[en] The codes and standards applicable to plant life extension have not been identified in the US at this time. However, there are several initiatives underway to establish the specific codes and standards pertaining to nuclear plant life extension (PLEX). The Board of Nuclear Codes and Standards (BNCS), the American Society of Mechanical Engineers (ASME), the American Society for Testing of Materials (ASTM) and the Institute of Electrical and Electronics Engineers (IEEE) have all formed groups to formulate codes and standards for mechanical and electrical components pertaining to PLEX. The potential need for codification to address PLEX is based on the following: current design codes generally do not address inservice degradation processes; there are currently no codification standards for many nuclear components, for example, turbines, diesel generators; there are no code requirements on obtaining the operating records important to PLEX; and there are currently no codes or standards pertaining to critical plant structures, for example, spent fuel pool, condenser intake structure. This paper will summarize the activities of each of these code writing bodies and give a status report and background information on any proposed code recommendations to support PLEX

[en] A nuclear reactor is described which includes: a reactor vessel having a plurality of nozzles, a thick concrete primary shield surrounding the reactor vessel, a reactor cavity comprising a small spacing provided between the reactor vessel and the concrete primary shield; passages traversing the concrete primary shield, and a plurality of liquid-coolant-carrying pipes positioned within the passages in the concrete primary shield, and the liquid-coolant-carrying pipes welded to the nozzles of the reactor vessel, the passages in the concrete primary shield having an oblong configuration, and motion-limiting keys provided within each of the passages to limit motion of the liquid-coolant-carrying pipes therein; the liquid-coolant-carrying pipes each having a carbon steel nozzle portion welded under shop conditions to a stainless steel ring portion in order to preclude the necessity of performing bimetallic welds in the field when joining the liquid-coolant-carrying pipes to the nozzles of the reactor vessel, and the bimetallic welds constituting a location for possible pipe rupture and a resulting loss of coolant fluid

[en] A reactor pressure vessel disposed in a cavity has coolant inlet or outlet pipes extending through passages in the cavity walls and welded to pressure nozzles. The cavity wall has means for directing fluid away from a break at a weld away from the pressure vessel, and means for inhibiting flow of fluid toward the vessel. (author)

No abstract available

[en] The invention relates to an installation for transferring fuel loads between a storage building and the enclosure of a nuclear reactor. The fuel rods are transported in a basket supported by a carriage horizontally moving along a track through an opening formed in the retaining wall. The basket is pivotally mounted on the carriage about an axis passing through the counter of gravity thereof and a space is reserved under the carriage for allowing the basket to point. Extra features of operation are provided. This installation permits to load and unload fuel loads most safely

[en] A fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area is described. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track which serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. (author)

[en] An emergency core cooling system is described which discharges neutron absorber material, such as borated water, from a separate system of accumulators into the area beneath the reactor head and above the upper support plate structure for further distribution through coolant conducting devices to the top of the reactor core

[en] Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

[en] The emergency cooling system for a nuclear-reactor comprises a pressurized enclosure tightly sealed by an obturating head forming a pressure chamber, a reactor-core in said enclosure, an inlet and an outlet for allowing the coolant to flow, characterized in that it comprises an accumulator containing a pressurized neutron-absorbent connected to the obturating head for providing said pressure-chamber with an emergency-coolant when the coolant pressure within the reactor falls below the pressure within the accumulator, and means in communication with the pressure-chamber and the core for introducing the emergency coolant into the latter

[en] An emergency core cooling system for a nuclear reactor which preferably is supplemental to the main emergency core cooling system incorporated in the reactor at the time of construction is described. Under circumstances of a rupture in the reactor primary coolant piping and consequent drop in reactor coolant pressure, emergency supplemental coolant is supplied from tanks or accumulators through check valves into the head closure plenum area. From there, the coolant is distributed downwardly through hollow support columns and through control rod guide thimbles to the top of the fuel assemblies which comprise the reactor core. The pressure and flow of the emergency supplemental coolant is sufficiently great to overcome the normal upward flow of primary coolant through the core, the result being that the supplemental coolant causes collapse of bubbles otherwise generated by the heat producing fuel rods thereby permitting the supplemental coolant to effectively and efficiently carry away heat generated in the core