A pressurized water reactor (PWR) includes a vertical cylindrical pressure vessel and a nuclear reactor core disposed in a lower vessel section. A hollow cylindrical central riser is disposed concentrically inside the pressure vessel. A downcomer annulus is defined between the central riser and the pressure vessel. A reactor coolant pump (RCP) includes (i) an impeller disposed above the nuclear reactor core and in fluid communication with the downcomer annulus to impel primary coolant downward through the downcomer annulus, (ii) a pump motor disposed outside of the pressure vessel, and (iii) a drive shaft operatively connecting the pump motor with the impeller. The PWR may include an internal steam generator in the downcomer annulus, with the impeller is disposed below the steam generator. The impeller may be disposed in the downcomer annulus. The RCP may further comprise a pump casing that with the impeller defines a centrifugal pump.

A pressurized water nuclear reactor (PWR) includes a pressure vessel having a lower portion containing a nuclear reactor core comprising a fissile material and an upper portion defining an internal pressurizer volume. A condenser is secured to, and optionally supported by, the upper portion of the pressure vessel. A condenser inlet is in fluid communication with the internal pressurizer volume. A heat sink is in fluid communication with the condenser such that the condenser operates as a passive heat exchanger to condense steam from the internal pressurizer volume into condensate while rejecting heat to the heat sink. A condenser outlet connects with the pressure vessel to return condensate to the pressure vessel. A single metal forging having a first end welded to the pressure vessel and a second end welded to the condenser inlet may provide the fluid communication between the condenser inlet and the internal pressurizer volume.

A pressurized water reactor (PWR) comprises: a nuclear core comprising a fissile material; a cylindrical pressure vessel having a vertically oriented cylinder axis and containing the nuclear core immersed in primary coolant water; and a hollow cylindrical central riser disposed concentrically with and inside the cylindrical pressure vessel. A downcomer annulus is defined between the hollow cylindrical central riser and the cylindrical pressure vessel. The hollow cylindrical central riser has a radially expanding upper orifice that merges into an annular divider plate that separates an upper plenum above the annular divider plate from a lower plenum below the annular divider plate. The upper plenum is in fluid communication with the radially expanding upper orifice and the lower plenum is in fluid communication with the downcomer annulus. A weir may extend away from a bottom wall of the lower plenum into the lower plenum. An emergency core cooling system (ECCS) return line nozzle may be arranged to inject water into the upper plenum. A pump support plate spans the inner diameter of the cylindrical pressure vessel and forms a portion of the pressure boundary of the cylindrical pressure vessel, and reactor coolant pumps (RCPs) are supported by the pump support plate. Alternatively, reactor coolant pumps (RCPs) are supported by an arcuate annular ledge formed in the upper portion of the cylindrical pressure vessel.

A method for retrofitting a nuclear power plant which include dismantling and removing all the components of the primary circuit apart from the LWR reactor vessel, which is essentially emptied of all material and neutralized, subsequently replacing a part of these components with subassemblies that are each made up of an integrated SMR reactor and a mixed concrete/metal structure, which mixed concrete/metal structure is also used as a reactor pit for the SMR reactor, which reactor pit is advantageously filled with water, anchoring the SMR to the inside of the reactor building and advantageously contributing to the third confinement barrier while ensuring minimal disruption to the infrastructure of the reactor building.

The present disclosure discloses a method and an apparatus for determining a rod position of a control rod of a pressurized water reactor, a rod position measurement system and a non-volatile computer-readable storage medium. In the method for determining a rod position of a control rod, a rod position determination condition of each rod position of a full stroke is determined according to a Gray code bit signal voltage of the rod position of the full stroke, and Gray code bits and corresponding Gray code bit signal voltage values are enabled to correspond to rod positions, so that according to a real-time signal and by means of a rod position interval and the rod position determination condition, each Gray code bit signal, which is currently obtained in real time, can be accurately positioned to each rod position point, so that rapid and accurate full-stroke rod position measurement is realized.

A hybrid power generation system which includes a nuclear facility comprising a nuclear reactor and an exclusion zone. A thermal energy storage, a nuclear steam supply system located, and a solar energy collection system are all located within the exclusion zone. The thermal energy storage vessel contains a thermal mass composition operable to store thermal energy. The nuclear steam supply system has a nuclear reactor and a working fluid loop. The working fluid loop is configured to circulate a working fluid from a steam generator through the thermal energy storage vessel to absorb thermal energy and heat the working fluid for introduction to an electricity generating system. The solar energy collection system includes a heat transfer loop heated via a solar collector. The heat transfer loop is configured to circulate a heated heat transfer fluid to add thermal energy to the thermal mass composition in the thermal energy storage vessel.