This collection includes most of the ASU Theses and Dissertations from 2011 to present. ASU Theses and Dissertations are available in downloadable PDF format; however, a small percentage of items are under embargo. Information about the dissertations/theses includes degree information, committee members, an abstract, supporting data or media.

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Description
A pressurized water reactor (PWR) nuclear power plant (NPP) model is introduced into Positive Sequence Load Flow (PSLF) software by General Electric in order to evaluate the load-following capability of NPPs. The nuclear steam supply system (NSSS) consists of a reactor core, hot and cold legs, plenums, and a U-tube

A pressurized water reactor (PWR) nuclear power plant (NPP) model is introduced into Positive Sequence Load Flow (PSLF) software by General Electric in order to evaluate the load-following capability of NPPs. The nuclear steam supply system (NSSS) consists of a reactor core, hot and cold legs, plenums, and a U-tube steam generator. The physical systems listed above are represented by mathematical models utilizing a state variable lumped parameter approach. A steady-state control program for the reactor, and simple turbine and governor models are also developed. Adequacy of the isolated reactor core, the isolated steam generator, and the complete PWR models are tested in Matlab/Simulink and dynamic responses are compared with the test results obtained from the H. B. Robinson NPP. Test results illustrate that the developed models represents the dynamic features of real-physical systems and are capable of predicting responses due to small perturbations of external reactivity and steam valve opening. Subsequently, the NSSS representation is incorporated into PSLF and coupled with built-in excitation system and generator models. Different simulation cases are run when sudden loss of generation occurs in a small power system which includes hydroelectric and natural gas power plants besides the developed PWR NPP. The conclusion is that the NPP can respond to a disturbance in the power system without exceeding any design and safety limits if appropriate operational conditions, such as achieving the NPP turbine control by adjusting the speed of the steam valve, are met. In other words, the NPP can participate in the control of system frequency and improve the overall power system performance.
ContributorsArda, Samet Egemen (Author) / Holbert, Keith E. (Thesis advisor) / Undrill, John (Committee member) / Tylavsky, Daniel (Committee member) / Arizona State University (Publisher)
Created2013
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Description
For a (N+1)-bus power system, possibly 2N solutions exists. One of these solutions

is known as the high-voltage (HV) solution or operable solution. The rest of the solutions

are the low-voltage (LV), or large-angle, solutions.

In this report, a recently developed non-iterative algorithm for solving the power-

flow (PF) problem using the holomorphic embedding

For a (N+1)-bus power system, possibly 2N solutions exists. One of these solutions

is known as the high-voltage (HV) solution or operable solution. The rest of the solutions

are the low-voltage (LV), or large-angle, solutions.

In this report, a recently developed non-iterative algorithm for solving the power-

flow (PF) problem using the holomorphic embedding (HE) method is shown as

being capable of finding the HV solution, while avoiding converging to LV solutions

nearby which is a drawback to all other iterative solutions. The HE method provides a

novel non-iterative procedure to solve the PF problems by eliminating the

non-convergence and initial-estimate dependency issues appeared in the traditional

iterative methods. The detailed implementation of the HE method is discussed in the

report.

While published work focuses mainly on finding the HV PF solution, modified

holomorphically embedded formulations are proposed in this report to find the

LV/large-angle solutions of the PF problem. It is theoretically proven that the proposed

method is guaranteed to find a total number of 2N solutions to the PF problem

and if no solution exists, the algorithm is guaranteed to indicate such by the oscillations

in the maximal analytic continuation of the coefficients of the voltage power series

obtained.

After presenting the derivation of the LV/large-angle formulations for both PQ

and PV buses, numerical tests on the five-, seven- and 14-bus systems are conducted

to find all the solutions of the system of nonlinear PF equations for those systems using

the proposed HE method.

After completing the derivation to find all the PF solutions using the HE method, it

is shown that the proposed HE method can be used to find only the of interest PF solutions

(i.e. type-1 PF solutions with one positive real-part eigenvalue in the Jacobian

matrix), with a proper algorithm developed. The closet unstable equilibrium point

(UEP), one of the type-1 UEP’s, can be obtained by the proposed HE method with

limited dynamic models included.

The numerical performance as well as the robustness of the proposed HE method is

investigated and presented by implementing the algorithm on the problematic cases and

large-scale power system.
ContributorsMine, Yō (Author) / Tylavsky, Daniel (Thesis advisor) / Armbruster, Dieter (Committee member) / Holbert, Keith E. (Committee member) / Sankar, Lalitha (Committee member) / Vittal, Vijay (Committee member) / Undrill, John (Committee member) / Arizona State University (Publisher)
Created2015
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Description
A nonlinear dynamic model for a passively cooled small modular reactor (SMR) is developed. The nuclear steam supply system (NSSS) model includes representations for reactor core, steam generator, pressurizer, hot leg riser and downcomer. The reactor core is modeled with the combination of: (1) neutronics, using point kinetics equations for

A nonlinear dynamic model for a passively cooled small modular reactor (SMR) is developed. The nuclear steam supply system (NSSS) model includes representations for reactor core, steam generator, pressurizer, hot leg riser and downcomer. The reactor core is modeled with the combination of: (1) neutronics, using point kinetics equations for reactor power and a single combined neutron group, and (2) thermal-hydraulics, describing the heat transfer from fuel to coolant by an overall heat transfer resistance and single-phase natural circulation. For the helical-coil once-through steam generator, a single tube depiction with time-varying boundaries and three regions, i.e., subcooled, boiling, and superheated, is adopted. The pressurizer model is developed based upon the conservation of fluid mass, volume, and energy. Hot leg riser and downcomer are treated as first-order lags. The NSSS model is incorporated with a turbine model which permits observing the power with given steam flow, pressure, and enthalpy as input. The overall nonlinear system is implemented in the Simulink dynamic environment. Simulations for typical perturbations, e.g., control rod withdrawal and increase in steam demand, are run. A detailed analysis of the results show that the steady-state values for full power are in good agreement with design data and the model is capable of predicting the dynamics of the SMR. Finally, steady-state control programs for reactor power and pressurizer pressure are also implemented and their effect on the important system variables are discussed.
ContributorsArda, Samet Egemen (Author) / Holbert, Keith E. (Thesis advisor) / Undrill, John (Committee member) / Tylavsky, Daniel (Committee member) / Karady, George G. (Committee member) / Arizona State University (Publisher)
Created2016