Nuclear science in engineering education : a selected list of references for instructors /

Includes bibliographical references.

In-pile loop test for the pebble bed reactor fuel development program : final report /

"Final report to Sanderson and Porter for subcontract no. S-3 under USAEC Prime Contract no. AT(30-1)-2378."

The effect of nuclear radiation on ceramic reactor-fuel materials /

"June 30, 1963."

The effect of nuclear radiation on metallic fuel materials /

"September 30, 1963."

International nuclear spent fuel storage programs : hearing before the Subcommittees on International Security and Scientific Affairs and on International Economic Policy and Trade of the Committee on Foreign Affairs, House of Representatives, Ninety-sixth Congress, first session, May 3, 1979.

Mode of access: Internet.

The Illinois Spent Nuclear Fuel and High-Level Waste Inspection and Escort Program /

"5/99"--Colophon.

Proceedings of the U.S. Nuclear Regulatory Commission Tenth Water Reactor Safety Research Information Meeting, held at National Bureau of Standards, Gaithersburg, Maryland, October 12-15, 1982 /

Includes papers describing research sponsored by the Office of Nuclear Regulatory Research, NRC.

Proceedings of the U.S. Nuclear Regulatory Commission Eleventh Water Reactor Safety Research Information Meeting, held at National Bureau of Standards, Gaithersburg, Maryland, October 24-28, 1983 /

Includes bibliographies.

Nuclear reactor control engineering.

Includes bibliography.

Principles of nuclear reactor engineering,

Bibliographical footnotes.

Illinois spent nuclear fuel transportation.

Mode of access: Internet.

Reactor structural materials: engineering properties as affected by nuclear reactor service,

Includes bibliography.

Chemical reactions at nuclear fuel-clad interfaces

This thesis presents a study of the chemical reactions that may occur at the fuel- clad interfaces of fuel elements used in advanced gas-coooled reactors (A.G.R.) The initial investigation involved a study of the inner surfaces of irradiated stainless steel clad and evidence was obtained to show that fission products, in particular tellerium, were associated with reaction products on these surfaces. An accelerated rate of oxidation was observed on the inner surfaces of a failed A.G.R. fuel pin. It is believed that fission product caesium was responsible for this enhancement. A fundamental study of the reaction between 20%Cr/25%Ni/niobium stabilised stainless steel and tellerium was then undertaken over the range 350 - 850 degrees C. Reaction occurred with increasing rapidity over this range and long term exposure at ≤ 750 degrees resulted in intergranular attack of the stainless steel and chromium depletion. The reaction on unoxidised steel surfaces involved the formation of an initial iron-nickel-tellerium layer which subsequently transformed to a chromium telluride product during continued exposure. The thermodynamic stabilities of the steel tellurides were determined to be chromium telluride > nickel telluride > iron telluride. Oxidation of the stainless steel surface prior to tellerium exposure inhibited the reaction. However reaction did occur in regions where the oxide layer had either cracked or spalled.

TiN-based coatings on fuel cladding tubes for advanced nuclear reactors

Titanium nitride (TiN) thin films are coated on HT-9 and MA957 fuel cladding tubes and bars to explore their mechanical strength, thermal stability, diffusion barrier properties, and thermal conductivity properties. The ultimate goal is to implement TiN as an effective diffusion barrier to prevent the inter-diffusion between the nuclear fuel and the cladding material, and thus lead to a longer lifetime of the cladding tubes. Mechanical tests including hardness and scratch tests for the samples before and after thermal cycle tests show that the films have a high hardness of 28GPa and excellent adhesion properties despite the thermal treatment. Thermal conductivity measurements demonstrate that the thin TiN films have very minimal impact on the overall thermal conductivity of the MA957 and HT-9 substrates, i.e., the thermal conductivity of the uncoated HT-9 and MA957 substrates was 26.25 and 28.44 W m-1 K-1, and that of the coated ones was 26.21 and 28.38W m-1 K-1, respectively. A preliminary Ce diffusion test on the couple of Ce/TiN/HT-9 suggests that TiN has excellent material compatibility and good diffusion barrier properties.