subjects/Keywords: Engenharia nuclear;
VHTR;
TermoidrΓ‘ulica nuclear;
HTR-10;
CFD;
ANSYS CFX
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APA ·
Chicago ·
MLA ·
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CSE |
Export
to Zotero / EndNote / Reference
Manager
APA (6th Edition):
GΓMEZ RODRΓGUEZ, A. (2019). Uma metodologia termo-fluido-dinΓ’mica computacional para avaliação de reatores que operam a altΓssimas temperaturas com leitos de combustΓveis esfΓ©ricos
. (Doctoral Dissertation). Universidade Federal de Pernambuco. Retrieved from https://repositorio.ufpe.br/handle/123456789/36154
Chicago Manual of Style (16th Edition):
GΓMEZ RODRΓGUEZ, Abel. “Uma metodologia termo-fluido-dinΓ’mica computacional para avaliação de reatores que operam a altΓssimas temperaturas com leitos de combustΓveis esfΓ©ricos
.” 2019. Doctoral Dissertation, Universidade Federal de Pernambuco. Accessed January 24, 2021.
https://repositorio.ufpe.br/handle/123456789/36154.
MLA Handbook (7th Edition):
GΓMEZ RODRΓGUEZ, Abel. “Uma metodologia termo-fluido-dinΓ’mica computacional para avaliação de reatores que operam a altΓssimas temperaturas com leitos de combustΓveis esfΓ©ricos
.” 2019. Web. 24 Jan 2021.
Vancouver:
GΓMEZ RODRΓGUEZ A. Uma metodologia termo-fluido-dinΓ’mica computacional para avaliação de reatores que operam a altΓssimas temperaturas com leitos de combustΓveis esfΓ©ricos
. [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:
GΓMEZ RODRΓGUEZ A. Uma metodologia termo-fluido-dinΓ’mica computacional para avaliação de reatores que operam a altΓssimas temperaturas com leitos de combustΓveis esfΓ©ricos
. [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
URL: http://hdl.handle.net/1853/52196
► 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)
▼ 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 reactor designs under investigation by the nuclear industry include the General Atomics GT-MHR and the Department of Energy MHTGR-350. Both reactors use the same hexagonal prismatic block, TRISO fuel compact, and circular coolant channel array design.
Steady state temperature, pressure, and mass flow distributions are determined for the base reference designs and also for a range of values of the important parameters. Core temperature distributions are obtained with reduced computational cost over more highly detailed computational fluid dynamics codes by using efficient, correlations and first-principles-based approaches for the relevant thermal fluid and thermal transport phenomena. Full core 3-D heat conduction calculations are performed at the individual fuel pin and lattice assembly block levels. The fuel compact is treated as a homogeneous medium with heat generation. A simplified 1-D fluid model is developed to predict convective heat removal rates from solid core nodes. Downstream fluid properties are determined by performing a channel energy balance down the axial node length. Channel exit pressures are then compared and inlet mass flows are adjusted until a uniform outlet pressure is reached. Bypass gaps between assembly blocks as well as coolant channels are modeled. Finite volume discretization of energy, and momentum conservation equations are then formed and explicitly integrated in time. Iterations are performed until all local core temperatures stabilize and global convective heat removal matches heat generation.
Several important observations were made based on the steady state analyses for the MHTGR and GT-MHR. Slight temperature variation in the radial direction was observed for uniform radial powers. Bottom-peaked axial power distributions had slightly higher peak temperatures but lower core average temperatures compared to top and center-peaked power distributions. The same trend appeared for large bypass gap sizes cases compared to smaller gap widths. For all cases, peak temperatures were below expected normal operational limits for TRISO fuels. Bypass gap flow for a 3 mm gap width was predicted to be between 10 and 11% for both reactor designs. Single assembly hydrodynamic and temperature results compared favorably with those available in the literature for similar prismatic HTGR thermal hydraulic, computational fluid dynamics analyses.
The method developed here enables detailed local and core wide thermal analysis with minimal computational effort, enabling advanced coupled analyses of high temperature reactors with thermal feedback. The steady state numerical scheme also offers a potential for select transient scenario modeling and a wide variety of design optimization studies.
Advisors/Committee Members: Garimella, Srinivas (advisor), Rahnema, Farzad (committee member), Graham, Samuel (committee member).
Subjects/Keywords: Thermal hydraulics; VHTR; MHTGR
…Temperature Reactor (VHTR) Background
HTGRs are gas reactor systems with coolant outlet… …temperatures up to 850Β°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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Record Details
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APA ·
Chicago ·
MLA ·
Vancouver ·
CSE |
Export
to Zotero / EndNote / Reference
Manager
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
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2557
► 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)
▼ 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 repository at
Yucca Mountain. Ongoing efforts on closed-fuel cycle options and difficulties in
opening and safeguarding such a repository have led to investigations of alternative
waste management strategies. One potential strategy would make use of fuels containing
transuranic (TRU) nuclides in nuclear reactors. This would prolong reactor operation on
a single fuel loading and by doing so, would reduce current HLW stockpiles. The
analysis has already shown that high-temperature gas-cooled reactors (HTGRs) and their
Generation IV extensions, very-high-temperature reactors (VHTRs), have encouraging
performance characteristics that will allow for prolonged operation with no intermediate
refueling, as well as for transmutation of TRUs.
The objective of this research was to show that TRU-fueled VHTRs have the possibility
of prolonged operation on a single fuel loading while retaining their Generation IV safety
features. In addition, this research evaluated performance characteristics, and identified
operational domains of these systems, as well as the possibility of HLW reduction.
A whole-core, 3-D model of a power size prismatic
VHTR with a detailed temperature
distribution was developed for calculations with the SCALE 5.1 code package. Results
of extensive criticality and depletion calculations with multiple fuel loadings showed that
VHTRs are capable and suitable for autonomous operation when loaded with TRU fuel.
Advisors/Committee Members: Tsvetkov, Pavel V. (advisor), Charlton, William S. (committee member), Petrova, Guergana (committee member), Poston, John W. (committee member).
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 ·
Chicago ·
MLA ·
Vancouver ·
CSE |
Export
to Zotero / EndNote / Reference
Manager
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
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2559
► 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)
▼ 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 one of these targeted wastes. Partitioning and transmutation in fission reactors is one
possible incineration option and could potentially serve as a source of nuclear fuel
required for sustainability of energy resources.
The objective of this research was to evaluate the neutronic performance of the
pebble-bed Very High Temperature Reactor (
VHTR) configurations with various fuel
loadings. The configuration adjustments and design sensitivity studies specifically
targeted the achievability of spectral variations. The development of several realistic
full-core 3D models and validation of all modeling techniques used was a major part of
this research effort. In addition, investigating design sensitivities helped identify the
parameters of primary interest.
The full-core 3D models representing the prototype and large scale cores were
created for use with SCALE 5.0 and SCALE 5.1 code systems. Initially the models
required the external calculation of a Dancoff correction factor; however, the recent release of SCALE 5.1 encompassed inherent double heterogeneity modeling capabilities.
The full core 3D models with multi-heterogeneity treatments are in agreement with
available pebble-bed High Temperature Test Reactor data and were validated through
benchmark studies. Analyses of configurations with various fuel loadings have
indicated promising performance and safety characteristics. It was found that through
small configuration adjustments, the pebble-bed design can be tweaked to produce
desirable spectral shifts. The future operation of Generation IV nuclear energy systems
would be greatly facilitated by the utilization of minor actinides as a fuel component.
This would offer development of new fuel cycles, and support sustainability of a fuel
source.
Advisors/Committee Members: Tsvetkov, Pavel V. (advisor), Allen, Donald (committee member), Horvat, Vladimir (committee member), McDeavitt, Sean (committee member).
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 ·
Chicago ·
MLA ·
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CSE |
Export
to Zotero / EndNote / Reference
Manager
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.
MateriΓ‘ly pro reaktory IV. generace: Materials for IV. generation power plants.
Degree: 2018, Brno University of Technology
URL: http://hdl.handle.net/11012/32439
► The aim of this thesis is to describe material needs of chosen Generation IV nuclear reactors β VHTR, SCWR, and MSR with the emphasis on…
(more)
▼ The aim of this thesis is to describe material needs of chosen Generation IV nuclear reactors β
VHTR, SCWR, and MSR with the emphasis on the
VHTR. It describes the main principle and a constructional arrangement of
VHTR components with its material demands. SCWR and MSR reactors are described in terms of its material requirements. The middle part of this thesis is dedicated to the available material candidate analysis. Main material properties are discussed and at the end of this part the suitability of chosen materials for a flange joint in
VHTR heat exchanger is assessed. The last chapter (the practical part) is focused on an experimental proposal of changeability of older materials by the new ones using the electron beam welding method.
Advisors/Committee Members: Ε najdΓ‘rek, Ladislav (advisor), Svoboda, Pavel (referee).
Subjects/Keywords: JadernΓ½; reaktor; IV. generace; VHTR; materiΓ‘l; SCWR; MSR; jadernΓ‘ elektrΓ‘rna; 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). MateriΓ‘ly 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. “MateriΓ‘ly 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. “MateriΓ‘ly pro reaktory IV. generace: Materials for IV. generation power plants.” 2018. Web. 24 Jan 2021.
Vancouver:
Hovorka M. MateriΓ‘ly 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. MateriΓ‘ly 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
URL: http://hdl.handle.net/1969.1/157998
► 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)
▼ 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) is a major accident scenario in which the primary coolant loop is broken, resulting in a loss of forced circulation of helium into the reactor vessel. With the onset of natural circulation, coolant flow reverses and is driven by buoyancy forces. The goal of the research is to simulate this accident condition on a 1/16th scaled model and visualize the flow behavior in the upper plenum of the
VHTR. The facility was designed and constructed from a set of scaling parameters and outfitted with various instrumentation to characterize the depressurized conduction cooldown (DCC) event. Particle image velocimetry (PIV) is a nonintrusive optical laser technique used to obtain an instantaneous velocity field and was successfully applied to this system. Throughout the preliminary tests, the number of frames to be averaged to reach a statistically steady state was obtained from 1,000 images. The performance of the PIV method is validated with a flowmeter and analytic flowrate equation. The uncertainty of PIV system was also quantified.
Single jet tests are performed to provide a basic understanding of the simplest turbulent buoyant jet mixing in the upper plenum. By the Morton length scale, it was observed that the buoyant jet behaves like a plume and self-similarity is obtained for the axial velocity profiles. Q criterion is applied to identify the eddy structures of the turbulent jet mixing as a way to characterize the mechanism of vortex-pair mixing on the dome surface. Subsequent triple jet experiments are performed and compared with the results from single jet tests. Velocity distributions along the concave wall show that higher wall shear stress is obtained in single jet tests. The experiment results will provide the benchmark data for the PIV validation.
Advisors/Committee Members: Anand, Nagamangala K. (advisor), Hassan, Yassin A. (advisor), Chen, Hamn-Ching (committee member), Lau, Sai C. (committee member).
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
URL: http://hdl.handle.net/1969.1/154093
► 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)
▼ 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 (
VHTR). Texas A&M University built a 1/8 scale experimental facility for the air-cooled Reactor Cavity Cooling System (RCCS) based on General Electric Modular High Temperature Gas Cooled Reactor (MHTGR) design to study the thermal hydraulic phenomena occurring in the upper plenum. The facility consists of four vertical parallel riser ducts welded to the upper plenum which has two exhaust chimneys. Blowers are used to drive air through in-line heaters which are connected to the bottom end of the riser ducts. Experiments were conducted to measure the temperature spatial profile in the plenum. Type T thermocouples were mounted on six moveable racks inside the upper plenum, which were moved during the experiments to measure the temperature profile across 6 different planes. Measurements were taken for four different cases with different boundary conditions. Two cases operated with heated air flow in all four risers, whereas the other two were performed with flow in a single riser only. The obtained temperature profiles were asymmetric and suggested the presence of reverse flow from one of the chimneys in both single riser cases and in one of the four riser cases. The other four riser case exhibited a symmetric temperature spatial profile indicating even distribution of the flow across the exhaust chimneys.
Advisors/Committee Members: Hassan, Yassin (advisor), Marlow, William (committee member), King, Maria (committee member).
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
URL: http://hdl.handle.net/1928/23215
► 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)
▼ 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 cooled reactor that operates at an exit temperature of up to 1273 K, making it ideal for generating electricity at a plant thermal efficiency upwards of 48% and the co-generation of process heat for hydrogen production and other industrial uses. Extensive thermal-hydraulics and safety analyses of VHTRs are being conducted using Computational Fluid Dynamics (CFD) and heat transfer codes, in conjunction with experiments and prototype demonstrations. These analyses are challenging, largely due to the 3-D simulation of the helium flow in the 10 m long coolant channels in the reactor core and the need to examine the effects of helium bypass flow in the interstitial gaps between the core fuel elements. This research, performed at the UNM-ISNPS, developed an effective thermal-hydraulics analyses methodology that markedly reduces the numerical meshing requirements and computational time. It couples the heliums 1-D convective flow and heat transfer in the channels to 3-D heat conduction in graphite and fuel compacts of
VHTR fuel elements. Besides the helium local bulk temperature, the heat transfer coefficient is calculated using a Nusselt number correlation, developed and validated in this work. In addition to omitting the numerical meshing in the coolant channels, the simplified analysis methodology effectively decreases the total computation time by a factor of ~ 33 - 40 with little effect on the calculated temperatures (< 5 K), compared to a full 3-D thermal-hydraulics analysis. The developed convective heat transfer correlation accounts for the effect of entrance mixing in the coolant channels, where z/D < 25. The correlation compares favorably, to within + 12%, with Taylor's (based on high temperature hydrogen heat transfer) and to within + 2% of the calculated results for full 3-D analyses of a
VHTR single channel module and multiple channels in the fuel elements. The simplified methodology is used to investigate the effects of helium bypass flow in interstitial gaps between fuel elements and of the helium bleed flow in control rod channels on calculated temperatures in the
VHTR fuel elements. Thermal-hydraulics analysis of a one-element high and of a full height
VHTR 1/6 core are also conducted. Results show that the interstitial bypass flow increases the temperatures near the center of the core fuel elements by 10-15 K, while reducing the temperatures along the edges of the elements by ~30 K. Without bypass flow, hotspots may occur at the location of burnable poison rods in the fuel elements, depending on the assumed volumetric heat generation rate in the rods. The helium bleed flow through the control rod channels reduces temperatures near them by 2-5 K, and only slightly increases the temperatures within the rest of the core fuel elements. In the
VHTR 1/6 core thermal-hydraulics analysis, the helium bypass flow decreases the heat…
Advisors/Committee Members: El-Genk, Mohamed, Tournier, Jean-Michel, Rodriguez, Sal, El-Genk, Mohamed.
Subjects/Keywords: thermal-hydraulics; bypass flow; VHTR; Nusselt number correlation; helium coolant; prismatic fuel element
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MLA ·
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Export
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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
URL: http://hdl.handle.net/1853/41195
► 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)
▼ 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 rates in burnable poison (BP) pins. However, in traditional methods, multi-group cross sections are generated from single bundle calculations with specular reflection boundary condition, in which the energy spectral effect in the core environment is not taken into account. This approximation introduces errors to the absorption cross sections especially for BPs neighboring reflectors and control rods.
In order to correct the BP absorption cross sections in whole core diffusion calculations, energy spectrum reconstruction (ESR) methods have been developed to reconstruct the fine group spectrum (and in-core continuous energy spectrum). Then, using the reconstructed spectrum as boundary condition, a BP pin cell local transport calculation serves an imbedded module within the whole core diffusion code to iteratively correct the BP absorption cross sections for improved results.
The ESR methods were tested in a 2D prismatic High Temperature Reactor (HTR) problem. The reconstructed fine-group spectra have shown good agreement with the reference spectra. Comparing with the cross sections calculated by single block calculation with specular reflection boundary conditions, the BP absorption cross sections are effectively improved by ESR methods. A preliminary study was also performed to extend the ESR methods to a 2D Pebble Bed Reactor (PBR) problem. The results demonstrate that the ESR can reproduce the energy spectra on the fuel-outer reflector interface accurately.
Advisors/Committee Members: Rahnema, Farzad (Committee Chair), Petrovic, Bojan (Committee Member), Zhang, Dingkang (Committee Member).
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
URL: http://hdl.handle.net/1957/11999
► 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)
▼ 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 be built for the validation of computer codes for the safe operation of the
VHTR. The scaling of the prismatic core configuration has been analyzed previously for a large break loss of coolant accident. This is a scaling analysis for the pebble bed core configuration.
As part of the full scaling analysis, the bottom up scaling of the pebble bed core for pressure drop and radial heat transfer were conducted. Radiation is the dominant form of heat transfer at high temperatures and was scaled using the two methods of treating radiation in a packed bed of spheres. The results of scaling were compared using FLUENT, a computational fluid dynamics code, using the setup, run, and comparison of a 1/80 azimuthally and 1/4 radial full scale prototype and scaled model. The temperature profiles across the core under natural circulation like conditions were determined for both models. The model and prototype temperature profiles had significant variation at the boundary, but only a few degree variation away from the boundary. Additionally, the radiation transport equation and radiation conductivity were compared, and distortions quantified for the FLUENT models.
Advisors/Committee Members: Woods, Brian G. (advisor), Wu, Qiao (committee member).
Subjects/Keywords: VHTR; Pebble bed reactors Β β 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 NRCβs 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
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2012-05-10984
► 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)
▼ 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 constitutes a challenging multi-parametric optimization problem. The reactors and fuel cycle performance parameters may be strongly dependent on minor variations in the system's input data. Proven discrepancies in nuclear data evaluations could affect the validity of the system optimization metrics.
This study first analyzes various advanced AP1000-
VHTR fuel cycle scenarios by assessing their TRU destruction and their U consumption minimization capabilities, and by computing reactor performance parameters such as the time evolution of the effective multiplication factor keff, the reactors' energy spectrum or the isotopic composition/activity at EOL. The performance metrics dependence to prompt neutron fission spectrum discrepancies is then quantified to assess the viability of one strategy. Fission spectrum evaluations are indeed intensively used in reactors' calculations. Discrepancies higher than 10% have been computed among nuclear data libraries for energies above 8MeV for 235U.
TRU arising from a 3wt% 235U-enriched UO2-fueled AP1000 were incinerated in a
VHTR. Fuels consisting of 20%, 40% and 100% of TRU completed by UO2 were examined. MCNPX results indicate that up to 88.9% of the TRU initially present in a
VHTR fueled with 20% of TRU and 80% of ThO2 were transmuted. Additionally, the use of WgPu instead of RgPu should reduce the daily consumption of 235U by 1.3 and augment core lifetime.
To estimate the system metrics dependence to fission spectrum discrepancies and validate optimization studies outputs, the VTHR 235U fission spectrum distribution was altered successively in three manners. keff is at worst lowered by 1.7% of the reference value and the energy spectrum by 5% between 50meV and 2MeV when a significantly distorted fission spectrum tail is used. 233U, 236Pu and 237Pu inventories and activities are multiplied by 263, 523 and 34 but are still negligible compared to 239Pu mass or the total activity.
The AP1000-
VHTR system is in conclusion not dependent on the selected fission spectrum variations. TRU elimination optimization studies in AP1000-
VHTR systems will be facilitated by freeing performance metrics dependency from 1 input parameter.
Advisors/Committee Members: Tsvetkov, Pavel V. (advisor), McDeavitt, Sean M. (committee member), Ehlig-Economides, Christine (committee member).
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
URL: http://hdl.handle.net/1969.1/153485
► 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)
▼ 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 (HTGR) design to have a 1000 ^(0)C coolant outlet. This increases fuel efficiency and allows for other industrial applications. During the design process several studies are performed to develop safety codes for the reactor. One major accident of interest is the Pressurized Conduction Cooldown (PCC) scenario. The PCC scenario involves loss of forced coolant to the core but the loop stays pressurized. This results in a large buoyancy force that through natural convection reverses the flow of the core coolant loop to circulate into the upper plenum of the
VHTR. Computer codes may be developed to simulate the phenomenon that occurs in a PCC scenario, but benchmark data is needed to validate the simulations. There are currently no experimental models to provide benchmark data for the PCC scenario. This study will cover the design, construction, and testing of a 1/16th scaled model of a
VHTR that uses Particle Image Velocimetry (PIV) for flow visualization in the upper plenum. Three tests were run for a partially heated core at statistically steady state, and PIV was used to generate the velocity field of three naturally convective adjacent jets. Recirculation between the jets occurred until the jets reached the mixing point three cm from the outlet where turbulent mixing was observed. A sensitivity analysis was performed to confirm 1000 image pairs was sufficient to correctly represent the flow. The results were then validated by comparing the PIV results with experimental data and calculated values.
Advisors/Committee Members: Anand, Nagamangala (advisor), Hassan, Yassin (advisor), Chen, Hamn (committee member), Lau, Sai (committee member).
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
URL: http://hdl.handle.net/1969.1/4382
► 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)
▼ 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 worldwide. If implemented, these technologies could also be a
source of nuclear fuel materials required for sustainability of nuclear energy.
The objective of this research was to evaluate performance characteristics of Very
High Temperature Reactors (VHTRs) and their variations due to configuration
adjustments targeting achievability of spectral variations. The development of realistic
whole-core 3D
VHTR models and their benchmarking against experimental data was an
inherent part of the research effort. Although the performance analysis was primarily
focused on prismatic core configurations, 3D pebble-bed core models were also created
and analyzed.
The whole-core 3D models representing the prismatic block and pebble-bed cores
were created for use with the SCALE 5.0 code system. Each of the models required the
Dancoff correction factor to be externally calculated. The code system DANCOFF-MCThe whole-core/system 3D models with multi-heterogeneity treatments were
validated by the benchmark problems. Obtained results are in agreement with the
available High Temperature Test Reactor data. Preliminary analyses of actinide-fueled
VHTR configurations have indicated promising performance characteristics. Utilization
of minor actinides as a fuel component would facilitate development of new fuel cycles
and support sustainability of a fuel source for nuclear energy assuring future operation of
Generation IV nuclear energy systems.
was utilized to perform the Dancoff factor calculations.
Advisors/Committee Members: Tsvetkov, Pavel V. (advisor), Allen, G. Donald (committee member), Peddicord, Kenneth L. (committee member).
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
URL: http://hdl.handle.net/1969.1/ETD-TAMU-2556
► 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)
▼ 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 first 500 years. One of the most viable approaches suggests creating new transmutation fuels containing TRUs for use in thermal and fast nuclear reactors. Irradiation of TRUs results in their transmutation and ultimate incineration by fission. The objective of this thesis is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled Very High Temperature Reactor (
VHTR) systems. This effort was focused on the prismatic core configuration. The 3D core models were created for use in calculations with the SCALE 5.1 code system. As part of the research effort, basic nuclear characteristics of TRUs were taken into consideration. The potential variations of PWR spent fuel compositions were modeled with the International Atomic Energy Agency (IAEA) Nuclear Fuel Cycle Simulation System, VISTA. The
VHTR configurations with varying TRU compositions were analyzed assuming a single-batch core operation. Their performance was compared to the
VHTR cases with low enriched uranium (LEU). The analysis shows that TRUs can be effectively utilized in the
VHTR systems. The TRU-fueled VHTRs exhibit favorable performance characteristics.
Advisors/Committee Members: Tsvetkov, Pavel V. (advisor), Charlton, William S. (committee member), Hassan, Yassin A. (committee member), Pasciak, Joe (committee member).
Subjects/Keywords: VHTR; Spent nuclear fuel; TRU-fuel
…Control rod block dimensions ¦ ¦ ¦ ¦ ¦ ¦ ¦ ¦ ¦. 63
25
3-D whole-core VHTR model with… …spent fuel ¦ ¦ ¦ ¦ ¦ ¦β¦ 50
XVI
Selected TRU vectors from PWR spent nuclear fuel for VHTR… …analysisβ¦ 51
XVII
VHTR core specifications ¦ ¦ ¦ ¦ ¦ ¦ ¦ ¦ ¦β¦. 54
XVIII
Fuel assembly block… …for the TRU-fueled VHTR analysis ¦ ¦ ¦β¦ 73
XXVI
Beginning-of-life multiplication factors… …Cooled Reactor (SCWR).
Very High Temperature Reactor (VHTR).
I.A.5…
Record Details
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Record Details
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APA ·
Chicago ·
MLA ·
Vancouver ·
CSE |
Export
to Zotero / EndNote / Reference
Manager
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
URL: https://submit-etda.libraries.psu.edu/catalog/10843
► 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)
▼ 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 Convection (D-LOFC) for the Very High Temperature Gas-Cooled Reactor (
VHTR). This event involves the gravity driven ingress of air into the helium filled reactor vessel, ultimately leading to a possible oxidation of graphite components in the vessel. The present study performs a scoping experiment using water and brine as simulant fluids to study the exchange phenomena. To design the test apparatus, a scaling analysis is performed to maintain the exchange time ratio to be unity with the Gas Turbine Modular Helium Reactor (GT-MHR) reference system for the vertical standpipe break. The apparatus consists of two rectangular acrylic compartments connected by pipes and is designed to investigate the effects of the break angle and break length. The break angle is varied from horizontal to vertical at every 15 degrees for L/D = 0.63, 3.0 and 5.0. The volumetric exchange rate is obtained by measuring the time rate of change of mixture density using a hydrometer. A flow visualization study is performed to gain physical understanding of the phenomenon. In general, the present results show similar characteristic phenomena to those found in previous studies for the initial stage of ingress, where the mixture density changes linearly with time. As the ingress progresses, however, it is found that the mixing phenomena inside the compartment and the compartment geometry make significant impacts on the ingress rate. Unlike the previous studies, the present results show that the average exchange rate for the entire ingress event can be up to 70% lower than that obtained from the initial stage alone.
Advisors/Committee Members: Seungjin Kim, Thesis Advisor/Co-Advisor, Seungjin Kim, Thesis Advisor/Co-Advisor.
Subjects/Keywords: gravity driven exchange; vhtr; air ingress; buoyancy driven exchange; water brine
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Record Details
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APA ·
Chicago ·
MLA ·
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CSE |
Export
to Zotero / EndNote / Reference
Manager
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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Open Access Theses and Dissertations
