A pressurized water reactor (PWR) includes a vertical cylindrical pressure vessel having a lower portion containing a nuclear reactor core and a vessel head defining an integral pressurizer. A reactor coolant pump (RCP) mounted on the vessel head includes an impeller inside the pressure vessel, a pump motor outside the pressure vessel, and a vertical drive shaft connecting the motor and impeller. The drive shaft does not pass through the integral pressurizer. The drive shaft passes through a vessel penetration of the pressure vessel that is at least large enough for the impeller to pass through.

An advanced initial core fuel configuration is for improving the fuel management efficiency and thus economics for a nuclear reactor. The advanced initial core fuel configuration includes a plurality of fuel assemblies having different average enrichments of uranium 235 and arranging the fuel assemblies in an initial core configuration structured to emulate a known equilibrium reload cycle core at least in terms of spatial reactivity distribution. The resulting average enrichment within the initial core ranges from below about 1.0 percent weight of uranium 235 to about 5.0 percent weight of uranium 235. An advanced lattice design is also disclosed.

A suspended basket includes a plurality of plates, tie rods, and adjustable length threaded tie rod couplings connecting threaded ends of the tie rods with threaded features of the plates. Control rod drive mechanisms (CRDMs) with CRDM motors are mounted in the suspended basket, which is suspended in a pressure vessel above a nuclear reactor core to control insertion of control rods into the reactor core. In one embodiment each adjustable length threaded tie rod coupling is a turnbuckle coupling that includes a sleeve threaded onto the threaded end of the tie rod and onto the threaded feature of the plate, and the sleeve is rotatable to adjust the position of the tie rod respective to the plate. Guide frames may be mounted in the suspended basket between the CRDMs and the nuclear reactor core to guide portions of the control rods withdrawn from the nuclear reactor core.

Thorium-based fuel bundles according to one or more embodiments of the present invention are used in existing PHWR reactors (e.g., Indian 220 MWe PHWR, Indian 540 MWe PHWR, Indian 700 MWe PHWR, CANDU 300/600/900) in place of conventional uranium-based fuel bundles, with little or no modifications to the reactor. The fuel composition of such bundles is 60+ wt % thorium, with the balance of fuel provided by low-enriched uranium (LEU), which has been enriched to, according to one or more embodiments, 13-19.95% .sup.235U. According to various embodiments, the use of such thorium-based fuel bundles provides (1) 100% of the nominal power over the entire life cycle of the core, (2) high burnup, and (3) non-proliferative spent fuel bundles having a total isotopic uranium concentration of less than 12 wt %. Reprocessing of spent fuel bundles is also avoided.

Single-loop nuclear power plant with a pressurized coolant, comprising a power generating unit and a throttling device having an impeller, which are interconnected by an outlet pipe and a feed pipe, and a steam turbine connected to the throttling device and to a condenser connected to the throttling device, vertically divided into a steam zone, a high pressure zone, and a low pressure zone by horizontal partitions. The high pressure zone is connected to the the feed pipe and is connected to the low pressure zone by throttling nozzles provided in the partition between said zones, and the low pressure zone is connected to the steam zone by a vertical pipe which passes through the horizontal sealed partitions and the high pressure zone. The single-loop nuclear power plant is provided with an electric motor to rotate the impeller.

A nuclear reactor can include a pressure vessel for containing a pressurized moderator at a first pressure. The nuclear reactor can also include a plurality of fuel channels for a coolant fluid at a second pressure. The plurality of fuel channels are fluidly connected at inlet ends thereof to a coolant supply conduit and are adapted to receive nuclear fuel bundles and to be mounted within the pressure vessel and surrounded by the moderator. The outlet ends of the fuel channels are fluidly connected to a coolant outlet conduit to enable the coolant fluid to circulate from the coolant supply conduit through the fuel channels to the coolant outlet conduit. The plurality of fuel channels maintain separation between the coolant fluid circulating within the fuel channels and the moderator.

The present invention provides a top nozzle for use with PWR nuclear reactors and power plants, and in particular, VVER nuclear reactors. The top nozzle includes a plate portion having a peripheral portion; a hub portion spaced from the plate portion; a plurality of support portions extending from the plate portion to the hub portion; and at least one deflector portion extending inwardly from the peripheral portion at an acute angle with respect to the plate portion.

An improved measurement apparatus employs a support that includes a housing and that further includes a follower that is movably situated on the support. The follower is biased in a direction generally away from the housing. The housing and follower are receivable between a spring/dimple pair, and a sensor detects the distance between the housing and the follower when interposed between the spring and dimple. An indicator apparatus employs the signal to output and indication that is representative of the amount of compressive loading that will be applied to a fuel rod by the spring/dimple pair, and the indication is based at least in part upon the distance between the portion of the housing that is engaged with the dimple and the portion of the follower that is engaged with the spring.

A nuclear reactor system that, in one embodiment, utilizes natural circulation to circulate a primary coolant in a single-phase through a reactor core and a heat exchange sub-system. The heat exchange subsystem is located outside of the nuclear reactor pressure vessels and, in some embodiments, is designed so as to not cause any substantial pressure drop in the flow of the primary coolant within the heat exchange sub-system that is used to vaporize a secondary coolant in another embodiment, a nuclear reactor system is disclosed in which the reactor core is located below ground and all penetrations into the reactor pressure vessel are located above ground.

A reactor in-core instrument handling system in which the signal leads are routed from the instrument sensors through an outer sheath through the upper reactor internals and out of and around the sheath in a substantially tightly wound spiral before exiting the reactor vessel.