Abstract: A system for removing sludge from a sludge-bearing fluid flowing in a vertically oriented nuclear steam generator so that the sludge is not present therein in sufficient quantities to corrode components (e.g., heat transfer tubes) located in the steam generator. The system comprises a tube sheet having a plurality of apertures therethrough for receiving the ends of the heat transfer tubes so that the tubes are supported by the tube sheet. The tube sheet has a top surface thereon and also a recess formed in the top surface for separating and for collecting the sludge from the sludge-bearing fluid. The recess has a substantially frusto-conical wall generally converging at the bottom end of the recess for providing a settling basin for the sludge after the sludge is separated from the sludge-bearing fluid. A pipe extends from adjacent the bottom end of the recess to the exterior of the heat exchanger for transporting the sludge from the recess to the exterior of the heat exchanger.

Abstract: A hydro-ball in-core instrumentation system employs detector strings each comprising a wire having radiation sensitive balls affixed diametrically at spaced positions therealong and opposite tip ends of which are transportable by fluid drag through interior passageways through which primary coolant is caused to flow selectively in first and second opposite directions for transporting the detector strings from stored positions in an exterior chamber to inserted positions within the instrumentation thimbles of the fuel rod assemblies of a pressure vessel, and for return. The coolant pressure within the detector passageways is the same as that within the vessel; face contact, disconnectable joints between sections of the interior passageways within the vessel facilitate assembly and disassembly of the vessel for refueling and routine maintenance operations.

Abstract: A hydro-ball in-core instrumentation system employs detector strings each comprising a wire having radiation sensitive balls affixed diametrically at spaced positions therealong and opposite tip ends of which are transportable by fluid drag through interior passageways. In the passageways primary coolant is caused to flow selectively in first and second opposite directions for transporting the detector strings from stored positions in an exterior chamber to inserted positions within the instrumentation thimbles of the fuel rod assemblies of a pressure vessel, and for return. The coolant pressure within the detector passageways is the same as that within the vessel; face contact, disconnectable joints between sections of the interior passageways within the vessel facilitate assembly and disassembly of the vessel for refueling and routine maintenance operations.

Abstract: A method of providing a nuclear steam supply system that is constructed to include a barge or flotation base as an integrated portion thereof. The system is tested for operability and safety at the factory and then towed along navigable coastal or inland waterways, or overseas to foreign locations, to a prepared foundation site at its point of use. The overall unit is so constructed and arranged as to permit the removal of top and side portions to permit passage under low bridges and through narrow locks without requiring disassembly of safety class piping and wiring. The method further provides for a complete nuclear power plant constructed on multiple barges at a separate factory site in parallel with plant site preparation, i.e., in estuaries or caves.

Abstract: A method and apparatus for extending the life of a core of a water moderated nuclear reactor by spectral shift comprising the displacement of a portion of the moderator in the core by a gas having a low neutron cross-section during the initial stage of the life of the core and displacing the gas and replacing it with moderator during the latter part of the life of the core.

Abstract: A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core.

Abstract: The reactor and its containment, instead of being supported on a solid concrete pad, are supported on a truss formed of upper and lower reinforced horizontal plates and vertical walls integrated into a rigid structure. The plates and walls from chambers within which the auxiliary components of the reactor, such as valves, pumping equipment and various tanks, are disposed. Certain of the chambers are also access passages for personnel, pipe chases, valve chambers and the like. In particular the truss includes an annular chamber. This chamber is lined and sealed by a corrosion-resistant liner and contains coolant and serves as a refueling cooling storage tank. This tank is directly below the primary-coolant conductor loops which extend from the reactor above the upper plate. The upper plate includes a sump connected to the tank through which coolant flows into the tank in the event of the occurrence of a loss-of-coolant accident.