Abstract: A nuclear reactor annealing apparatus and method for annealing a vessel of the reactor by utilizing a non-aqueous fluid, a means for replacing water in the vessel with the non-aqueous fluid, a means for heating the non-aqueous fluid to a predetermined temperature and for a predetermined length of time. The non-aqueous fluid creates less pressure than water at the same temperature and is in communication with said vessel so that heating the non-aqueous fluid in turn heats the vessel thereby annealing the vessel. A can device is used to anneal only a portion of the vessel walls.

Abstract: A method of in-service ultrasonic detection of intergranular stress corrosion cracking conditions in pipe sections of nuclear power plants. A plurality of ultrasonic return signal features and an associated general form of algorithmic combination are selected on the basis of ultrasonic pipe studies to use in distinguishing return signals from cracking conditions and geometric reflectors. A calibration pipe sample having a prearranged configuration with geometric reflectors substantially corresponding to the unknown pipe section and at least one known cracking condition is provided.

Abstract: A technique for storing spent fuel rods from a nuclear reactor is disclosed herein. This technique utilizes a housing including a closed inner chamber for containing the fuel rods and a thermally conductive member located partially within the housing chamber and partially outside the housing for transferring heat generated by the fuel rods from the chamber to the ambient surroundings. Particulate material is located within the chamber and surrounds the fuel rods contained therein. This material is selected to serve as a heat transfer media between the contained cells and the heat transferring member and, at the same time, stand ready to fuse into a solid mass around the contained cells if the heat transferring member malfunctions or otherwise fails to transfer the generated heat out of the housing chamber in a predetermined way.