Abstract: An apparatus for repairing a shroud in which one or more seam welds are cracked. The repair involves the attachment of a splice bracket to the shroud so that the bracket bridges the cracked weld seam and is applicable to both vertical and horizontal weld seams. The bracket is intended to structurally replace the shroud seam weld which is cracked. Multiple splice brackets can be placed along the length of a crack. Each splice bracket is attached to the shroud by a plurality of tapered fastener assemblies.

Abstract: A method for mitigating initiation or propagation of a crack in a surface of a metal component in a boiling water reactor. The method includes the step of injecting a solution or suspension of a pH-adjusting compound into the bulk water of the reactor. The compound has the property of changing the pH of high-temperature water inside the crack from a value outside a predetermined pH range (namely, pH 6.0 to 8.0) to a value within the predetermined pH range without causing any significant change of the bulk water pH. The growth rate of the crack when the crack pH is outside the predetermined pH range is greater than the growth rate of the crack when the crack pH is within the predetermined pH range.

Abstract: A lifting assembly is configured to facilitate attaching of the lifting bar to a load bearing water rod in a nuclear fuel assembly. A connecting member is rigidly secured to the water rod and is provided with one of a male engaging portion and a female receiving portion. A lifting member including the lifting bar is engageable with the connecting member and includes the other of the male engaging portion and the female receiving portion. The male engaging portion and the female receiving portion are correspondingly shaped, and after insertion of the connecting member into or about the lifting member, the connecting member and lifting bar can be rotated relative to each other to prevent disassembly of the lifting assembly. A spring is disposed between the lifting bar and the water rod and urges the male engaging portion toward a locked position. The lifting assembly facilitates the attaching and detaching of the lifting bar to a load bearing water rod.

Abstract: A storage cage for the storage and transport of a plurality of fuel assemblies includes a baseplate having apertures formed therein. A plurality of casings each have an interior and each stand on the baseplate for receiving and storing a fuel assembly. Each of the casings is associated with at least one of the apertures leading into the interior of the casing for supplying and discharging cooling liquid. Each of the casings has a plurality of slots formed therein for supplying and discharging cooling gas. At least one supporting wall forms a load-bearing part with the baseplate and the casings are fastened to the at least one supporting wall.

Abstract: A deflector is disposed in a steam vent region above the upper ends of part-length fuel rods to deflect upwardly flowing liquid laterally outwardly into the interstices of the full-length fuel rods adjacent the steam vent region. The deflector is supported by a spacer or from the upper tie plate and can be removed from the fuel bundle for access to the part-length fuel rods. The deflector may comprise a flat plate, an inverted pyramid, an inverted cone, a multi-sided pyramidal configuration or a swirl device wherein the horizontally projected area of the deflector is substantially coextensive with the horizontal cross-sectional area of the steam vent area. In this manner, a higher density liquid is provided in the interstices of the full-length rods while the lower density steam flows into the steam vent volume for flow upwardly out of the fuel bundle.

Abstract: A passive emergency water system for cooling the atmosphere inside containment in a water-cooled nuclear reactor comprises an elevated water reservoir connected to an in-containment heat exchanger. The heat exchanger promotes natural convection of containment atmosphere and heat transfer. Heat from containment atmosphere is transferred to the reservoir by the in-containment heat exchanger via a convective return flowpath. The heat exchanger is preferably an elevated tube bank and the reservoir is preferably external to containment. Baffle walls can be used inside containment to promote circulation of containment atmosphere.

Abstract: A filter for a control rod drive of a nuclear reactor is adapted to be attached to a connector extending upwardly from a piston. The filter includes a base portion and a perforated filter portion, the base portion having a radially open recess adapted to receive the connector in a sideways movement. The recess is bordered by a pair of spring arms of substantially square cross-section which are adapted to resiliently grip the connector upon installation.

Abstract: A depressurizing system (1) for depressurizing plants, operating with pressurized steam and including a steam head (16), by injection of cold water under gravity from a reservoir (2) located at a higher level than the pressurized plant and connected thereto by a delivery duct (24) which forms a syphon (24a, 24b, 24c) and which is connected to the narrow section (10) of an ejector (6) into which steam flows in the event of an incident, the steam being drawn from a condenser (7) located downstream of the ejector (6), causing a drop in pressure which draws cold water from the tank (2) through the syphon which is thus overcome, enabling cold water to be injected into the pressurized plant.

Abstract: The tube to be replaced is extracted and an internal surface of the bottom head (1b) is then machined around the through opening (10), a hollow shell (20) is attached and fixed on the machined surface (15) of the bottom head (1b), around the through opening (10), by welding along a peripheral edge (19), the hollow shell is bored in order to produce a passage hole (24) through its wall, in extension of the through opening (10) and a welding bevel (23). A replacement through-tube (18) is fixed on the wall of the hollow shell (20) by depositing a welding metal (25) in the bevel (23). The invention applies in particular to the replacement of vessel bottom head penetrations of a pressurized water nuclear reactor.

Abstract: A dry transfer system for spent nuclear fuel (SNF) assemblies includes a transfer container with a sliding sleeve for vertical translation therein. The sliding sleeve includes a number of compartments for receiving a corresponding number of fuel assemblies. The container includes an integral and remotely controllable hoist with a number of individually actuated grapples for latching onto a corresponding number of fuel assemblies. The system further includes a loading stand with an elevator for raising and lowering a fuel basket that also includes a number of compartments for moving fuel assemblies up to and in alignment with the sleeve of the transfer container that is landed on the loading stand.

Abstract: In a pressure vessel of a nuclear reactor containing a core assembly enclosed within a core shroud, the core shroud spaced radially inwardly of a side wall of the pressure vessel with an annular pump deck located in an annular radial space between the core shroud and the side wall of the pressure vessel, the improvement wherein the shroud is removably secured to an annular support leg extending upwardly from the bottom of the pressure vessel; and further wherein the annular pump deck is provided in the form of a plurality of removable segments.

Abstract: A device for collecting reactor-meltdown products from a reactor pressure vessel includes an expansion chamber for receiving the reactor-meltdown products and for receiving a coolant, such as cooling water, for cooling the reactor-meltdown products spreading in the expansion chamber. The expansion chamber is disposed laterally of the reactor pressure vessel and has a floor and a cooling system in the floor.

Abstract: A reactor container is composed of a dry well and a wet well and, in the reactor container, there is provided with a dry well cooling system, the dry well cooling system including, in one aspect, an in-dry-well heat exchanger disposed in the dry well of the reactor container, an in-dry-well blower connected to a primary side of the in-dry-well heat exchanger, a circulation pipe connected to a secondary side of the in-dry-well heat exchanger, a normal cooling system connected to the secondary side of the in-dry-well heat exchanger through the circulation pipe and including an equipment cooling pump, an equipment cooling heat exchanger and a seawater pump, and a standby cooling system connected to and branched from the circulation pipe.

Abstract: A neutron-irradiated structural material having a defect in a nuclear reactor internals is repaired by covering the structural material to be repaired with a plate to cover over a portion thereof having the defect, and then welding the plate and the structural material by locally applying pressure on the surface of the plate and adding energy to the portion to which the pressure is applied thereby to generate thermal energy in the contact surfaces between the plate and the structural material.

Abstract: The invention relates to a fuel element for a pressurized-water reactor comprising a number of fuel rods arranged in a square lattice and wherein certain positions in the lattice contain guide tubes for receiving control rods, wherein the fuel element can be used for higher burnup by manufacturing the guide tubes of the element such that their irradiation-induced growth is reduced and their corrosion resistance is increased compared with conventionally manufactured guide tubes. This is achieved by finally heat-treating the guide tubes in two steps, the first step being a beta quenching and the second step a heat treatment in the alpha-phase region.

Abstract: The reactor vessel is placed inside a reactor pit having a vertical axis and including a bottom (5) arranged below and aligned with the vessel bottom head. In the event of meltdown of the reactor core as a result of accidental operation of the nuclear reactor, which may lead to perforation of the vessel bottom head, the temperature at a plurality of points (14) distributed over the surface (8) of the reactor pit bottom (5) in line with the vessel bottom head is measured continuously. Optical measurement of the temperatures on the bottom (5) of the reactor pit is preferably carried out by using a plurality of optical fibers (10) arranged on the reactor pit bottom (5), along the length of which Bragg gratings (14), each capable of reflecting a light signal having a particular wavelength, are distributed.

Abstract: A basket for a cask for transporting, storing, and containing pressurized water nuclear fuel assemblies, including an internal assembly of sleeves has a plurality of sleeves arranged in a uniform pattern and secured within a cylindrical shell. Each of the plurality of independent sleeves being sized to secure and contain a fuel assembly. The internal assembly of sleeves includes a set of single sleeves centrally positioned in the basket, a set of corner sleeves including a single sleeve in each corner of the basket, and a set of double sleeves. An internal support structure is positioned within the basket and includes a cross-shaped central support element and four angle shaped corner supports. Neutron poison material for absorbing neutrons is secured to an inner wall of each of the plurality of sleeves for maintaining fission reactions within the basket below a critical level necessary to sustain a fission reaction.

Abstract: The method serves to manufacture tubes for constituting sheaths for nuclear fuel rods. A bar is made out of a zirconium-based alloy containing 50 ppm to 250 ppm iron, 0.8% to 1.3% by weight niobium, less than 1600 ppm oxygen, less than 200 ppm carbon, and less than 120 ppm silicon. The bar is heated to a temperature in the range 1000.degree. C. to 1200.degree. C. and is quenched in water. A blank is extruded after heating to a temperature in the range 600.degree. C. to 800.degree. C. and cold-rolled in at least four passes in order to obtain a tube, with intermediate heat treatment being performed between passes at temperatures in the range 560.degree. C. to 620.degree. C. A final heat treatment is performed at a temperature in the range 560.degree. C. to 620.degree. C., all of the heat treatments being performed under an inert atmosphere or a vacuum.

Abstract: A method and an apparatus for repairing a core plate having a cracked support structure. The method entails the step of providing a vertical compression load path between the core plate and one or more control rod guide tubes. This is accomplished by installing a plurality of load-bearing elements between the core plate and the flanges of the guide tubes. Each load-bearing element is placed between the core plate and a guide tube flange so that core plate pressure loads are reacted by the weight of the fuel rather than the core plate support structure. The load-bearing element may take the form of a solid ring, a C-shaped spring collar or a hinged collar.

Abstract: The invention relates to a process for the treatment of natural apatites with a view to the use thereof for storing irradiated fuel elements.This process consists of contacting a natural apatite powder with an acid aqueous solution, e.g. dilute HNO.sub.3 or HCl for between a few seconds and 10 minutes.In this way, the physicochemical properties of the natural apatite are improved by modifying its surface composition in order to make it close to that of hydroxyapatite. This modified apatite can be used as a filling material in irradiated fuel element storage containers.