Abstract: The present disclosure is directed to systems and methods useful for the construction and operation of a Modular Integrated Gas High-Temperature Reactor (MIGHTR). The MIGHTR includes a reactor core assembly disposed at least partially within a core baffle within a first high-pressure shell portion, a thermal transfer assembly disposed at least partially within a flow separation barrel within a second high-pressure shell portion. The longitudinal axes of the first high-pressure shell portion and the second high-pressure shell portion may be collinear. The reactor core assembly may be accessed horizontally for service, maintenance, and refueling. The core baffle may be flexibly displaceably coupled to the flow separation barrel. Coolant gas flows through the reactor core assembly and into the thermal transfer assembly where the temperature of the coolant gas is reduced. A plurality of coolant gas circulators circulate the cooled coolant gas from the thermal transfer assembly to the reactor core assembly.

Abstract: A power module includes a reactor vessel containing a coolant and a reactor core located near a bottom end of the reactor vessel. A riser section is located above the reactor core, wherein the coolant circulates past the reactor core and up through the riser section. In one embodiment, a coolant deflector shield includes flow-optimized surfaces, wherein the flow-optimized surfaces direct the coolant towards the bottom end of the reactor vessel. In another embodiment, the reactor housing includes an inward facing portion that varies a flow pressure of the coolant and promotes a circulation of the coolant past a baffle assembly and towards the bottom end of the reactor vessel.

Abstract: A pressurized water nuclear reactor (PWR) includes a once through steam generator (OTSG) disposed in a generally cylindrical pressure vessel and a divider plate spaced apart from the open end of a central riser. A sealing portion of the pressure vessel and the divider plate define an integral pressurizer volume that is separated by the divider plate from the remaining interior volume of the pressure vessel. An internal control rod drive mechanism (CRDM) has all mechanical and electromagnetomotive components including at least a motor and a lead screw disposed inside the pressure vessel. Optionally CRDM units are staggered at two or more different levels such that no two neighboring CRDM units are at the same level. Internal primary coolant pumps have all mechanical and electromagnetomotive components including at least a motor and at least one impeller disposed inside the pressure vessel. Optionally, the pumps and/or CRDM are arranged below the OTSG.

Abstract: A natural circulation boiling water reactor provides to a chimney with a plurality of tubes. That is to say, each of the plurality of tubes partitions the coolant flow path above a core. Thus, unlike the conventional natural circulation boiling water reactor providing the flow path partition wall grid in which the plate members are made integral by welding and coolant flow paths are partitioned, the chimney of the natural circulation boiling water reactor can reduce the number of welded portions because the edges of the four corners of each flow path do not need to be welded. The natural circulation boiling water reactor can avoid removal as a single unit, as in the case of the flow path partition wall grid in the conventional natural circulation boiling water reactor, by detaching each tube. The chimney can be easily detached from the reactor pressure vessel.

Abstract: An upper plenum structure of a cooled pressure vessel for a prismatic very high temperature reactor which secures a space for coolant to supply to a core and also supports an upper reflector located inside a graphite structure on top of the core. The upper plenum structure includes a cavity structure where the coolant goes down in the upper plenum structure, a plurality of upper reflector supports formed with the cavity and supporting the upper reflector located on top thereof, and a plurality of coolant distributing blocks. Each of the coolant distributing blocks is coupled with a bottom portion of a respective one of the upper reflector supports and is located on top of the core in order to distribute the coolant collected in a cavity, formed by the upper reflector support, to the core. The coolant distributing blocks cooperate with the upper reflector supports to define the cavity structure.

Abstract: A direct vessel injection-type pressurized light water reactor (DVI-PLWR), in which an emergency core cooling water (ECC) is directly injected into a downcomer of a reactor vessel, is disclosed. In order to reduce the ratio of ECC bypass from the downcomer to a broken area of a cold leg in the case of a cold leg guillotine break (CLGB), such as a double-ended guillo4tine break (DEGB), a plurality of corrugations, having a V-shaped cross-section, are vertically arranged around each of the inner surface of a pressure vessel and the outer surface of a core barrel at regular intervals, with a vertical groove formed between two neighboring corrugations. The grooves phase-separate the ECC from a high-speed lateral flow of fluid running in the downcomer, and the separated ECC stagnates in the form of vortexes in the grooves, prior to flowing down to the lower section of the downcomer due to gravity.

Abstract: Within an upper plenum of a nuclear reactor, a portion of a heated coolant flows radially outward from a central portion of a core barrel (30) towards outlet nozzles (12) in a region of an upper core plate (21) extending outside of an outer periphery of the core along an inner wall of a core barrel (30). Portions of the coolant flows beneath the outlet nozzles (12). Thus, streams of heated coolant flowing in opposite directions may collide with each other. After collision, the flow directions of the heated coolant are changed to flow upward. Due to the collision, the coolant flow behavior becomes complicated and unstable, making it difficult to measure the temperature of the heated coolant with an outlet pipe (42) connected to the outlet nozzle (12).

Abstract: A mixing apparatus for a plurality of turbulently flowing fluid flows varying in temperature and/or composition includes a mixing chamber having a non-circular or predetermined cross section and a straight or curved center line. A plurality of single-conduit and parallel or radial deflector elements are disposed beside the mixing chamber and staggered in the direction of the center line. The deflector elements receive a fluid flow being oriented at an angle relative to the center line and staggered laterally. The deflector elements deflect the fluid flow tangentially into the mixing chamber. The mixing chamber has an outlet opening for the exit of a mixed fluid flow. In a gas-cooled, high-temperature nuclear reactor with a circular outline, the mixing apparatus has a horizontal, annular mixing chamber with a plurality of sectors. Horizontal annular conduits receive at least one fluid flow.

Abstract: A calandria for use in conducting the hot coolant of a nuclear reactor transversely. The calandria includes an upper plate and a lower plate which support tubes. The plates and tubes are enclosed in a shell which extends above the upper plate and has a supporting flange. The lower plate has holes for transmitting coolant into the region between the plates. The shell has openings whose boundaries mate with the outlet nozzles of the reactor. The tubes are of stainless steel and are dimensioned so that they have mass, stiffness and strength such that they are not subject to failure by the transverse flow of the coolant even at a high velocity.

Abstract: In a liquid metal-cooled fast neutron nuclear reactor which has a vessel sealed by a horizontal slab on which is suspended a core cover plug, the lower part of said plug comprises a deflecting grid, optionally duplicated by a thermal protection grid. Each of the grids is formed from a peripheral plate and modular plates defining between them passages for the liquid metal. The peripheral plates can be fixed to a perforated part of an external ferrule or can be suspended on a core cover plate by tie bolts and/or by certain of the control rod guidance sleeves. The modular plates are fixed to the other sleeves and traversed by sodium sampling tubes and tubes for housing thermocouples.

Abstract: A nuclear reactor plant housed in a steel pressure vessel with a small high temperature reactor, together with a steam generator arranged over said reactor and with at least two circulating blowers connected in parallel with each other and following the steam generator in line and is provided with a decay heat removal system located in the primary loop. This system is arranged above the steam generator and a shut-off device is provided between the two component. The shut-off device, which may actuated both passively--by difference pressure--and actively, by means of a drive, is designed so that in normal operation no hot gas may enter the decay heat removal system. In the decay heat removal mode the hot gas is conducted into the decay heat removal system and the steam generator is closed off the hot gas. Simultaneously, the cold gas coming from the decay heat removal system is returned through the steam generator or a gas conduit parallel to it, to the reactor core.

Abstract: An improved, grid-type flow distribution baffle for both providing lateral support for an array of heat exchange tubes and for generating a sludge-sweeping water current within a steam generator is disclosed herein. Generally, the improved baffle comprises a plurality of grid members for defining an array of square, tube-capturing cells, wherein each wall of the cell is closely adjacent and contactable with the side of the tube. Each of the cells is circumscribed by a flange at one of its ends which includes a substantially circular opening for allowing the tube to extend through the cell. The flange includes four flat portions spaced 90.degree. apart for rendering the flange flush with the center portion of each of the square walls which form the cells of the baffle. These flat portions allow the flange to closely circumscribe the tube without coming into contact with it, thereby avoiding "point contact" between the flange and the tube.

Abstract: A gas-cooled nuclear reactor includes a central core located in the lower portion of a prestressed concrete reactor vessel. Primary coolant gas flows upward through the core and into four overlying heat-exchangers wherein stream is generated. During normal operation, the return flow of coolant is between the core and the vessel sidewall to a pair of motor-driven circulators located at about the bottom of the concrete pressure vessel. The circulators repressurize the gas coolant and return it back to the core through passageways in the underlying core structure.If during emergency conditions the primary circulators are no longer functioning, the decay heat is effectively removed from the core by means of natural convection circulation. The hot gas rising through the core exits the top of the shroud of the heat-exchangers and flows radially outward to the sidewall of the concrete pressure vessel.

Abstract: A flow exchange element is presented which lowers temperature gradients in fuel elements and reduces maximum local temperature within high temperature gas-cooled reactors. The flow exchange element is inserted within a column of fuel elements where it serves to redirect coolant flow. Coolant which has been flowing in a hotter region of the column is redirected to a cooler region, and coolant which has been flowing in the cooler region of the column is redirected to the hotter region. The safety, efficiency, and longevity of the high temperature gas-cooled reactor is thereby enhanced.

Abstract: It is desirable to being able to vary the temperature of a loop exposed to a flow of hot gases in gas cooled nuclear reactor installations without thereby affecting the gas temperature in the other loops. It should be possible in an extreme case to take the affected loop out of service. This is achieved by arranging under the heat absorbing components in the hot gas channel a means for mixing gas connected with a cold gas line.

Abstract: A gas cooled, high temperature nuclear reactor is provided with a base plate arranged under the reactor core and over the bottom of the prestressed concrete pressure vessel serving as the bottom shield. The bottom shield comprises at least two plates arranged coaxially with respect to each other, one above the other. Each plate comprises several partially interconnected parts with the lower plate being placed at an axial and vertical distance from the bottom liner of the prestressed concrete pressure vessel and also from the upper plate.

Abstract: An improved design for shielding the component parts of a gas turbine power plant from radiation in a gas transport conduit is disclosed. The radiation shield and shielded gas conduit utilize a shielding arrangement comprising a displacement body arranged centrally within the conduit and a shielding annulus surrounding the displacement body. A shielding device provides for a first annular flow path for reactor cooling gas between the displacement body and shielding annulus and a second annular flow path for reactor cooling gas between the shielding annulus and the conduit housing. The shielding device provides protection against radiation without significantly disrupting the flow of the cooling gas in the conduits.