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You searched for +publisher:"Georgia Tech" +contributor:("Dr. Bojan Petrovic"). Showing records 1 – 3 of 3 total matches.

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Georgia Tech

1. Hartmangruber, David Patrick. Using MAVRIC sequence to determine dose rate to accessible areas of the IRIS nuclear power plant.

Degree: MS, Mechanical Engineering, 2010, Georgia Tech

The objective of this thesis is to determine and analyze the dose rate to personnel throughout the proposed IRIS nuclear power plant. To accomplish this objective, complex models of the IRIS plant have been devised, advanced transport theory methods employed, and computationally intense simulations performed. IRIS is an advanced integral, light water reactor with a 335 MWe expected power output (1000 MWth). Due to its integral design, the IRIS pressure vessel has a large downcomer region. The large downcomer and the neutron reflector provide a great deal of additional shielding. This increase in shielding ensures that the IRIS design easily accomplishes the regulatory dose limits for radiation workers. However, The IRIS project set enhanced objectives of further reducing the dose rate to significantly lower levels, comparable or below the limit allowed for general public. The IRIS nuclear power plant design is very compact and has a rather complex geometric structure. Programs that use conventional methods would take too much time or would be unable to provide an answer for such a challenging deep penetration problem. Therefore, the modeling of the power plant was done using a hybrid methodology for automated variance reduction implemented into the MAVRIC sequence of the SCALE6 program package. The methodology is based on the CADIS and FW-CADIS methods. The CADIS method was developed by J.C. Wagner and A. Haghighat. The FW-CADIS method was developed by J.C. Wagner and D. Peplow. Using these methodologies in the MAVRIC code sequence, this thesis shows the dose rate throughout most of the inhabitable regions of the IRIS nuclear power plant. This thesis will also show the regions that are below the dose rate reduction objective set by the IRIS shielding team. Advisors/Committee Members: Dr. Bojan Petrovic (Committee Chair), Dr. Dingkang Zhang (Committee Member), Dr. Nolan Hertel (Committee Member).

Subjects/Keywords: Dose rate; Variance reduction parameters; IRIS; MAVRIC; Nuclear power plants; Radiation dosimetry; Light water reactors; Radiation Dosage Statistics

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APA (6th Edition):

Hartmangruber, D. P. (2010). Using MAVRIC sequence to determine dose rate to accessible areas of the IRIS nuclear power plant. (Masters Thesis). Georgia Tech. Retrieved from http://hdl.handle.net/1853/37123

Chicago Manual of Style (16th Edition):

Hartmangruber, David Patrick. “Using MAVRIC sequence to determine dose rate to accessible areas of the IRIS nuclear power plant.” 2010. Masters Thesis, Georgia Tech. Accessed February 23, 2019. http://hdl.handle.net/1853/37123.

MLA Handbook (7th Edition):

Hartmangruber, David Patrick. “Using MAVRIC sequence to determine dose rate to accessible areas of the IRIS nuclear power plant.” 2010. Web. 23 Feb 2019.

Vancouver:

Hartmangruber DP. Using MAVRIC sequence to determine dose rate to accessible areas of the IRIS nuclear power plant. [Internet] [Masters thesis]. Georgia Tech; 2010. [cited 2019 Feb 23]. Available from: http://hdl.handle.net/1853/37123.

Council of Science Editors:

Hartmangruber DP. Using MAVRIC sequence to determine dose rate to accessible areas of the IRIS nuclear power plant. [Masters Thesis]. Georgia Tech; 2010. Available from: http://hdl.handle.net/1853/37123

2. Hon, Ryan Paul. Creation of a whole-core PWR benchmark for the analysis and validation of neutronics codes.

Degree: MS, Mechanical Engineering, 2013, Georgia Tech

This work presents a whole-core benchmark problem based on a 2-loop pressurized water reactor with both UOâ‚‚and MOX fuel assemblies. The specification includes heterogeneity at both the assembly and core level. The geometry and material compositions are fully described and multi-group cross section libraries are provided in 2, 4, and 8 group formats. Simplifications made to the benchmark specification include a Cartesian boundary, to facilitate the use of transport codes that may have trouble with cylindrical boundaries, and control rod homogenization, to reduce the geometric complexity of the problem. These modifications were carefully chosen to preserve the physics of the problem and a justification of these modifications is given. Detailed Monte Carlo reference solutions including core eigenvalue, assembly averaged fission densities and selected fuel pin fission densities are presented for benchmarking diffusion and transport methods. Three different core configurations are presented in the paper namely all-rods-out, all-rods-in, and some-rods-in. Advisors/Committee Members: Dr. Farzad Rahnema (Committee Chair), Dr. Bojan Petrovic (Committee Member), Dr. Dingkang Zhang (Committee Member).

Subjects/Keywords: Whole-core transport; Neutronics; PWR benchmarks; Nuclear reactors; Neutron transport theory

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APA (6th Edition):

Hon, R. P. (2013). Creation of a whole-core PWR benchmark for the analysis and validation of neutronics codes. (Masters Thesis). Georgia Tech. Retrieved from http://hdl.handle.net/1853/47611

Chicago Manual of Style (16th Edition):

Hon, Ryan Paul. “Creation of a whole-core PWR benchmark for the analysis and validation of neutronics codes.” 2013. Masters Thesis, Georgia Tech. Accessed February 23, 2019. http://hdl.handle.net/1853/47611.

MLA Handbook (7th Edition):

Hon, Ryan Paul. “Creation of a whole-core PWR benchmark for the analysis and validation of neutronics codes.” 2013. Web. 23 Feb 2019.

Vancouver:

Hon RP. Creation of a whole-core PWR benchmark for the analysis and validation of neutronics codes. [Internet] [Masters thesis]. Georgia Tech; 2013. [cited 2019 Feb 23]. Available from: http://hdl.handle.net/1853/47611.

Council of Science Editors:

Hon RP. Creation of a whole-core PWR benchmark for the analysis and validation of neutronics codes. [Masters Thesis]. Georgia Tech; 2013. Available from: http://hdl.handle.net/1853/47611

3. Gros, Emilien B. Liquid-Salt-Cooled Reactor start-up with natural circulation under Loss-of-Offsite-Power (LOOP) conditions.

Degree: MS, Nuclear and Radiological Engineering, 2012, Georgia Tech

The Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR) was modeled using the neutronics analysis code SCALE6.0 and the thermal-hydraulics and kinetics modeling code RELAP5-3D with objective to devise, analyze, and evaluate the feasibility and stability of a start-up procedure for this reactor using natural circulation of the coolant and under the Loss Of Offsite Power (LOOP) conditions. This Generation IV reactor design has been studied by research facilities worldwide for almost a decade. While neutronics and thermal-hydraulics analyses have been previously performed to show the performance of the reactor during normal operation and for shutdown scenarios, no study has heretofore been published to examine the active or passive start-up of the reactor. The fuel temperature (Doppler) and coolant density coefficient of reactivity of the LS-VHTR were examined using the CSAS6 module of the SCALE6.0 code. Negative Doppler and coolant density feedback coefficients were calculated. Two initial RELAP5 simulations were run to obtain the steady-state conditions of the model and to predict the changes of the thermal-hydraulic parameters during the shutdown of the reactor. Next, a series of step reactivity additions to the core were simulated to determine how much reactivity can be inserted without jeopardizing safety and the stability of the core. Finally, a start-up procedure was developed, and the restart of the reactor with natural convection of the coolant was simulated. The results of the simulations demonstrated the potential of a passive start-up of the LS-VHTR. Advisors/Committee Members: Dr. Bojan Petrovic (Committee Chair), Dr. Davor Grgic (Committee Member), Dr. Graydon Yoder (Committee Member), Dr. Scott Duncan (Committee Member), Dr. Srinivas Garimella (Committee Member).

Subjects/Keywords: Liquid-salt; Reactor; Start-up; Loss of offsite power; Molten salt reactors; Nuclear power plants Power supply; Electric power failures

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

APA (6th Edition):

Gros, E. B. (2012). Liquid-Salt-Cooled Reactor start-up with natural circulation under Loss-of-Offsite-Power (LOOP) conditions. (Masters Thesis). Georgia Tech. Retrieved from http://hdl.handle.net/1853/43745

Chicago Manual of Style (16th Edition):

Gros, Emilien B. “Liquid-Salt-Cooled Reactor start-up with natural circulation under Loss-of-Offsite-Power (LOOP) conditions.” 2012. Masters Thesis, Georgia Tech. Accessed February 23, 2019. http://hdl.handle.net/1853/43745.

MLA Handbook (7th Edition):

Gros, Emilien B. “Liquid-Salt-Cooled Reactor start-up with natural circulation under Loss-of-Offsite-Power (LOOP) conditions.” 2012. Web. 23 Feb 2019.

Vancouver:

Gros EB. Liquid-Salt-Cooled Reactor start-up with natural circulation under Loss-of-Offsite-Power (LOOP) conditions. [Internet] [Masters thesis]. Georgia Tech; 2012. [cited 2019 Feb 23]. Available from: http://hdl.handle.net/1853/43745.

Council of Science Editors:

Gros EB. Liquid-Salt-Cooled Reactor start-up with natural circulation under Loss-of-Offsite-Power (LOOP) conditions. [Masters Thesis]. Georgia Tech; 2012. Available from: http://hdl.handle.net/1853/43745

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