Abstract: A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Abstract: An electrical ground fault detection and limitation system for employment with a nuclear reactor utilizing a liquid metal coolant. Elongate electromagnetic pumps submerged within the liquid metal coolant and electrical support equipment experiencing an insulation breakdown occasion the development of electrical ground fault current. Without some form of detection and control, these currents may build to damaging power levels to expose the pump drive components to liquid metal coolant such as sodium with resultant undesirable secondary effects. Such electrical ground fault currents are detected and controlled through the employment of an isolated power input to the pumps and with the use of a ground fault control conductor providing a direct return path from the affected components to the power source.

Abstract: To qualify the heights of inwardly projecting nuclear fuel rod-positioning stops formed in the sidewall of tubular ferrules 12 utilized in fuel bundle spacers, an automated gauging apparatus, seen in FIG. 6, is provided to handle successive ferrules on an automated basis. The apparatus includes an orienting station 32 for longitudinally reorienting each ferrule, if necessary, and angularly orienting each ferrule in a pickup position. A mechanism 68 picks each ferrule from its pickup position and places it on an arbor 84 with the rod-positioning stop surfaces bearing against arbor reference surfaces. A plurality of eddy current probes 96 (FIG. 8) are positioned to measure the proximity of ferrule peripheral surface sites adjacent the stops and thus provide output gauging signals indicative of the heights of the various stops.

Abstract: An improved nuclear fuel element for service in power generating, water cooled nuclear reactors, comprising a fuel cladding container of an alloy of zirconium provided with a barrier lining of unalloyed zirconium metal metallurgically bonded to the container's inner surface, and enclosed therein fissionable nuclear fuel including an additive of aluminum silicate.

Abstract: In a post-accident sample needle system, at the sample station, an improved seal for a PASS liquid sample needle is disclosed. In such systems, a bottle with a septum is impaled on paired needles. The first needle provides venting, and the second needle provides for the placement of potentially radioactive samples in the bottle through the septum. A needle of the hypodermic type having an angled point is provided with a flared end remote from the point. A rubber O-ring fits over the needle adjacent the flare having an inner circumferential diameter that permits passage along the length of the needle and trapping at the end of the needle adjacent the flare. The O-ring is captured at the flared end of the needle by a conventional compression nut. The nut, when tightened, compresses the O-ring to circumferentially expand beyond the edges of the flare to form a seal between the needle and O-ring and to form a secondary seal between the bore and O-ring.

Abstract: Tubular parts are delivered in successive pairs to separate rotary positioners where they are held for machining in parallel by twin laser cutting beams directed thereat by a commonly mounted pair of focussing heads. The rotary positioners are commonly rotated through a programmed series of angular motions in conjunction with a programmed series of joint X and Y axes linear motions of the focusing heads to scan the cutting beam over the surfaces of the parts and thus machine identical feature cuts in the sidewall of each part. Machining is performed under an inert gas atmosphere, and cuttings are vacuumed from the machining sites through apertured mandrels inserted into the part bores.

Abstract: An apparatus and process for control rod assembly and construction for a nuclear reactor is disclosed wherein the control rod is of cruciform configuration having a plurality of and preferably four flat planar members. Each planar member is made up with an assembly of square cross sectioned tubular members. Each square cross sectioned tubular member includes a defined cylindrical center and a constant thickness surrounding tubular metallic shell to provide the contained neutron absorbing materials in a cylindrically shaped pressure vessel. The constant thickness surrounding tubular shell constituting the cylindrical shaped pressure vessel has four corner sections added to generate a modified square exterior profile. These added four corner sections comprise an addition to the tubular side wall thickness of the shell. Appropriate chamfers and rounding are provided to the square profile to impart both stress relief and convenient points for automated welding between side-by-side square sections.

Abstract: An apparatus and process for the construction of a cruciformed section control rod is disclosed. The control rod is characterized by having square exterior sectioned tubes with central cylindrical volumes for the containment of neutron absorbing poisons. In the process, these tubes are welded together at the corners of the square sections to form the planar members of the cruciformed shaped rods. An apparatus and process is disclosed in which the rods are all held by a jig in the same configuration as is ultimately fabricated. Thereafter, the jig is drawn through fixtures, which fixtures accurately guide the rods to be joined by welding past paired and opposed laser welding heads. By the expedient of positioning the laser and rapidly drawing the braced together tubes by the laser, control rod fabrication occurs. There results a solid, integral control rod constituting a suitable pressure vessel for neutron absorbers for control of nuclear reactions.

Abstract: To form longitudinally spaced pairs of angularly spaced stops in the walls of fuel rod-positioning ferrules utilized in nuclear fuel bundle spacers, the ferrules, cut from tubular stock, are longitudinal oriented in a first station and angularly oriented in a second station using a reference notch cut in one ferrule edge. Successive ferrules are picked from the second station and placed in a die of a punch press. Ferrule exterior surface back-up is provided by the die upon closure, and interior surface back-up is provided by a mandrel inserted in the ferrule bore as the punches converge into stop-forming engagement with the ferrule wall. Formed ferrule discharge is induced by placement of the next ferrule in the die.

Abstract: To assemble nuclear fuel bundle spacers comprised of a multiplicity of rod-positioning ferrules, automated apparatus is provided comprising an indexing table mounting a plurality of identical, angularly spaced ferrule fixtures which are successively positioned at a first station for receipt of a pair of properly oriented ferrules. The table is indexed to present the ferrule pairs in turn to a second station where a spring is loaded in pre-assembly relation thereto. At the next table index position, a third station manipulates the successive fixture-mounted ferrule pairs to interlock the spring with each ferrule pair as a subasssembly which is then indexed in turn to a fourth station for assembly into a spacer fixture.

Abstract: A spacer assembly for imparting a swirl to a passing steam water mixture within a fuel channel of a boiling water reactor is disclosed. The spacer assembly is placed at periodic intervals along the inside of a square cross sectioned fuel channel so as to separate the fuel containing rods from each other and the fuel channel sides. Each discrete spacer element is constructed of a matrix of helically twisted sheet metal spacing elements for the serendipitous purposes of providing separation of the steam and water, with water being deposited on the fuel rods surfaces. Each helical element is a twisted planar metallic member and includes a rotation of the metal surfaces in the order of 450.degree. over a typical two inches of length. For each 90.degree. interval of rotation from the top and bottom of each helical element, paired lattice connections are provided for attachment to adjoining elements. Intermediate of the lattice connectors at 45.degree.

Abstract: A steam separating system for boiling water nuclear reactors comprising a combination of mechanical steam separators with steam collecting and transferring units.

Abstract: An intermediate heat transport system is disclosed for providing steam generation from the secondary nonradioactive liquid sodium heat extraction loop from a sodium cooled nuclear reactor. The transport system includes a unitary module combining the steam generating heat exchanger, the pump for the circulation of the liquid sodium coolant, and the surge volume required for differential expansion between the sodium and the vessels that contain the sodium. Two concentric cylindrical and vertically standing vessels are provided for containing the liquid sodium; one vessel is outer and larger; the other inner vessel is of a double wall construction and open on its lower end hung from the top of the longer vessel. The outer and larger cylindrical vessel has four feedwater inlet plenums (typical number) at the bottom and four steam outlet plenums at the top. Tube sheets terminate each plenum to a tube bundle extending between the inlet and outlet plenums.

Abstract: The control rod drive is removed from the top of the reactor. The control rod drive is attached at its upper end to the bottom end of the control rod drive guide tube. The control rod guide tube in turn couples at a rotationally locked bayonet fitting to the top of the control rod drive housing. Locking of the bayonet fitting occurs thru the orificed fuel support casting. The orificed fuel support casting, is locked from rotation at a pin located on the reactor core plate. The control rod guide tube in turn locks rotationally to the orificed fuel support casting. During refueling, the fuel bundles and the orificed fuel support casting are removed, freeing the control rod guide tube and its depending drive for rotation. Thereafter, a fitting captures the control rod guide tubes at coolant inflow apertures and rotates the guide tubes and the depending drives.

Abstract: Both gadolinia nuclear fuel rods and urania fuel rods are non-destructively tested to determine the enrichment concentration of the fuel pellets therein by serially conveying the gadolinia rods and the urania rod along separate test channels through an active scanner where they are irradiated from a common neutron source, and the neutron-induced gamma ray emissions are counted as a measure of the enrichment concentrations at a multiplicity of points distributed along the rod length. The enrichment counts are corrected for background radiation, variations in fuel density, and, in the case of gadolinia rods, for gadolinia content by passing these rods through a magnetometer.

Abstract: An apparatus and process for control rod assembly and construction for a nuclear reactor is disclosed wherein the control rod is of cruciform configuration having a plurality of and preferably four flat planar members. Each planar member is made up with an assembly of square cross sectioned tubular members. Each square cross sectioned tubular member includes a defined cylindrical center and a constant thickness surrounding tubular metallic shell to provide the contained neutron absorbing materials in a cylindrically shaped pressure vessel. The constant thickness surrounding tubular shell constituting the cylindrical shaped pressure vessel has four corner sections added to generate a modified square exterior profile. These added four corner sections comprise an addition to the tubular side wall thickness of the shell. Appropriate chamfers and rounding are provided to the square profile to impart both stress relief and convenient points for automated welding between side-by-side square sections.

Abstract: Nuclear fuel elements for use in the core of a nuclear reactor include an improved composite cladding having a zirconium barrier layer metallurgically bonded on the inside surface of a zirconium alloy tube, wherein the inside surface of the barrier is alloyed with preselected elemental impurities to improve oxidation resistance. The zirconium barrier layer forms a shield between the zirconium alloy tube and a core of nuclear fuel material enclosed in the composite cladding. The alloy layer formed on the barrier surface acts to inhibit cracking during the tube production fabrication step and limits oxidation in the event that the cladding is breached during operation of the reactor, allowing the entrance of water or steam into the fuel element.

Abstract: A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

Abstract: In a nuclear boiling water reactor, an improved lower tie-plate and fuel channel interface for a boiling water reactor fuel bundle is disclosed. The fuel bundle has a lower tie-plate for supporting fuel rods and permitting the introduction of fluid interior of the fuel bundle. An upper tie-plate maintains the lower tie-plate supported rods in side-by-side relation and has apertures for discharging a mixture of water and steam. The fuel rods extend between the tie-plates for the generation of steam with some of the fuel rods forming a threaded connection fastening the tie-plates together. A polygon sectioned channel, preferably square, surrounds the tie-plates and fuel rods for the confining of fluid flow between the tie-plates interior of the bundle. The interface of the channel as it surrounds the lower tie-plate is reconfigured.

Abstract: In a boiling water reactor, provision is made to sample the core bypass region immediate the top guide to determine the physical and chemical constituents of the moderating water. A conduit for a local power range monitor is fitted with a measurement assembly. The conduit and measurement assembly are inserted up to the vicinity of the top guide. A tube opening is provided to the bypass region immediate the top guide. During reactor operation, the saturated liquid in this region flashes to a steam water mixture (18% steam) at constant enthalpy and is rapidly removed from the reactor to measuring equipment in the reactor building. During removal, the radiolytic disassociated gases (namely hydrogen and oxygen) partition to the steam phase where their recombination is retarded and accurate measurement of their constituent content can be made. Also, temperature of the steam water mixture decreases thereby preserving unstable species like hydrogen peroxide that degrade much more rapidly a high temperatures.