Masters Theses

Date of Award

12-1981

Degree Type

Thesis

Degree Name

Master of Science

Major

Nuclear Engineering

Major Professor

Hall C. Roland

Committee Members

P. F. Pasqua, L. F. Miller

Abstract

Graphite, due to its excellent nuclear, physical, and mechanical properties, is widely used in nuclear reactors; specially, it is utilized effectively, and in large quantities, as neutron moderator, reflector, and supporting structure in high-temperature gas-cooled reactors (HTGR's). This large application of graphite has justified extensive studies of its properties under simulated reactor core conditions.

One of the critical properties of nuclear graphite is its creep in a neutron-irradiation environment, referred to as "irradiation-induced creep". Knowledge of this creep and its dependence on neutron fluence, temperature, and stress is of paramount importance to the reactor designer to insure the structural integrity and dimensional stability of the graphite core. This knowledge can only be obtained by conducting rather complex and costly irradiation experiments.

It is the aim of this work to present the thermal design and analysis of a new graphite irradiation-induced creep test capsule, designated the OC-6 capsule. This capsule is cylindrical in shape and contains two cylindrical graphite specimen columns, one of which will be kept under constant compressive stress during the course of the irradiation experiment. The experiment is to be performed in the Oak Ridge Research Reactor (ORR), where the capsule will be placed at the E-5 position of the core for a total period of several reactor power-cycles; the post-irradiation studies will be done after its removal from the core.

The main concern in the thermal design of the capsule is to make it capable of maintaining a constant temperature of 1200 C on the surface of the specimens, while not allowing that of the surrounding graphite block, at thermocouple locations, to exceed 950 C. Gamma heating will be the only heat source in the capsule. A modified version of the HEATING5 computer code was utilized to perform the heat transfer calculations for a computational model of the capsule in three-dimensional cylindrical coordinates.

In order to achieve the required high temperature, it was found that a thermal radiation reflector was required around the specimen columns. Tungsten was chosen for this reflector. Each tungsten sleeve is separated from the corresponding specimen column and the graphite block (surrounding both) by two annular gas gaps, referred to as the Inner Gap (IG) and Middle Gap (MG), respectively. A third annular gas gap, referred to as the Outer Gap (OG), separates the graphite block from the stainless-steel containment of the capsule which is in contact with the reactor coolant.

Because of the non-flat axial distribution of the gamma-heating rate, and due to the flat temperature distribution required for the specimens, the width of the MG and OG must be changed, stepwise, in the axial direction; and a proper mixture of helium and neon must be used, as the sweep gas, to flow through the gaps. For a fixed power profile, the widths of the gaps and the conductivity of the gas mixture are the only two available tools in controlling the distribution and the level of the temperature within the specimens.

The main goal of the thermal design has been achieved by being able to maintain about half of the length of each specimen column (30 cm) at a rather flat temperature of 1200 C. The required gap widths, and their lengths, have been calculated. The overall level of temperature control has also been calculated for various He-Ne gas mixtures in the sweep gas.

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