Nuclear Engineering ETDs

Amy Drumm

Publication Date

1-28-2015

Abstract

The purpose of this work was to develop and validate a design tool for the transients that are of interest in designing the decay heat removal system for a Fluoride salt-cooled High-temperature Reactor (FHR). The loop of interest operates with a single phase, near ambient pressure molten salt coolant. Most of the codes used for LWR safety and licensing were designed to model more complex heat transfer problems. A single phase, one dimensional, natural circulation code is more suited to validating with simple university experiments. A code designed for lead bismuth eutectic at the Bhabha Atomic Research Centre (BARC) was modified to use it to model the flow in the decay heat removal system in an FHR. The discretization of the governing equations was modified to account for non-uniform mesh spacing and to allow for an explicit (forward Euler), semi-implicit (Crank Nicolson), or fully implicit (backward Euler) solution. The models used to approximate the density integral were also modified to account for a variable mesh scheme. The code was benchmarked against the original code developed at BARC, tested for parameter sensitivities, and then validated with experimental data. The most sensitive areas in the code were tested with a simple validation experiment. The experimental data was compared to the computational results to determine how accurately the methods used in the code can model the physical process involved in the transients being studied. The code was found to model the overall flow and temperature profiles fairly well for the startup, heater step, and loss of heat sink transients involved in an FHR. In general, the code overestimated the flow rate and underestimated the temperatures. Inaccurate modeling of the loss coefficients and parasitic heat loss could be the largest contributing factors these effects.

Keywords

Numerical Methods, Fluoride salt-cooled High-temperature Reactor (FHR), Validation Study, Thermal Hydraulics

Document Type

Thesis

Language

English

Degree Name

Nuclear Engineering

Level of Degree

Masters

Department Name

Nuclear Engineering

First Committee Member (Chair)

Prinja, Anil

Second Committee Member

Mousseau, Vincent

Third Committee Member

Blandford, Edward

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