A control rod assembly for a nuclear reactor having a reactor core and a pressurized fluid source, including a control rod disposed within a control rod sleeve, a lead screw that is selectively secured to the control rod, a trip latch that is secured to a bottom end of the lead screw, the trip latch being selectively securable to a top end of the control rod, a control rod drive motor that is operably connected to the lead screw, and a valve that is in fluid communication with the pressurized fluid source of the nuclear reactor and is movable between a first position and a second position, wherein in the second position of the gas valve the trip latch is in an open position.

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 method includes assessing operational characteristics of the fuel assembly, the assessing including determining if the fuel assembly is to be placed in a controlled location in the reactor core, a controlled location being positioned adjacent to a control blade that is to be utilized, and configuring the sidewalls of the outer channel by making at least a first select sidewall of the outer channel a reinforced sidewall, the remaining sidewalls of the outer channel, other than the at least a first select sidewall, being non-reinforced sidewalls. The entirety of the reinforced sidewall as a whole is at least one of thicker and made from a material that is more resistant to radiation-induced deformation as compared to an entirety of the non-reinforced sidewalls.

A method and apparatus of limiting power of a boiling water nuclear reactor system includes a reactor pressure vessel, a reactor core disposed in the reactor pressure vessel, a core shroud surrounding the reactor core, a downcomer region disposed between an inner surface of the reactor pressure vessel and the core shroud, a steam line connected to an upper end of the reactor pressure vessel and a condenser system that receives steam from the reactor pressure vessel. A portion of the condenser system condensate is returned to the reactor pressure vessel of the boiling water reactor inside the core barrel above the core rather than into the downcomer. Returning the condensate in this way increases the effectiveness of an isolation condenser system or if the condensate is a portion of the feedwater from the main condenser it provides an effective means to regulate core flow and core power.

Nuclear reactors have very few systems for significantly reduced failure possibilities. Nuclear reactors may be boiling water reactors with natural circulation-enabling heights and smaller, flexible energy outputs in the 0-350 megawatt-electric range. Reactors are fully surrounded by an impermeable, high-pressure containment. No coolant pools, heat sinks, active pumps, or other emergency fluid sources may be present inside containment; emergency cooling, like isolation condenser systems, are outside containment. Isolation valves integral with the reactor pressure vessel provide working and emergency fluid through containment to the reactor. Isolation valves are one-piece, welded, or otherwise integral with reactors and fluid conduits having ASME-compliance to eliminate risk of shear failure. Containment may be completely underground and seismically insulated to minimize footprint and above-ground target area.

To provide a corrosion mitigation method for carbon steel pipe that can further reduce corrosion of the carbon steel pipe. In a BWR plant, oxygen is injected from an oxygen injection device 30 into a clean up system pipe 18 which is constituted by a Cr-containing carbon steel pipe containing Cr in a range of larger than 0.052 wt % and less than 0.4 wt % and being in communication with a RPV 3, and reactor water of 150 C. having a dissolved oxygen concentration of 30 g/L is generated. The reactor water is brought into contact with an inner surface of the clean up system pipe 18 to perform an oxidizing treatment on the inner surface, and an oxide film containing Cr is formed on the inner surface. Thus, after the oxide film is formed, hydrogen is injected into the reactor water in the RPV 3 through a water supply pipe 11 in communication with to the RPV 3, and even when the dissolved oxygen concentration in the reactor water in contact with the inner surface of the clean up system pipe 18 is reduced to 2 g/L, corrosion of the clean up system pipe 18 is remarkably mitigated.

Controlled-debris elements inhibit the formation of a fibrous/particulate debris bed that unduly increases the pressure head loss through the perforated plates of strainers in a nuclear power plant emergency core cooling system. In a loss of cooling accident, pumps draw cooling water through the plates, which retain on their surfaces fibrous material in the circulating water to prevent it from reaching the pumps while permitting entrained particulate matter to pass through the perforations. The controlled-debris elements have a specific gravity substantially the same as the circulating water so they are entrained in the cooling water that is drawn toward the strainers and intimately intermix with the fibrous and particulate matter in the cooling water. The elements are configured to provide open structures in the bed formed on the plate surfaces to distribute fibers in the flow away from the surface and maintain cavities between the elements for the particulates.

An exemplary embodiment can include an apparatus including: an internal-external hybrid nuclear reactor, which can include: at least one reciprocating internal engine; and at least one external reactor integrated with said at least one reciprocating internal engine. The reciprocating engine can receive nanofuel (including moderator, nanoscale molecular dimensions & molecular mixture) internally in an internal combustion engine that releases nuclear energy. A method of operating the hybrid nuclear reactor can include operating the reciprocating internal engine loaded with nanofuel in spark or compression ignition mode. A method of cycling the reciprocating internal engine, can include compressing nanofuel; igniting nanofuel; capturing energy released in nanofuel, which is also the working fluid; and using the working fluid to perform mechanical work or generate heat.

Nuclear reactors have very few systems for significantly reduced failure possibilities. Nuclear reactors may be boiling water reactors with natural circulation-enabling heights and smaller, flexible energy outputs in the 0-350 megawatt-electric range. Reactors are fully surrounded by an impermeable, high-pressure containment. No coolant pools, heat sinks, active pumps, or other emergency fluid sources may be present inside containment; emergency cooling, like isolation condenser systems, are outside containment. Isolation valves integral with the reactor pressure vessel provide working and emergency fluid through containment to the reactor. Isolation valves are one-piece, welded, or otherwise integral with reactors and fluid conduits having ASME-compliance to eliminate risk of shear failure. Containment may be completely underground and seismically insulated to minimize footprint and above-ground target area.

When all fuel assemblies loaded in a region excluding an outermost periphery of the reactor core in an Nth operation cycle belong to the first fuel assembly, and all fuel assemblies loaded in the region excluding the outermost periphery of the reactor core in a (N+m) th (m>1) operation cycle belong to the second fuel assembly, the number of new loaded second fuel assemblies in the (N+m) th operation cycle is greater than the number of new loaded second fuel assemblies in a (N+m1) th operation cycle which is one operation cycle before the (N+m) th operation cycle, and a cycle burnup in the (N+m) th operation cycle is greater than a cycle burnup in the (N+m1) th operation cycle.