Abstract: In a preheating steam generator, the secondary water is introduced into a preheating space (34) by deformable extensions (52) from inverted J tubes (50) fixed to a semi-torroidal supply collector (48). The structure may be in the form of three tubular segments interconnected by swivel connections or couplings. The fitting of the extensions (52) between the tubes (50) and a plate (54) partly upwardly sealing the preheating space (34) is ensured by dismantable fixing members, without any welding.

Abstract: A steam generator which finds particular application as a superheater and reheater for transfer of heat from the gas coolant of a nuclear power reactor to a secondary fluid medium. The lower pressure reheater is located inside of the nuclear pressure vessel containing a reactor core, as is the high pressure main steam tube bundle which is comprised of an economizer/ evaporator tube bundle stage and an initial superheater tube bundle stage. The remaining superheater tube bundles which comprise the intermediate superheater and the finishing superheater are located outside of the nuclear pressure vessel, where they are heated regeneratively by reheat steam which is emerging from the nuclear pressure vessel at a temperature higher than required by the reheat turbine, The main steam system can be designed for either subcritical or supercritical pressure operation.

Abstract: A pressurized water reactor plant includes a reactor vessel (1), a reactor coolant pump (5) having an inlet connected by coolant pipe (2) to an outlet of the reactor vessel, a steam generator (10) having an inlet connected by a high-temperature side pipe (4) to the outlet of the pump (5) and an outlet connected by low-temperature side pipe (6) to an inlet of the reactor vessel. The steam generator has a horizontally extending shell (11) in the interior of which are horizontal U-shaped tube bundles (30) supported on tube support plates (20) and a pair of separator plates (18) which divide the interior of the shell into downcomer space (19). A partition plate (21) divides the interior of the shell into a U-shaped tube region (16) and a straight tube region (17). A moisture separator (27) is disposed above the heat transfer tubes (15) and a drain pipe (28) extends therefrom for draining impurities from the moisture separator to the U-shaped tube region (16) to which blowdown pipe (26) is connected.

Abstract: Stayrod arrangement for enhancing removal of sludge from nuclear heat exchangers. The stayrod arrangement includes a rod disposed adjacent a plurality of heat exchange tubes defining a tube lancing lane therebetween, the tube lane having sludge deposits therein. The rod has planer side portions oriented parallel to the direction of a fluid stream intended to flush the sludge deposits from the tube lane. An anti-rotation connector is connected to the rod for fixing the orientation of the rod with respect to the fluid stream.

Abstract: In a power station installation which is designed as a combined installation, i.e. consists of a gas turbine group (2, 3, 4, 5, 6, 8) and of a steam cycle (13, 16, 17, 18, 19, 20) downstream of the gas turbine group, the steam rate being generated in a waste heat boiler (13) by heat exchange with the exit gases from the gas turbine (5), a heat exchanger (7), which is a component of a closed cycle, with a high-temperature reactor (12) as the heat source is placed downstream of the last compressor (4) of the gas turbine group and upstream of the combustion chamber (8). In this heat exchanger (7), the compressed air undergoes vigorous caloric processing before it flows into the downstream combustion chamber (8) in which the final caloric processing of the hot gases for charging the gas turbine (5) is accomplished. The essential point here is that the pollutant emissions of any type, in particular CO.sub.

Abstract: A pump assembly which overcomes a shaft and cover cracking problem comprising a means for heating the seal purge water flowing along the shaft before it exits into an annulus (mixing region) thus reducing the temperature difference between the cooler seal purge water and the hotter product water prior to the mixing of the two waters.

Abstract: In a preheating steam generator, the feed water is distributed in a feed space (34) by inverted J-shaped tubes (50), so as to mix with the recirculation water passing out of the water-steam separators, which ensures the preheating of said feed water. In order to prevent the latter rising again above the upper edge (30a) of skirt (30) externally defining the space (34), use is made of one or more of the four procedures consisting of partly obstructing by a plate (52) the top of space (34), downwardly extending tube (50), increasing the diameter in the vicinity of their end and upwardly extending skirt (30).

Abstract: A nuclear reactor coolant pump for pumping reactor coolant fluid in a reactor coolant system includes a casing, a rotor and an impeller. The casing defines an inlet nozzle for receiving the fluid, a peripheral outlet nozzle for discharging the fluid, and a passage interconnecting the inlet and outlet nozzles through which the fluid can flow from the inlet to outlet nozzle. The rotor extends axially through the casing and has an end disposed adjacent passage. The impeller is mounted to the end of the rotor and disposed in communication with the passage and offset axially from the outlet nozzle of the casing. The impeller is rotatable with the rotor for drawing fluid into the casing through the inlet nozzle and discharging fluid from the casing tangentially through the peripheral outlet nozzle after flow through the passage. The outlet nozzle is composed of first and second wall portions defined above and below a plane extending generally parallel to the rotation axis of the impeller.

Abstract: A magnet pump has a magnet coupling for magnetically coupling an inner ring yoke and an outer ring yoke so as to magnetically transmit a torque from driving part to driven part of the pump to rotate the driven part. The magnet pump comprises magnetized magnet pieces bonded to the outer peripheral surface of the inner ring yoke and magnetized magnet pieces bonded to the inner peripheral surface of the outer ring yoke; non-magnetic cylindrical covers covering the surface of the inner ring yoke carrying the magnet pieces and the surface of the outer ring yoke carrying the magnet pieces, respectively; and mold resin parts filling the gap between the surface of the inner ring yoke and the associated non-magnetic cylindrical cover and the gap between the surface of the outer yoke ring and the associated non-magnetic cylindrical cover.

Abstract: A nuclear reactor internal pump has a pump shaft connected to a motor and rotatably passing through the bottom of a nuclear reactor pressure vessel, an impeller fitted to a front end portion of the pump shaft, a diffuser disposed on the downstream side of the impeller and having an inner cylinder fixed to the bottom of the reactor pressure vessel to surround the pump shaft, a purge water passage defined between the pump shaft and the inner cylinder to communicate with the inside of the reactor pressure vessel and a purge water supply device for supplying purge water to reactor pressure vessel through the passage. Further, the nuclear reactor internal pump comprises a clad dropping preventive device for trapping clad coming from the reactor pressure vessel in the passage to prevent the clad from dropping.

Abstract: A novel liquid nuclear reactor is described which comprises a reactor vessel that is connected through upper and lower liquid conduit means to one or more satellite tanks that contain a heat exchanger means and pump means.

Abstract: A liquid metal nuclear reactor has a sodium tank, having a closed arcuate bottom and upstanding cylindrical wall for containing a reactor core in a pool of liquid metal coolant, which is disposed in a guard tank having a bottom and upwardly extending cylindrical shell, and a support inside the guard tank adjacent the closed bottom of the sodium tank to support the two tanks in spaced relationship to each other. The sodium tank has an open top spaced from an overlying closure deck, with the cylindrical shell of the guard tank secured to the closure deck. The support is preferably an inner ledge. Support legs are also provided about the bottom periphery of the guard tank adapted to rest on a base mat of a reactor cavity.

Abstract: A system for variably delaying reactor trip and auxiliary feedwater system actuation resulting from a low water condition in a steam generator includes a plurality of sensors for determining both whether a low water condition exists in each steam generator and the power output level of the reactor. The length of the variable time delay is determined based upon the severity of the steam generator low water level condition and the reactor power output level. The reactor is then tripped and the auxiliary feedwater system actuated at the expiration of the time delay period if the steam generator water level is not restored.

Abstract: A fluidic device, in particular a vortex diode, has a vortex chamber formed by spaced apart end walls and a peripheral side wall. An axial port is provided in one end wall and a further port permitting tangential flow into or out of the chamber is provided in the other end wall.

Abstract: The apparatus for simulation of the operation of a pressurized water reactor comprises a reactor vessel with a heater, a steam generating vessel divided into a primary portion and a secondary portion. The reactor vessel and the primary portion of the steam generating vessel are connected in a primary feed loop equipped with a coolant pump, while the secondary portion of the steam generating vessel and the condenser unit are joined together in a secondary feed loop provided with a feed pump. With such an apparatus a steam generating vessel heating pipe break or fracture can be simulated when the reactor vessel and a primary part of an additional steam generating unit are joined in another primary feed loop parallel to the primary feed loop having the cooling pump and a secondary part of the additional steam generating unit is connected with the aid of at least two branch pipes to the secondary feed loop.

Abstract: In a nuclear or fossil steam turbine, a novel arrangement of structural elements is utilized to support a stationary blade ring, that is normally supported on a radial wall of a turbine inner casing, on an upstream separate blade ring. Thus, the stationary blade ring is separated from undesirable thermal deformations that naturally occur in turbine inner casing structures. A downstream stationary blade ring is fit to the upstream separate blade ring to maintain its relative position, and is attached using connection bolting and alignment pins or dowels. The inventive system is compact, low cost, and reduces bladepath seal leakage, leading to improved heat rate.

Abstract: A heat radiator useful for expelling waste heat from a power generating system aboard a space vehicle is disclosed. Liquid to be cooled is passed to the interior of a rotating bubble membrane radiator, where it is sprayed into the interior of the bubble. Liquid impacting upon the interior surface of the bubble is cooled and the heat radiated from the outer surface of the membrane. Cooled liquid is collected by the action of centrifical force about the equator of the rotating membrane and returned to the power system. Details regarding a complete space power system employing the radiator are given.

Abstract: A novel concrete core support structure for nuclear reactors is described.

Abstract: A DC electromagnetic pump for circulating liquid-metal coolant through a nuclear reactor. The metal is circulated through ducts which are disposed in an annulus. Opposite walls of the ducts serve as electrodes to transmit current radially with respect to the annulus through the liquid metal in the ducts. The radial current is supplied by a homopolar generator. The walls of the ducts joining the electrodes are of composite structure. They include metal plates insulated from the electrodes backed up by insulators and are capable of withstanding the pressure of the liquid metal. These composite wall structures may also be of thin metal strips of low electrical conductivity backed up by sturdy insulators. A circumferential electric field is supplied to the liquid metal by a toroidal electromagnet which has core sections interposed between the ducts. The windings of the electromagnet extend over the ducts to suppress flux leakage.

Abstract: A nuclear reactor steam generating plant having a pressurized water reactor with a single discharge line to a steam generator, with the primary coolant passing through the steam generator tubes divided into two streams in the channel head outlet section that is divided into two subsections. Discharge nozzles on the outlet subsections direct the streams of coolant to hermetically sealed pumps directly attached to the discharge nozzles, with the pumps directing the cooled primary coolant back to the reactor vessel. A single removable closure is provided for access to both subsections of the outlet section of the channel head, with a removable portion provided on the divider plate that forms the subsections.

Abstract: A liquid metal circulation system is disclosed which is suitable for use in connection with a nuclear reactor. The circulation system of the invention includes an electromagnetic pump having a central core body featuring a helically configured, six pole magnet, optionally directing flow to one or a series of flow couplers, which generates sufficient liquid metal flow to be used as the primary circulation means for the nuclear reactor. The increased safety, design simplicity, and efficiency of such circulation system overcomes many of the drawbacks of conventional circulation systems.

Abstract: A nuclear power plant (heating plant) with a helium cooled high temperature reactor with spherical fuel elements, located in a cylindrical prestressed concrete pressure vessel, suitable for supplying heat (for local or remote heating or the generation of process steam) with a capacity of approximately 50 to 300 MWth. The high temperature reactor core is located out of center in the prestressed concrete pressure vessel, with at least two heat exchangers installed adjacent and offset in the upward direction relative to it. The heat exchangers are each connected on the secondary side to an intermediate circulation loop including an intermediate heat exchanger and circulating pump, and have an auxiliary circulation loop with a recooling system connected in parallel with them for the removal of decay heat. The auxiliary loops are closed in normal operation.

Abstract: A mechanism for anti-seismic connection between the pump base and the dome of a nuclear power station is characterized by the fact that the pump base (1) is equipped externally with a flange (3) seated with a slight axial play and significant radial play within a socket (4) which is a fixed cover closed by and able to slide axially with a slight play within a cylindrical reinforcement (7a) of the dome (7). A radial damping mechanism (8) is installed between the flange (3) and the socket (4) to permit slow displacements between these two entities, but to block abrupt displacements. The lower end of the socket (4) comprises a collar (15) equipped with a metallic journal (16) which can slide axially within a bearing (17) firm with the internal wall of the dome (7).

Abstract: A double tube modular coil steam generator is provided in which a multiplicity of inner tubes conducting water are individually surrounded by outer tubes containing liquid metal as a heat transfer agent. The double tubes form into coils, providing a large surface area while conserving space. Immersion of the double tube coil in hot liquid metal, e.g., from the core of a nuclear reactor, causes efficient transfer of heat across the liquid metal in the outer tube to the water in the inner tube, creating superheated steam, which can be cycled to a turbine and converted to electrical power.The efficiency, reliability and safety of the multiple double tube design of the steam generator obviates the necessity of many secondary heat removal and emergency components in addition to conserving space and material. The modular design allows ease of operation, fabrication and repair.

Abstract: A nuclear reactor is provided characterized by a circulating liquid metal cooling system comprising a heat exchanger having a closed intermediate heat transfer fluid circuit in the form of a helical coil juxtaposed with the liquid metal coolant system, which effectively transfers heat through the intermediate circuit to a secondary fluid circuit. The intermediate heat transfer fluid circuit, which completely separates the liquid metal cooling system from the secondary fluid circuit, prevents potentially dangerous reactions between the primary liquid metal coolant and the secondary fluid (e.g., water).The efficiency, reliability and safety of the heat exchanger features eliminates the need for many secondary heat removal and emergency components in the design of the nuclear reactor.

Abstract: Disclosed is a nuclear reactor employing a liquid metal and containing a reactor vessel, and at least one satellite vessel, and a hot and a cold leg connecting means which combine to form a continuous pathway for the flow of the liquid metal between the reactor vessel and the at least one satellite vessel. Each of the hot leg and cold leg connecting means contain a bellows assembly which are readily accessible for repair or replacement without draining the reactor vessel of the liquid metal.

Abstract: An orifice insert for placement in nuclear steam generator tubes comprises a sleeve having a top end, a bottom end located in use in the vicinity of the opening of the steam generator tube, and a longitudinal axis. An orifice is disposed in and extends the entire length of the sleeve. At least a portion of the outer diameter of the sleeve is substantially the same as the inner diameter of the steam generator tube. The orifice comprises at least a lower and an upper cylindrical portion which are coaxial with the longitudinal axis of the sleeve, wherein the diameter of the upper cylindrical portion is smaller than the diameter of the lower cylindrical portion.

Abstract: A steam generator level measurement system of the type having a reference leg which is kept full of water by a condensation pot, has a separator pot in the connecting line between the condensation pot and the steam phase in the steam generator to remove excess liquid from the steam externally of the steam generator. The separator pot has an expansion chamber which slows down the velocity of the steam/liquid mixture to aid in separation, and a baffle, which directs steam introduced at the top of the chamber on one side of the baffle to flow downward and then upward for discharge on the other side of the baffle to avoid direct liquid penetration into the line connected to the condensate pot.

Abstract: A steam control value for a steam chest such as used in a nuclear plant with an improved muffler portion. The muffler portion includes window openings of predetermined dimensions. A ring portion is below the window opening having a plurality of pockets. Each pocket is aligned with one of the window openings. This design greatly reduces vibrations which may damage the valve.

Abstract: A nuclear reactor coolant pump impeller/shaft assembly, including an impeller nut with a plurality of conduits formed therein in fluid communication with both ends of a passageway formed in the impeller nut in contact with a portion of the shaft, the fluid inlet and fluid outlet of the pump. Due to this conduit/passageway combination, coolant is effectively pumped into the passageway to heat the shaft to a temperature approximately equal to that of the impeller, thus creating thermal balance of the pump's component parts and providing improved pump efficiency.

Abstract: A support system for connection to an outer surface of a J-shaped steam generator for use with a nuclear reactor or other liquid metal cooled power source. The J-shaped steam generator is mounted with the bent portion at the bottom. An arrangement of elongated rod members provides both horizontal and vertical support for the steam generator. The rod members are interconnected to the steam generator assembly and a support structure in a manner which provides for thermal distortion of the steam generator without the transfer of bending moments to the support structure and in a like manner substantially minimizes forces being transferred between the support structure and the steam generator as a result of seismic disturbances.

Abstract: Steam generator, particularly for a pressurized water nuclear reactor, which has in per se known manner a water box in which circulates the hot water from the reactor core, and then enters a series of heat exchange tubes installed in a tube plate and extending within a secondary ferrule traversed by the secondary water, wherein the water box, secondary ferrule and tube plate are three independent, disassemblable assemblies.

Abstract: The invention relates to a device for thermal protection of a component of a fast-neutron nuclear reactor consisting of at least one shell (10, 11) of a diameter smaller than the diameter of the passage (2) for the component (4) but greater than the diameter of the component (4). The length of the shell (10, 11) fixed under the flange (5) of the component (4) is greater than the slab thickness (1). Circulation of blanket gas (13) is established in the annular space (7) between the component (4) and the passage (2).The invention applies in particular to nuclear reactors cooled with liquid sodium whose upper level is under an argon blanket.

Abstract: A steam generator recirculating system for use with the secondary coolant of the steam generators of a pressurized water nuclear reactor is comprised of two separate recirculation loops. Each loop contains a common discharge line into which a secondary coolant discharge line and blowdown discharge line from at least one steam generator feeds. The common discharge line contains a recirculating pump and a heat exchanger, with bypasses provided about each. The common discharge lines discharge into branch feedlines back to the steam generators. The system may also be used as a blowdown system leading to a steam generator blowdown processing system and may include a water purification means and used to purify the secondary coolant water. Cross-connecting lines are provided to permit optional use of a particular loop with various steam generators.

Abstract: A method and an apparatus for feed-water control in a steam generating plant such as a boiling water nuclear reactor having motor-driven feed-water pumps and turbine-driven feed-water pumps for supplying feed-water to a steam generator, in which the feed-water flow through the turbine-driven feed-water pumps is controlled to be maintained constant, while reducing the recirculation flow discharged from the turbine-driven feed-water pumps and fed back to the inlet thereof, so as to reduce the amount of feed-water supplied to the steam generator by the motor-driven feed-water pumps, whereby variation of water level in the steam generator during switching between the two kinds of feed-water pumps is suppressed.

Abstract: A double tank type fast breeder reactor including a reactor vessel, and a tank of a secondary cooling system located outside the reactor vessel. The reactor vessel contains a core and installations of a primary cooling system, and the tank contains circulation pumps of the secondary cooling system and steam generators alternately located in the tank, and a body of sodium serving as a secondary coolant located at outer peripheries of the circulation pumps and steam generators. The tank includes an annular portion defined in the tank between walls of the tank and reactor vessel serving as a path of circulation of the secondary coolant.

Abstract: A model steam generator including a system for facilitating the inspection of the sample tubes within the boiler vessel of the model generator is disclosed herein. The system includes means for detachably connecting the tubesheet from the primary and secondary sides of the boiler vessel. In the preferred embodiment, both end of the tubesheet and the abutting ends of the primary and secondary sides of the boiler vessel are circumscribed by tapered flanges. The tubesheet is detachably connected from the primary and secondary sides of the boiler vessel by means of Grayloc.RTM.-type annular clamps which are circumscribed by grooves for receiving the abutting flanges at the tubesheet joints. Additionally, the system includes a frame for suspending the secondary side of the boiler vessel, a wheeled cart having a jack for both laterally and vertically moving the primary side of the boiler vessel and the tubesheet into a clamping position onto the secondary side of the boiler vessel.

Abstract: Steam generation and reheat apparatus and method. After flowing in series through a superheater and an evaporator, a heat supply fluid is flowed through both an economizer and a reheater in parallel flow to improve the Rankine Cycle thermal efficiency in plants such as nuclear power plants where the heat supply fluid temperatures are limited.

Abstract: To control leaks between the primary and secondary circuits of a pressurized water reactor system, a pressure equalization is brought about by shutting off the live steam line and the feedwater line connected to the secondary circuit of the steam generator and by increasing the pressure prevailing in the resulting isolated secondary circuit by the injection of gas.

Abstract: The invention relates to a secondary heat transfer circuit for a liquid metal-cooled nuclear reactor, as well as to a steam generator which can be used in such a circuit.According to the invention, a downstream buffer tank and an upstream buffer tank are arranged within the steam generator. The upstream buffer tank is annular and surrounds and communicates with an area of the generator by which the liquid metal flows downwards between the exchange zone and the discharge tube. The pressure of the neutral gas pocket in the downstream buffer tank is higher than that of the pocket of the upstream buffer tank.Application to fast neutron nuclear reactors.

Abstract: Device for the post-accident cooling of the confinement enclosure of a pressurized water nuclear reactor, wherein it comprises in series a turbine supplied by the hot, humid air contained in the enclosure, a condenser in which the air from the turbine is dried and cooled by thermal contact with an external cooling fluid, and a compressor actuated by the turbine, and which returns the dried air into the enclosure.

Abstract: Device for producing steam by heat exchange between a heat-transfer liquid metal and feed water, comprising a cylindrical casing (31) receiving liquid metal in its upper part through at least one nozzle (35, 36). The liquid metal enters in a space bounded by a central sleeve (38), a lower perforated distribution plate (40) and a first perforated plate (45). A second perforated plate (46) parallel to the latter is separated therefrom by perforated crosspieces (47). The perforated plates (45 and 46) are disposed between the sleeve (38) and a peripheral sleeve (39). The three plates (40, 45 and 46) have identical perforations which are aligned in the vertical direction. The distribution plate (40) is provided with stiffening ribs (44) ensuring the distribution of the liquid metal thereon. The invention applies, in particular, to fast breeder nuclear reactor steam generators cooled by liquid sodium.

Abstract: Apparatus for superheating steam. In accordance with one aspect of the invention, the apparatus is provided with two banks of inclined tubes extending upwardly from an outlet header to respective inlet headers. The banks of tubes are disposed in the flow path of main steam through the apparatus and provide a flow of vapor for adding superheat to the main steam. In accordance with another aspect of the invention, the tubes extend upwardly to closed ends whereby a lower header acts as both a vapor inlet and condensate outlet. In accordance with another aspect of the invention, the vapor is provided by a fossil fuel-fired vapor generator to superheat main steam which has been provided by a nuclear steam generator.

Abstract: The invention relates to a secondary heat transfer circuit for a liquid metal nuclear reactor.Each loop of the main circuit has in order a steam generator, a pump, and at least one heat exchanger positioned in the reactor vessel. A downstream buffer tank is located in the pipe connecting the generator to the pump, whilst the upstream buffer tank can be positioned either in the generator, or outside the latter.Application to the generation of electric power by means of a fast neutron reactor.

Abstract: Device for safety injection on a pressurized-water nuclear reactor, including at least two separate injection trains each consisting of at least one set of means (60, 61, 62) for pumping water from a reservoir (58, 59) arranged outside the safety enclosure, and a main injection pipe (64, 65) connected to a distributor drum (70, 70') for the distribution of the water into each of the cold arms of the primary circuit. From each of the main pipes (64, 65) there is tapped off outside the enclosure an auxiliary pipe (84, 85) in which is located a block valve (96, 97). The auxiliary pipes (84, 85) include a branching-off (86, 87, 88, 89) for placing the auxiliary pipe (84, 85) into communication firstly with a cold arm (52, 53, 54, 55) and secondly with a hot arm (56, 56'). The invention is applicable in the event of leakage in the primary circuit of a pressurized-water nuclear reactor.

Abstract: A self-contained modular nuclear reactor which can be prefabricated at a factory location, nuclear-certified at the factory, transported to a field location for final assembly and connection to a large-scale electric-power generating facility. The modular reactor includes a prefabricated nuclear heat supply module and a plurality of shell segments which can be assembled about the heat supply module and which provide a form for the pouring and curing of a cementatious biological shield about the heat supply module. The modular reactor includes passive shutdown heat removal systems sufficient to render the reactor safe in an emergency. A large-scale power plant arrangement is disclosed which incorporates a plurality of the modular reactors.

Abstract: The present invention pertains to a means of connecting a primary pump (intended for sodium or the like) and the delivery pipe within a breeder reactor of the swimming-pool type, permitting limited movement of the pump and the pipe in relation to one another to occur, while also maintaining adequate sealing between low pressure areas and high pressure areas within said pump. The connecting component consists of a cylindrical sleeve (11) which, in a resting position, is situated along the same axis (1) as the delivery pipe for the pump. A cylindrical guide (3) which is fastened to the pump is fitted into the aforementioned pipe, and a sealing joint (13) is situated between the sleeve (11) and the guide (3). There is also a horizontal annular collar (12) which is fastened to the aforementioned sleeve (11), resting upon an annular rim (5) situated upon the upper flange (7) of the pipe (4).

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 process to produce energy with the fusion of hydrogen and carbon in a thermonuclear fusion reaction with an extension, wherein carbon targets are bombarded with a beam of fast protons at temperatures of 457 kev. The thermonuclear fusion reaction produces two helium atoms and two neutron particles.The neutrons are captured by a blanket of lithium 7 which cools the walls and carries the heat to a turbulent heat exchanger, then to a generator and turbine complex which generates energy, giving tritium as a sub-product.The helium atoms plus alpha particles and protons are extracted to the reactor's extension where, combined with the tritium emitted in the first cycle of the thermonuclear fusion reactor, they go to generate more energy.This process becomes self-maintaining once the thermonuclear fusion reactor is put into operation.

Abstract: A system for producing process heat which employs a high temperature gas cooled nuclear reactor and is adapted to provide process heat at temperatures higher than the reactor core temperature. The system includes a closed secondary loop having a working fluid heated to a first temperature by the coolant of the reactor core in an intermediate heat exchanger, and has a heat pump connected in the secondary loop which increases the temperature of the working fluid to a temperature above the core coolant temperature. Steam in an independent closed nuclear steam supply circuit is heated directly by the core coolant in a steam generator and provides steam for make-up driving power to the heat pump in the secondary loop. The steam supply circuit also supplies additional steam to generate electric power and process steam. The size of the intermediate heat exchanger is minimized to enable maximum utilization of the steam supply circuit.