Abstract: The invention concerns a method of estimating when dryout may occur in a nuclear light water reactor of the boiling water reactor kind. The method includes the use of a formula which expresses the local dryout property of the nuclear reactor. The formula includes at least a first and a second factor. The first factor is a first function that describes how the dryout property depends on the flow of the cooling medium through the nuclear fuel arrangement. The second factor is a second function that describes how the dryout property depends on the axial power profile of the nuclear fuel arrangement. The first and the second functions describe said flow dependence and said axial power profile dependence independently of each other. The invention also concerns a nuclear energy plant, a computer program product (23) and a method of operating a nuclear energy plant.

Abstract: An operation method and apparatus in a pressure vessel of a nuclear reactor is provided. The operation apparatus including a body and a guide inserted from the upper side of the nuclear reactor to an interior of the jet pump. The operation apparatus circulate water inside the pressure vessel. The guide is positioned at the end of a body of the operation apparatus, and is inclined with respect to the center axis of the body so as to be inserted into a side opening of the jet pump.

Abstract: An adjustable speed drive system for a primary loop recirculation pump and method of control in which a first breaker is connected to an auxiliary power supply system, an inverter is connected to the first breaker, a second breaker is connected to an output of the inverter, a gate driver supplies a gate pulse to the inverter, and an inverter control portion drives the gate driver. The adjustable speed drive system is combined with a backup adjustable speed drive system. When a malfunction in a main adjustable speed drive circuit portion including the inverter and the gate driver in the adjustable speed drive system used in a normal operation is detected, the main adjustable speed drive circuit portion is switched to the backup adjustable speed drive system, and an operation of a recirculation pump motor for driving a recirculation pump that controls a flow of a reactor core is continued.

Abstract: A containment vessel of a boiling-water nuclear power plant and a method of operating a condenser in a nuclear power plant, include a drain pipe which connects a top region of the containment vessel to a condensing chamber disposed in the containment vessel. The drain pipe draws off noncondensible gases from the surroundings of a building condenser in the containment vessel and thus maintains reliability of performance of the building condenser. The noncondensible gases flow automatically into the condensing chamber through the drain pipe. As a result, the building condenser may have a simple and cost-effective structure.

Abstract: A device is provided for mitigating vibration in a component of a nuclear reactor by removing vibration energy. To reduce vibration in the component, a device operatively connected to the component and including a magnet may be actuated within a conductive cylinder. This actuation may generate one or more eddy currents providing a damping function for removing vibration energy from the component, so as to alter vibration characteristics of the component.

Abstract: An assembly includes a base grid configured to be disposed below a pressure vessel and spaced vertically above a floor of a containment vessel to define a sump therebetween. The assembly further includes an annular wall extending vertically upwards from the floor and laterally bounding the base grid and the sump. The wall separates the sump from a suppression pool, an inlet passage extending through the wall and providing flow communication between the sump and the suppression pool, and an outlet passage extending through the wall and providing flow communication between the sump and the suppression pool.

Abstract: A system includes a containment vessel configured to prohibit a release of a coolant, and a reactor vessel mounted inside the containment vessel. An outer surface of the reactor vessel is exposed to below atmospheric pressure, wherein substantially all gases are evacuated from within the containment vessel.

Abstract: A constrained feedwater sparger end bracket assembly comprises a boiling water reactor vessel, a feedwater sparger including a conduit within the vessel, a feedwater sparger end bracket assembly connected to an end of the conduit at a sparger/bracket junction, and a clamp secured on the feedwater sparger end bracket assembly. The clamp constrains the feedwater sparger end bracket assembly in directions horizontal, vertical and radial to the vessel.

Abstract: An adjustable speed drive system for a primary loop recirculation pump capable of keeping a PLR pump operating even if a single malfunction occurred in a main adjustable speed drive circuit portion. The drive system for the primary loop recirculation pump has a backup main adjustable speed drive circuit portion, and a breaker for switching between an incoming destination and an outputting destination of the backup main adjustable speed drive circuit portion. A control signal of an inverter control portion is switched to control an inverter included in the backup main adjustable speed drive circuit portion, and when a malfunction occurred in the adjustable speed drive main circuit, by switching to the backup main adjustable speed drive circuit portion, the primary loop recirculation pump of a nuclear reactor is continuously controlled and operated.

Abstract: A protection system for a BWR may include a power-dependent high reactor pressure setpoint. The high reactor pressure setpoint that corresponds to at least one value of percent power in an operating domain of the reactor may be less than the high reactor pressure setpoint that corresponds to 100% power. A method of operating the BWR may include implementing the power-dependent high reactor pressure setpoint. The protection system may include a first high reactor pressure setpoint that corresponds to 100% power and at least one second high reactor pressure setpoint that corresponds to one or more values of percent power in the operating domain. The at least one second high reactor pressure setpoint may be less than the first high reactor pressure setpoint. The method of operating may include implementing the first high reactor pressure setpoint and the at least one second high reactor pressure setpoint.

Abstract: A control rod driving simulator has a control rod driving unit having a plurality of coil members vertically mounted therein, a control rod mounted vertically movably through the centers of the coil members, a cart having a reserve tank storing cooling water, a cooling fan for air-cooling the coil members and a control unit for controlling an operation of the cooling fan, a box containing a part of the control rod and a checking unit for checking the vertical movement of the control rod. When the simulator is connected to an actual control rod driving mechanism and a wave profile of current flowing in the coil members of the control rod driving unit is measured, a same wave profile of current as when it is connected to an actual control rod driving mechanism provided to a reactor core can be obtained.

Abstract: An apparatus and method are provided which are designed to structurally replace welds that attach a riser brace assembly to a jet pump riser pipe. The riser brace assembly may include an upper clamp assembly, a lower clamp assembly, and a plurality of mechanical fasteners to provide clamping forces to the upper clamp assembly and the lower clamp assembly. The upper clamp assembly and the lower clamp assembly sandwich a riser brace support, wherein the riser brace support is provided with at least one through-hole to accommodate the plurality of mechanical fasteners.

Abstract: A method for mitigating stress corrosion cracking of a component exposed to a high temperature water in a high temperature water system is provided. The method comprises the steps of lowering corrosion potential conditions to a desired low corrosion potential in the high temperature water environment; and introducing a first material comprising zinc into the high temperature water environment, such that the desired low corrosion potential facilitates transport of the first material into cracks in a structure communicative with the high temperature water environment.

Abstract: The invention concerns a device for spraying water in a feedwater tank of a thermal power plant. Because the spraying device comprises a catalyst for the conversion of the gases contained in the water, the gases released from the water during the spray degasification can be removed easily, reliably and completely. Thus, the degree of efficiency of the thermal circuit of the power plant and the lifespan of the components arranged in the circuit are increased.

Abstract: A variable-voltage variable-frequency power source for an electric motor that drives a re-circulation pump for a boiling water nuclear reactor. The power source has a semiconductor electric power converter and a speed controller for controlling the semiconductor electric power converter. When a part of the semiconductor electric power converter comes into an inoperative state, that part is electrically disconnected, and the variable-voltage variable-frequency power source temporarily stops outputting power, thereby idling the re-circulation pump. Thereafter, the power source re-starts outputting power before the re-circulation pump completely stops. Thus, the re-circulation pump keeps operating, without stopping.

Abstract: A remote monitoring system including a sub-system, local to a nuclear power plant and a remote sub-system at an operations base. Process data is collected manually by a hand-held computer and automatically by instrumentation on the power plant. The data collected is stored on a storage device before being transmitted via a communication link to a remote computer. The remote computer runs data analysis and diagnostic simulations on the process data from the power plant to predict future events.

Abstract: An integrated head assembly (100) is disclosed for a nuclear reactor. The preferred integrated head assembly includes a lift assembly (150) that supports the reactor vessel closure head (90) and integrated head assembly for removal, a separate support structure (202) supported by a ring beam (151) that sets atop the reactor vessel closure head, a shroud assembly (200), a seismic support system (300), a baffle assembly (500), a missile shield (400), and a CRDM cooling system. The CRDM cooling system draws cooling air into the baffle assembly, downwardly past the CRDMs (96), outwardly to upright air ducts (600), upwardly to an upper plenum (680), and out of the assembly through the air fans (190). In a second embodiment the integrated head assembly (1100) includes a missile shield (1400) and CRDM cooling system (1600) that permits access to individual CRDMs from above.

Abstract: A method of predicting component crack behavior in a nuclear reactor provides access to a crack growth behavior model over a global network. A user inputs water chemistry and material characteristics of a particular nuclear reactor and can perform crack growth rate predictions by accessing the behavior model over the computer network. A crack growth prediction profile or crack growth based result is output over the network according to an analysis by the behavior model.

Abstract: In order to allow a reliable and passive external cooling of a reactor pressure vessel of a boiling water reactor, outside the reactor pressure vessel a differential-pressure measurement pipe is provided and has a first and a second pipe leg which are connected to one another above a critical filling level. The first pipe leg is connected at the height of the critical filling level to the interior of the reactor pressure vessel. In a normal operating state, the differential-pressure measurement pipe is filled completely with water. In the event of the undershooting of the critical filling level, the water flows automatically out of the first pipe leg. As a result, the differential pressure between the two pipe legs rises, the differential pressure being detected by a measurement device via which the flooding of the exterior is then brought about.

Abstract: Control device for a cooling circuit which enables a pressure tank of a reactor to be cooled even when a serious failure occurs, without having to resort to the use of external energy. One or more rods is disposed below a bottom cap of the reactor pressure tank. Each rod is connected to an emergency cooling system by a transmission system. The end of each rod opposite the transmission system is arranged at a distance of approximately 3 cm from the bottom cap. A hydraulic system is used advantageously as a transmission system. Each rod is placed vertically on a hydraulic cylinder arranged in a parallel position with respect to the axis of the reactor pressure tank, and each cylinder is connected to the emergency cooling system. When several hydraulic cylinders are used, they are connected to each other by hydraulic lines.

Abstract: A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

Abstract: A method and system for thermal-dynamic modeling and performance evaluation of a nuclear Boiling Water Reactor (BWR) core design is presented. Based on a predetermined generic transient bias and uncertainty distribution in the change in critical power ratio (&Dgr;CPR/ICPR), shifted histograms of fuel rod critical power ratio (CPR) are generated. Ultimately, the operating limit minimum critical power ratio (OLMCPR) of the reactor is directly evaluated from a shifted histogram of fuel rod CPRs for a particular set of initial conditions which result in a probability calculation representing the number of fuel rods subject to a boiling transition (NRSBT) during the transient condition being equal to a predetermined value.

Abstract: A nuclear reactor containment cooling system includes a containment vessel having a drywell and a wetwell, a cooling condenser submerged in a cooling pool of water located outside the containment vessel, a vent line extending from the condenser to a suppression pool disposed in the wetwell, and at least one drain line extending from the condenser to a condensate drain tank located in the drywell. An end of the drain line is vertically submerged below the surface of a pool of water in the drain tank. To enhance flow, a blower can be located in the drain line. The containment cooling system can include a drywell gas recirculation subsystem coupled to the vent line, and including a suction pipe coupled to the vent line, at least one valve located in the suction pipe, at least one blower coupled to the suction line, and a discharge pipe in flow communication with the drywell.

Abstract: A stud enclosure for protecting a stud extending upwardly from a nuclear reactor pressure vessel (RPV) flange has a cylindrical can with a capped end and an open end. The capped end has an axially extending hole with a screw extending therein for fastening the stud enclosure to the stud. A seal ring is disposed adjacent the open end of the cylindrical can for sealing the ring on the RPV flange. A gas valve is disposed in the capped end of the cylindrical can for pressurizing the interior portion of the can with air.

Abstract: An apparatus for removing flammable gas is started only when a concentration change in the flammable gas in a primary containment vessel is detected, so that the decrease in activity of a catalyst due to a catalyst poison is prevented. This apparatus has a catalytic recombining apparatus equipped with a catalyst 1 for reacting the flammable gas, a case 2 that houses the catalyst, and lids 4a and 4b installed at openings 3a and 3b of the case 2. Adhesive parts 5a and 5b which adhere the lids 4a and 4b and the case 2 is melted at high temperatures, and heating elements 6a and 6b composed of an oxidizing catalyst of hydrogen is installed in contact with the adhesive parts 5a and 5b. Thus, the lids are not only closed under normal conditions but opened when the adhesive parts 5a and 5b being melted along with the temperature increase in the surrounding atmosphere.

Abstract: In a nuclear reactor, in-core stability is improved, power density is increased, and an economical natural-circulation reactor (or partial forced-circulation reactor) is achieved. The reactor core has a void reactivity coefficient between −0.07 and −0.03% &Dgr;k/k/% void fraction. This void reactivity coefficient range is achieved by, for example, the design of the by-pass portion and channel box, the enrichment distribution along the axial direction, the provision of blanket areas, and/or the arrangement of water rods and fuel rods within a channel box.

Abstract: A preventive maintenance method and apparatus for a structural member in a reactor pressure vessel according to the present invention reduce a tensile residual stress on a surface of the structural member by impinging a water jet from a nozzle onto a plane surface of a deflector to thereby change direction of flow of the water jet, and impinging the water jet after being deflected onto the surface of the structural member. This method and apparatus are applicable to a narrow space portion, and can improve a residual stress on the surface of the structural member and can also prevent damage such as stress corrosion cracking.

Abstract: A sealed operator is provided for operating an internal locking control element for a closure head for the inner container in a dual container hazardous waste storage system. The operator includes a cylindrical housing in which a bushing/operator element assembly is mounted for translational movement. The latter assembly includes a cylindrical bushing sealingly mounted in the housing, a spherical bearing assembly including an annular outer bearing press fit within the bushing and an annular inner bearing. An operator element is press fit within the inner surface of the inner bearing. The operator element includes a driving head which engages the locking control element for the closure head of the inner container.

Abstract: A blow-off device for live steam includes a filter for radioactive substances in order to increase operational safety in a nuclear power plant. The filter is preferably integrated in a housing of a muffler or sound absorber and is disposed on a slide which is movable along an axis under the effect of gravitational force and in particular also serves as a closure for an outlet opening of the muffler or sound absorber. A method of blowing off live steam is also provided.

Abstract: A nuclear power plant includes at least two reactor vessels enclosed within a single containment building. In a first embodiment, the reactor vessels and steam generators operate separately from one another. In a second embodiment the reactor vessels jointly provide heat to a common header having a plurality of steam generators associated therewith. In a third embodiment, the reactor vessels are completely integrated with one another by means of a single control and coolant system.

Abstract: A passive pulse generator can be connected to a line in order to activate a safety fitting in a nuclear power plant. The line can be closed by a closure element. In the event of a temperature change in the vicinity of the pulse generator, the closure element is actuated solely as a result of a thermal volume change of an actuating element, so that the line is released. The pulse generator therefore activates the fitting passively, without additional control pulses. A method for activating a fitting with a passive pulse generator is also provided.

Abstract: A steam power station, in particular a nuclear power station, includes a steam conduit leading through a wall and forming a fixed point with the wall for the introduction of forces and moments. A main valve is connected to the steam conduit at the fixed point, without a high-pressure pipe being interposed. Satellite valves, which have smaller nominal widths than the main valve, are fastened to the housing of the main valve, without a high-pressure pipe being interposed. At least one additional valve has a housing fastened to the housing of a satellite valve, without a high-pressure pipe being interposed and without any support.

Abstract: A restraint assembly which is a bolted lattice structure for preventing the in-core guide tubes from experiencing excessive flow induced vibrations is described. In one embodiment, the restraint assembly includes upper and lower restraint brackets secured to an in-core housing and in-core guide tube assembly. The restraint brackets are located approximately equal distance from the end supports of the in-core guide tube and in-core housing assembly. The upper restraint bracket is welded between sections of the in-core guide tube, and the lower restraint bracket is welded between the in-core guide tube and the in-core housing. Each bracket includes four flanges, and each flange has tapped, or threaded, openings therein. Each bracket also includes cylindrical sections extending from the flanges, and in one embodiment, the cylindrical sections have inner and outer diameters which match the inner and outer diameters of the in-core housing and the in-core guide tube.

Abstract: A method for determining the susceptibility of a sample of Zircaloy alloy to nodular corrosion. A specimen of such Zircaloy sample is annealed at a temperature within a temperature range bounded at its upper limit by the temperature T.sub.c being the temperature at equilibrium wherein sufficient solute would exist in the (.alpha.-matrix of the particular zircaloy to resist nodular corrosion, and bounded by a lower temperature, such temperature being the temperature of the (.alpha.+precipitate)/(.alpha.+.beta.+precipitate) transus for the particular Zircaloy sample. For Zircaloy-2, such temperature range is from approximately 825.degree. C. to 841.degree. C. The specimen is maintained at such selected temperature within such temperature range for a measured period, and subsequently exposed to steam at a fixed temperature and pressure for a fixed time.

Abstract: A reactor container is composed of a dry well and a wet well and, in the reactor container, there is provided with a dry well cooling system, the dry well cooling system including, in one aspect, an in-dry-well heat exchanger disposed in the dry well of the reactor container, an in-dry-well blower connected to a primary side of the in-dry-well heat exchanger, a circulation pipe connected to a secondary side of the in-dry-well heat exchanger, a normal cooling system connected to the secondary side of the in-dry-well heat exchanger through the circulation pipe and including an equipment cooling pump, an equipment cooling heat exchanger and a seawater pump, and a standby cooling system connected to and branched from the circulation pipe.

Abstract: A maintenance cover for use in a nuclear reactor including a pump deck having at least one pump diffuser opening is described. The cover includes, in one embodiment, a substantially cylindrical member having a first end and a second end. The first end has a diameter larger than the diameter of the pump diffuser opening. The substantially cylindrical member includes a substantially cylindrical frame having a plurality of window openings formed therein. Grating is secured to the frame and extends over the window openings. The cover further includes an end member secured to and positioned over the second end of the substantially cylindrical member. The end member also includes grating. A handle is secured to the substantially cylindrical member.

Abstract: The invention relates to a drive system for internal recirculation of coolant in a nuclear reactor (1), wherein the drive system for a main recirculation pump (10) for the coolant recirculation comprises a pump motor (11) connected to the main recirculation pump (10) and a static frequency converter (F) for connection to an alternating voltage network (16) and for feeding the pump motor (11). The frequency converter (F) comprises a rectifier (12) for connection between the alternating voltage network (16) and a d.c. intermediate link (15) and a first inverter (13) for connection between the d.c. intermediate link (15) and the pump motor (11). The invention is characterized in that the energy storage magazine, for example a flywheel (18), is connected to the d.c. intermediate link (15) via a second inverter (14).

Abstract: A reactor circulating pump system comprises a pump and a power supply apparatus for driving the pump. In the power supply apparatus, an AC/DC converter and a DC/AC converter provided with an AC voltage/frequency adjuster are connected in series between an in-plant AC bus and a reactor circulating pump, a DC voltage adjuster, a superconducting energy storage apparatus and an excitation power supply source are connected to the input side of the DC/AC converter. A controller is incorporated in the power supply apparatus so as to receive a signal indicating instantaneous power failure or power breakdown on the in-plant AC bus, or reactor trip, and change an AC voltage supplied to the reactor circulating pump with a ratio of AC voltage and frequency kept constant for controlling a pump delivery rate to have a predetermined value, while holding an output voltage from the superconducting energy storage apparatus at a predetermined value.

Abstract: An improved stainless steel composition and an improved stainless Ni-based alloy are disclosed for use in nuclear reactor environments. The improved stainless alloys include a stainless steel and a stainless nickel based alloy that are particularly well adapted for use in reactor components that are positioned adjacent to components formed from a zirconium based metal. The improved stainless steel is an austenitic stainless steel material that include less than approximately 0.2 percent manganese by weight, but does include sufficient austenitic stabilizer to prevent a martensitic transformation during fabrication or use in a reactor environment. In one preferred embodiment, the stainless alloy is formed without any significant amount of manganese.

Abstract: A device for mixing two fluids having different temperatures comprises a connecting branch (6) extending into a main pipe (1) from a secondary pipe (2), said connecting branch having at its end positioned in the main pipe a distribution casing (7) with double annular walls, of which an inner wall (9) defines a channel (10) extending axially and centrally in the main pipe (1), and having apertures which allow a fluid (B) to be conducted into and mixed with the fluid (A) passing through the main pipe. The apertures consist of a plurality of small apertures (11) which are formed in the inner wall (9) of the distribution casing (7) and provide intermixing of the two fluids directly in the channel (10) positioned centrally in the main pipe (1).

Abstract: A nuclear internal pump is disclosed in which a flywheel and a thrust disc acting as an auxiliary impeller are fixed on a motor shaft. Fixed orifices X and Y are formed respectively between outer circumference section at larger and smaller diameter sides of the thrust disc and larger and smaller diameter rings disposed to face the outer circumference section respectively, thereby to form a pressurized chamber between the two fixed orifices X and Y. The thrust disc is equipped with an axial thrust reducing device which has two sets of channels for circulating motor cooling water and for boosting the pressure in the pressurized chamber.

Abstract: A secondary-side residual-heat removal system for a pressurized-water reactor removes residual heat through a secondary side of a steam generator in certain modes of operation and malfunction. For this purpose, the respective steam generator has a connection on a live-steam and feed-water side to a secondary side of a safety condenser. The connection on the feed-water side is led to a supply connection of a swirl or vortex chamber valve having a control connection which is connected to a feed-water line part on the pressure side, with a downgradient from a main feed-water pump, and the swirl or vortex chamber valve has an outlet which opens out into a feed-water inflow space of the steam generator. A tertiary side of the safety condenser is fed from a water reservoir disposed at a geodetically higher level. The steam being generated is discharged into the atmosphere through a blowing-off line, with a separating device if appropriate.

Abstract: The Pressurized Water Reactor (PWR) plant steam generators, which are normally located inside the primary containment, are located outside, in separate structures. This arrangement provides for the removal or repair of the individual "leaky" steam generators, while the others remain operational. It also significantly reduces the size and thereby the cost of the containment.

Abstract: A shutdown cooling system for a nuclear reactor operates during lapse of normal power and emergency power and has an independent power source fo removing residual heat while cooling the seals of the main reactor coolant pump. A high pressure pump delivers cooling water to the reactor cooling pump seals, a low pressure pump circulates core coolant, and a cooling mechanism discharges the decay heat. Electrical power for the pumps and associated valves and controls is provided by a dedicated power source apart from the regular residual heat removal apparatus of the reactor, and apart from the emergency generators provided for regular power failure backup.

Abstract: A system for containment of a supercritical water oxidation reactor in the event of a rupture of the reactor. The system includes a containment for housing the reaction vessel and a communicating chamber for holding a volume of coolant, such as water. The coolant is recirculated and sprayed to entrain and cool any reactants that might have escaped from the reaction vessel. Baffles at the entrance to the chamber prevent the sprayed coolant from contacting the reaction vessel. An impact-absorbing layer is positioned between the vessel and the containment to at least partially absorb momentum of any fragments propelled by the rupturing vessel. Remote, quick-disconnecting fittings exterior to the containment, in cooperation with shut-off valves, enable the vessel to be isolated and the system safely taken off-line. Normally-closed orifices throughout the containment and chamber enable decontamination of interior surfaces when necessary.

Abstract: A boiling water reactor control rod drive housing support bars having wing assemblies to limit the travel of a control rod in the event that a control rod housing is ruptured. The wing assemblies are placed on support bars located in rows on opposing sides of the control rod drives. Each wing on the wing assemblies supports a quadrant of a control rod drive. The wing assemblies may be rotated from an operation position to a service position to allow replacement of the control rod drives.

Abstract: A boiling water reactor control rod drive housing support bars having radiation shields to limit the travel of a control rod in the event that a control rod housing is ruptured and protect persons working under the reactor vessel from radiation exposure. The radiation shield includes radiation shield cups for surrounding the bottom portion of a control rod drive. The radiation shield cups are placed on support bars located in rows on opposing sides of the control rod drives. The radiation shield may be easily raised and lowered automatically by the control rod drive installation/removal machine.

Abstract: A device for restricting the motion of a thimble tube in a nuclear reactor during normal reactor operating conditions. A cylindrical sleeve open at each end has upper, middle, and lower sections. The upper section has an over-expanded area adjacent the top end to aid in proper positioning in the lower core support plate. The middle section is formed from inner and outer sections attached at their ends and formed from material having different coefficients of thermal expansion. The inner portion of the middle section moves from a first normal relaxed state during cold conditions in the reactor to a second flexed state during hot conditions in the reactor to provide a clamping action on the thimble tube received by the sleeve. A baffle in the lower section reduces coolant flow velocity and pressure through the sleeve.

Abstract: Apparatus suitable for annealing the interior walls of a nuclear reactor which comprises a nuclear reactor shell, a cap securely attached to the nuclear reactor shell, wherein the cap has an inlet line for the introduction of a hot gas under high pressure and an outlet line for removing low pressure gas from the interior of the apparatus, means, such as a cylindrical shell, that helps define an annular space adjacent to the vertical walls of the reactor through which gas can flow and means for directing low pressure gas to the outlet line.

Abstract: A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.