This invention fundamentally changes the basic design principles adopted over the last 50 years that addresses the basic root cause for the station blackout threat faced by all nuclear power plants. The Fukushima nuclear accident that occurred in Japan in 2011 could have turned benign had the plant implemented this invention. It uniquely utilizes the decay heat directly from the reactor core through steam cycles to produce useable power onsite with one or a number of steam turbines of less capacities in combination with compatible electric generators. Such arrangement is reliable to be the onsite energy source. The electric power produced by generators attached to these steam turbines could support necessary all safety functions. The result is that during the first week of urgent threat to the nuclear reactor, there will always be electric power available to run the safety equipment, computers, lighting and other vital devices continuously.

An autonomous self-powered system for cooling radioactive materials comprising: a pool of liquid; a closed-loop fluid circuit comprising a working fluid having a boiling temperature that is less than a boiling temperature of the liquid of the pool, the closed-loop fluid circuit comprising, in operable fluid coupling, an evaporative heat exchanger at least partially immersed in the liquid of the pool, a turbogenerator, and a condenser; one or more forced flow units operably coupled to the closed-loop fluid circuit to induce flow of the working fluid through the closed-loop fluid circuit; and the closed-loop fluid circuit converting thermal energy extracted from the liquid of the pool into electrical energy in accordance with the Rankine Cycle, the electrical energy powering the one or more forced flow units.

A steam generator comprising a vessel having an inlet and an outlet, and in use a primary fluid flow enters the vessel through the inlet and exits the vessel through the outlet. A plurality of modules are connected in series and at least partially housed within the vessel, and each module comprises at least one tube. The modules are arranged such that at least one tube of one module is coaxial with at least one tube of an adjacent module so as to define a conduit through which a secondary fluid can flow from one module to an adjacent module.

A compact pressurized water nuclear reactor having connected to the reactor pressure vessel a plurality of horizontal pressure vessels, with all the horizontal pressure vessels connected to the reactor pressure vessel by a single connection between the respective nozzle of the reactor pressure vessel with the respective nozzle of each horizontal pressure vessel.

An autonomous self-powered system for cooling radioactive materials comprising: a pool of liquid; a closed-loop fluid circuit comprising a working fluid having a boiling temperature that is less than a boiling temperature of the liquid of the pool, the closed-loop fluid circuit comprising, in operable fluid coupling, an evaporative heat exchanger at least partially immersed in the liquid of the pool, a turbogenerator, and a condenser; one or more forced flow units operably coupled to the closed-loop fluid circuit to induce flow of the working fluid through the closed-loop fluid circuit; and the closed-loop fluid circuit converting thermal energy extracted from the liquid of the pool into electrical energy in accordance with the Rankine Cycle, the electrical energy powering the one or more forced flow units.

Disclosed is a gas turbine and pressurized water reactor steam turbine combined circulation system, using a heavy duty gas turbine and a pressurized water reactor steam turbine to form a combined circulation system. Heat of the tail gas of the gas turbine is utilized to raise the temperature of a secondary circuit main steam from 272.8 C., and the temperature of the secondary circuit main steam slides between 272.8 C. and 630 C. according to different pressurized water reactor steam yields and different input numbers and loads of the heavy duty gas turbine. The system has a higher heat efficiency than that of the pressurized water reactor steam turbines in the prior art; and as for the electric quantity additionally generated by gas, the heat efficiency of the system is also significantly higher than that of gas-steam combined circulation in the prior art.

Disclosed is a gas turbine and pressurized water reactor steam turbine combined circulation system, using a heavy duty gas turbine and a pressurized water reactor steam turbine to form a combined circulation system. Heat of the tail gas of the gas turbine is utilized to raise the temperature of a secondary circuit main steam from 272.8 C., and the temperature of the secondary circuit main steam slides between 272.8 C. and 630 C. according to different pressurized water reactor steam yields and different input numbers and loads of the heavy duty gas turbine. The system has a higher heat efficiency than that of the pressurized water reactor steam turbines in the prior art; and as for the electric quantity additionally generated by gas, the heat efficiency of the system is also significantly higher than that of gas-steam combined circulation in the prior art.

Supports with integrated sensors for nuclear reactor steam generators, and associated systems and methods, are disclosed. A representative method for forming a nuclear-powered steam generator includes forming an instrumented support, the instrumented support including a carrier portion and a retainer portion, with at least one of the carrier portion or the retainer portion being integrally formed with a sensor via an additive manufacturing process. The method can further include coupling the sensor to a communication link, supporting a helical steam conduit on the instrumented support, and installing the helical steam conduit and the instrumented support in a nuclear reactor. The helical steam conduit is positioned along a primary flow path, which is in turn positioned to circulate a heated primary flow in thermal communication with the helical steam conduit.

A compact pressurized water nuclear reactor having connected to the reactor pressure vessel a plurality of pressure vessels connected by nozzles, and connected by curved horizontal pressure vessel heads by having their central axis horizontal, with reduced stress and simple single connection between the respective nozzle of the reactor pressure vessel with the respective nozzle of each curved horizontal pressure vessel heads, all with the same internal and external design pressure.

A steam generator for a nuclear reactor comprises plenums proximate with a first plane, wherein the first plane intersects a bottom portion of a riser column of a reactor vessel. The steam generator may further comprise plenums proximate with a second plane, approximately parallel with the first plane, wherein the second plane intersects a top portion of the riser column of the reactor vessel. The steam generator may further include a plurality of steam generator tubes that convey coolant from a plenum located proximate with the first plane to one of the plenums proximate with the second plane.