Abstract: Chimneys include several combinable parts useable in nuclear reactors. The parts are modular and removably joinable without destruction for use in directing flow in an operating nuclear reactor and directly fit in storage area during non-operation. Chimney parts are joinable through flanges and connecting structures. Chimney parts may include partitions that divide or direct energetic coolant flow from a nuclear core as well as steam separating and drying structures. The parts each individually fit within storage areas of the nuclear plant, including equipment or buffer pools in the refueling floor of the plant. Methods move the chimney parts between the reactor and storage areas, and multiple parts may be stacked or nested in such moves. Methods are useable underwater and with storage pools to prevent exposure of chimney parts during an outage. During operation, chimneys are useable in place of existing single-piece chimneys.

Abstract: Chimneys include several combinable parts usable in nuclear reactors. The parts are modular and removably joinable without destruction for use in directing flow in an operating nuclear reactor and directly fit in storage area during non-operation. Chimney parts are joinable through flanges and connecting structures. Chimney parts may include partitions that divide or direct energetic coolant flow from a nuclear core as well as steam separating and drying structures. The parts each individually fit within storage areas of the nuclear plant, including equipment or buffer pools in the refueling floor of the plant. Methods move the chimney parts between the reactor and storage areas, and multiple parts may be stacked or nested in such moves. Methods are usable underwater and with storage pools to prevent exposure of chimney parts during an outage. During operation, chimneys are usable in place of existing single-piece chimneys.

Abstract: A plant with piping mounted on branch pipe, wherein said piping which introduces gas has a nozzle portion as a joint portion to a vessel and a branch portion connected with a branch pipe; and wherein an enlarged passage portion is formed at least at one of said branch portion and said nozzle portion, and a passage sectional area of said enlarged passage portion is larger than that of said piping other than said enlarged passage portion. Since the flow velocity of the gas flowing inside slows down at the enlarged passage portion, the occurrence of acoustic resonance at the branch portion or the nozzle portion can be suppressed. Accordingly, the fluctuation pressure of the gas flowing in the piping can be reduced even more.

Abstract: A jet pump which can restrain self-excited vibration in a connection portion between an inlet mixer pipe and a diffuser pipe without inhibiting a structural deformation due to thermal expansion and the like includes a slip joint structure connecting the inlet mixer pipe and the diffuser pipe to each other by inserting the inlet mixer pipe into an upper end opening of the diffuser pipe with a clearance left therebetween; and a self vibration damping structure configured such that when the clearance defined by an outer pipe wall of the inlet mixer pipe and an inner pipe wall of the diffuser pipe is widening or narrowing due to vibration of the inlet mixer pipe or the diffuser pipe, a flow path resistance inside a clearance flow path for pumped coolant water defined by the clearance is not smaller than a fluid inertia force all over the clearance flow path.

Abstract: In the boiling water nuclear plant of the present invention, a steam dryer is disposed in a reactor pressure vessel. Materials that have capability of capturing nitrogen compounds containing N-16 are supported on porous member. The porous members are placed in a region where steam goes through in the steam dryer. For example, both or either of perforated plates installed in the steam dryer is constituted of the porous member on which N-16 capture material is supported. When steam containing N-16 goes through the perforated plates, the N-16 is captured by the porous member, whereby the N-16 transfer amount into the turbine system is reduced.

Abstract: A jet pump has a plurality of nozzles installed to a nozzle base, a throat and a diffuser. A first nozzle straight-tube portion, a first nozzle narrowing portion, a second nozzle straight-tube portion, a second nozzle narrowing portion, and a nozzle lower end portion formed in those nozzles are disposed in this order from the nozzle base to a ejection outlet. A narrowing angle of the second nozzle narrowing portion is larger than of the first nozzle narrowing portion. The jet pump forms, in a lower end portion of the throat, a flow passage narrowing portion having a flow passage cross-sectional area that gradually diminishes. This flow passage narrowing portion is inserted into an upper end portion of the diffuser.

Abstract: Hard stops are useable in an operating nuclear reactor to separate and bias restrainer brackets and inlet mixers. Hard stops include a lip clamp that clamps to a restrainer bracket and a wedge member that biases against the inlet mixer. The wedge member and lip clamp are engaged such that the two components can slide against one another to bias the restrainer bracket and inlet mixer. The lip clamp includes a clamp arm and an engagement member to clamp opposite sides of the restrainer bracket. Ratchet assemblies maintain selective positioning various components of the hard stops. Hard stops may be used in several different numbers, positions, and configurations in repair or modification systems. Hard stops may be installed by determining location on an outside of a restrainer bracket for the hard stop, securing the hard stop at the location, and biasing the hard stop between two components at the location.

Abstract: A natural-circulation type boiling water reactor includes a plurality of divided chimneys provided above a reactor core and a number of fuel assemblies are charged in the reactor core. The natural-circulation type boiling water reactor is provided with a pressure equalization structure arranged on rectangular-columnar lattice plates of the divided chimneys for equalizing pressures in divided chimney portions so as to equalize the pressures of the divided chimneys with the pressure equalization structure.

Abstract: A support clamp assembly for mechanically securing a thermal sleeve to an elbow conduit in a jet pump of a nuclear reactor vessel, the support clamp assembly including: a tension shaft having a first end extendable through an opening in a sidewall of the elbow conduit and an opposite end with a head; a cruciform assembly having a base with an opening to receive the tension shaft and to abut the head of the shaft, wherein the cruciform assembly seats in the thermal sleeve; a boss slidable over the first end of the tension shaft and having a curved surface seating an outside surface of the elbow conduit, and a coupling device engaging the first end of the tension shaft and abutting the boss, wherein the coupling device places the tension shaft under tension to secure the cruciform assembly to the thermal sleeve and the boss to the elbow conduit.

Abstract: A boiling water reactor has a core disposed in the reactor pressure vessel and loaded with a plurality of fuel assemblies including transuranic nuclides. A ratio of Pu-239 in all of the transuranic nuclides included in the fuel assembly, which is loaded in the core, with a burnup of 0 is 3% or more but 45% or less. In the fuel assembly having a channel box and a plurality of fuel rods disposed in the channel box, a transverse cross section of a fuel pellet in the fuel rod occupies 30% or more but 55% or less of a transverse cross section of a unit fuel rod lattice in the channel box.

Abstract: A jet pump sensing line support clamp may be used for sensing line repair, replacement, and damage prevention or reduction. The clamp may affix to jet pump sensing line supports and confine the individual jet pump sensing lines. The clamp may provide for further access or securing of the lines in the support through the clamp. Methods of installing the clamp may include attaching and tightening the clamp against the sensing lines.

Abstract: A nozzle apparatus of a jet pump includes a nozzle base member, and a plurality of nozzles installed to the nozzle base member and forming a plurality of narrowing portions, in which a fluid passage cross-sectional area of a driving fluid passage formed in the nozzle is reduced.

Abstract: An embodiment of the present invention provides an apparatus and system for repairing and/or maintaining a position of a first component in relation to a second component. The first and second components may be located within a reactor pressure vessel of a nuclear powerplant. The apparatus and system may attach at least one bearing plate to a horizontal surface of the first component. The apparatus and system includes structure to apply a pre-load to the first component. This may assist in maintaining the position of the first component relative to the second component.

Abstract: An embodiment of the present invention takes the form of an apparatus or system that may reduce the level of vibration experienced by an inlet riser or other similar object within a reactor pressure vessel. An embodiment of the present invention may eliminate the need for welding the riser brace to the inlet riser. An embodiment of the present invention provides at least one riser brace clamp that generally clamps the riser brace to the inlet riser. After installation, the riser brace clamp may lower the amplitude of, and/or change the frequency of, the vibration experienced by the inlet riser.

Abstract: A device and method for stabilizing a dryer assembly in a reactor pressure vessel of a nuclear reactor including a spring assembly dimensioned and positioned within the reactor pressure vessel for applying a stabilizing force to the dryer assembly relative to the reactor pressure vessel.

Abstract: A high-temperature gas-cooled reactor steam generating system comprises a plurality of nuclear steam supply systems, a high-pressure cylinder (21), a low-pressure cylinder (22), a condenser (23), a condensate pump (24), a low-pressure heater (25), a deaerator (26), a water supply pump (27), and a high-pressure heater (28) which are sequentially connected end to end to form a close steam loop. On one hand, the inherent safety of the reactor is guaranteed and the generating system is simplified with the inherent safety. On the other hand, the scale economy of the steam engine system and other systems of a whole power station is guaranteed through batch copy, a shared auxiliary system and a scale effect.

Abstract: Example embodiment clamps may be used to clamp two components together, with a degree of freedom of motion between the clamped components, including both rotational and translational motion. Example embodiment clamps may clamp a BWR jet pump sensing line to a diffuser as a repair or installation of a sensing line support. Example clamps may include a ball subassembly that holds a component and allows rotation and twisting of the component within the clamp. Example clamps may further include jaws holding the ball subassembly and a biasing element that permits tightening of the clamp.

Abstract: Modified Kalina and modified Mayahi cycle heat engines are disclosed. Improvements in efficiency may be gained through various changes to these cycles as well as combining these cycles with boiling water reactors and other Rankine cycle power plants.

Abstract: A steam separator comprises an outer main swirler and an inner auxiliary swirler which is smaller than the main swirler. The swirlers are provided so as to be concentric on the inner wall at the lower side of the first stage inner cylinder. In the steam separator, when the gas-liquid two-phase flow which flows in the vicinity of the axial center of the first stage inner cylinder passes the auxiliary swirler, it is separated into steam and water by the centrifugal force. The separated water (droplets) is introduced into the main swirler. When the separated water (droplets) passes the main swirler, it is separated at the inner wall side of the first stage inner cylinder by the centrifugal force. Pressure loss in a steam separator is reduced and steam separation capability is increased without increasing the moisture from the steam separator.

Abstract: The present invention provides a nuclear fuel comprising an actinide nitride such as 233U, 234U, 235U, 236U, 238U, 232Th, 239Pu, 240Pu, 241Pu, 242Pu, 244Pu, 239Np, 239Am, 240Am, 241Am, 242Am, 243Am, 244Am, 245Am, 240Cm, 241Cm, 242Cm, 243Cm, 244Cm, 245Cm, 246Cm, 247Cm, 248Cm, 249Cm, 259Cm, 245Bk, 246Bk, 247Bk, 248Bk, 249Bk, 250Bk, 248Cf, 249Cf, 250Cf, 251Cf, 252Cf, 253Cf, 254Cf, 255Cf, 249Es, 250Es, 251Es, 252Es, 253Es, 254Es, 255Es, 251Fm, 252Fm, 253Fm, 254Fm, 255Fm, 256Fm, 257Fm, 255Md, 256Md, 257Md, 258Md, 259Md, 260Md, 253No, 254No, 255No, 256No, 257No, 258No and 259No, and optionally fission products such as 97Tc, 98Tc and 99Tc, suitable for use in nuclear reactors, including those based substantially on thermal fission, such as light and heavy water reactors, gas-cooled nuclear reactors, liquid metal fast breeders or molten salt fast breeders. The fuel contains nitrogen which has been isotopically enriched to at least about 50% 15N, most preferably above 95%.

Abstract: A device and method for stabilizing a dryer assembly in a reactor pressure vessel of a nuclear reactor including a spring assembly dimensioned and positioned within the reactor pressure vessel for applying a stabilizing force to the dryer assembly relative to the reactor pressure vessel.

Abstract: A pressure vessel of a reactor has a flow space between the reactor core and a separation device for separating water from water steam. The cross-section area of the flow section of the space expands upstream of the separation device. This makes it possible to pre-separate a water-water steam mixture coming from the reactor core. Due to the pre-separation, only the central area is provided with a cyclone device, and only drying equipment is disposed near the cyclone device. The novel approach makes it possible to reduce the height of the pressure vessel of a reactor.

Abstract: A ratio of the number of fuel assemblies loaded on a core to the number of control rod drive mechanisms is 3 or more. The fuel assembly itself contains mixed oxides of a low enrichment concentration uranium oxide containing 3 to 8 wt % in the average enrichment concentration of the fuel assembly, or mixed oxide containing not less than 2 wt %, but less than 6 wt % in the average enrichment concentration of fissile plutonium of. In the burner type BWR core on which the fuel assemblies are loaded, an average weight density of uranium, plutonium and minor actinides is 2.1 to 3.4 kg/L as a conversion at the value of unburned state.

Abstract: A resistance member (e.g., fuel holding portion of the lower tie plate) is provided at the lower end of the fuel assembly. Provision is made of a coolant ascending path in which said water rods have coolant inlet ports that are open in a region lower than the resistance member to upwardly guide the coolant, and a coolant descending path which has a coolant delivery port that is open in a region higher than the resistance member to downwardly guide the coolant. The coolant ascending path and the coolant descending path are communicated with each other at their upper end portions.

Abstract: A main steam system around a nuclear reactor which comprises two main steam nozzles in a reactor pressure vessel, each of the main nozzles being disposed at a symmetrical position with respect to a plane parallel to steam outlet faces of steam dryers provided inside the reactor pressure vessel and passing through a center of the reactor pressure vessel; and main steam pipes each connected to the main steam nozzles.

Abstract: A main steam system around a nuclear reactor which comprises two main steam nozzles in a reactor pressure vessel, each of the main nozzles being disposed at a symmetrical position with respect to a plane parallel to steam outlet faces of steam dryers provided inside the reactor pressure vessel and passing through a center of the reactor pressure vessel; and main steam pipes each connected to the main steam nozzles.

Abstract: A method and system of regulating an output voltage of a boiling water reactor nuclear reactor plant recirculation system motor generator are provided. The method includes sensing an alternator output voltage and transmitting an alternator output voltage signal to a voltage regulator circuit, sensing an alternator speed and transmitting an alternator speed signal to the voltage regulator circuit, comparing the alternator output voltage signal to the alternator speed signal with a volts per hertz divider network electrically coupled to the alternator output voltage sensing circuit and the alternator speed sensing device, adjusting a capacitive reactance of the voltage regulator with a lead compensation circuit electrically coupled in series with the volts per hertz divider network, and adjusting a current in a control winding of a saturable reactor.

Abstract: Single piece fuel element and fast spectrum boiling water reactor using elements of this type.

Abstract: A supercritical pressure water cooled reactor comprising: a reactor vessel including: a shell part for containing sub-critical pressure coolant, and an end part for containing supercritical-pressure coolant which is separated from the sub-critical pressure coolant in the reactor vessel. A core-support plate with through-holes, the core-support plate disposed-in and fixed to the reactor vessel so that the core-support plate divides space inside the reactor vessel into a supercritical-pressure portion and a sub-critical pressure portion. Fuel tubes with both open ends fixed to the through-holes, the open ends being communicated to the supercritical-pressure portion, outside of the fuel tubes being disposed in the sub-critical pressure portion; and nuclear fuel assemblies disposed in the fuel tubes.

Abstract: A supercritical water cooled reactor comprising: a reactor vessel including: a shell part for containing sub-critical pressure coolant, and an end part for containing supercritical-pressure coolant which is separated from the sub-critical pressure coolant in the reactor vessel; a core-support plate with through-holes, the core-support plate disposed in and fixed to the reactor vessel so that the core-support plate divides space inside the reactor vessel into a supercritical-pressure portion and a sub-critical pressure portion; fuel tubes with both open ends fixed to the through-holes, the open ends being communicated to the supercritical-pressure portion, outside of the fuel tubes being disposed in the sub-critical pressure portion; and nuclear fuel assemblies disposed in the fuel tubes.

Abstract: A fuel assembly for a boiling water reactor which is designed to allow water, during operation of the reactor, to flow upwards through the fuel assembly while absorbing heat from a plurality of fuel rods, whereby part of the water is transformed into steam. The fuel assembly comprises a first steam pipe (10a) arranged with its longitudinal axis parallel to the longitudinal axis of the fuel assembly and the steam pipe comprises an inlet for the steam arranged in the first end of the steam pipe and an outlet for the steam arranged in the second end of the steam pipe. The fuel assembly also comprises a second steam pipe (10b) arranged above and at a distance from the first steam pipe such that an opening is formed between the steam pipes. The outlet of the first steam pipe has a diameter which is larger than the diameter of the inlet of the second steam pipe.

Abstract: A method for controlling vessel chemistry in a boiling water reactor (BWR) includes a targeted injection of hydrazine (N2H4) to overcome intergranular stress corrosion cracking (IGSCC) and provide other advantages. The method does not require the injection of hydrogen as a reducing species nor the costly equipment needed to store and control the injection of hydrogen, but it is optional. The method involves: 1) adding a carefully selected amount of N2H4 at a carefully selected location such that reaction with hydrogen peroxide (H2O2) is targeted for reduction prior to treated vessel water (feed water combined with steam dryer/separator liquid effluent) entering the reactor core and 2) providing sufficient residence time to keep all but a tolerable amount of the N2H4 from entering the reactor core.

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: In each of a number of fuel assemblies loaded in a core of a boiling water reactor, a fuel holding portion of a lower tie plate holds lower end portions a plurality of fuel rods and at least one water rod. The water rod includes a rising pipe opened to a space in the lower tie plate below the fuel holding portion and introducing upward a coolant introduced to the rising pipe, and a falling pipe communicated with the rising pipe and introducing downward the coolant introduced through the rising pipe. The falling pipe has a coolant outlet opened to a second coolant passage defined between the fuel rods above the fuel holding portion. The rising passage is filled with the coolant during a period of rated power operation of the reactor, and a surface of the coolant is formed in the rising pipe during a period of non-rated power operation in which a flow rate of the coolant supplied to the fuel assemblies is lower than that during the period of rated power operation.

Abstract: A shroud attachment assembly which provides both structural support for the shroud and a seal between the shroud and shroud support is described. In one specific embodiment, the shroud attachment assembly includes a shroud flange formed integral with the shroud, and a pump deck flange formed integral with the pump deck. The shroud flange extends over the pump deck flange, and a plurality of studs extend through aligned openings in the flanges. In addition, a plurality of wedges are secured to the shroud flange by studs, and the wedges extend between the shroud flange and the pump deck flange. Generally, the studs which extend through aligned openings in the shroud and pump deck flanges transfer vertical loads from the shroud to the pump deck, and the wedges transfer horizontal loads from the shroud to the pump deck.

Abstract: A reactor core for a boiling water nuclear reactor includes a plurality of fuel assemblies (40) with a plurality of vertical fuel rods (10) and possibly occasional vertical water-filled rods or channels (32, 48, 49, 50, 51) which are surrounded by a fuel channel (1). Between the fuel assemblies (40) there are arranged water gaps (37a, 37b). At least one fuel assembly (40) has at least one outer, reduced corner portion (41) facing a gap (37a, 37bb), and the number of fuel rods (10) in each such fuel assembly (40) is reduced by a number corresponding to the number of reduced corner portions (41) therein.

Abstract: A reactor core for a boiling water nuclear reactor includes a plurality of fuel assemblies (40) with a plurality of vertical fuel rods (10) and possibly occasional vertical water-filled rods or channels (32, 48, 49, 50, 51) which are surrounded by a fuel channel (1). Between the fuel assemblies (40) there are arranged water gaps (37a, 37b). At least one fuel assembly (40) has at least one outer, reduced corner portion (41) facing a gap (37a, 37bb), and the number of fuel rods (10) in each such fuel assembly (40) is reduced by a number corresponding to the number of reduced corner portions (41) therein.

Abstract: In a pressure vessel of a nuclear reactor containing a core assembly enclosed within a core shroud, the core shroud spaced radially inwardly of a side wall of the pressure vessel with an annular pump deck located in an annular radial space between the core shroud and the side wall of the pressure vessel, the improvement wherein the shroud is removably secured to an annular support leg extending upwardly from the bottom of the pressure vessel; and further wherein the annular pump deck is provided in the form of a plurality of removable segments.

Abstract: A fuel assembly is provided with a coolant ascending path for making coolant rise and a water rod having a coolant descending path for conducting the coolant.A ratio of a flow area in a coolant inlet port of the smallest in coolant ascending path 13 on the downstream side than large diameter tube portion 3E to a flow area of the largest in the axial direction of coolant ascending path 13 in large diameter tube portion 3E is set to be 0.2-20%.In the normal operation, the declination degree from the liquid level in the coolant ascending path, corresponding to the coolant flow rate of the liquid level formed in the coolant ascending path can be controlled. Further, at the time of the excess the change speed of the liquid level can also be controlled.

Abstract: In a pressure vessel of a nuclear reactor containing a core assembly enclosed within a core shroud, the core shroud spaced radially inwardly of a side wall of the pressure vessel with an annular pump deck located in an annular radial space between the core shroud and the side wall of the pressure vessel, the improvement wherein the shroud is removably secured to an annular support leg extending upwardly from the bottom of the pressure vessel; and further wherein the annular pump deck is provided in the form of a plurality of removable segments.

Abstract: A self-aligning seal system for use in carrying out maintenance services in a nuclear fission reactor plant is disclosed. The self-aligning seal system enables an operator to block off openings in the lower portion of a water filled reactor pressure vessel for maintenance service in an underlying dry field without draining the coolant water from the reactor pressure vessel.

Abstract: A boiling water reactor includes a pressure vessel containing a reactor core, chimney, steam separator assembly, and steam dryer assembly therein, with the vessel being filled with reactor water to a normal water level through the steam separator assembly. A plurality of control rod drives extend downwardly from the bottom of the pressure vessel and are operatively joined to control rods extending downwardly into the reactor core. The chimney includes a plurality of channels disposed above the core and laterally spaced apart to define guide slots for receiving the control rods as they are selectively translated upwardly out of the core by the control rod drives. The chimney has a vertical height for increasing the normal water level above the reactor core and for providing a space for the control rods withdrawn from the reactor core by the bottom-mounted control rod drives.

Abstract: A boiling water reactor comprises a reactor pressure vessel accommodated in a vertical fashion, a core disposed at a low portion in the reactor pressure vessel, a plurality of control rods to be inserted from an upper side of the core into the spaces between adjacent fuel assemblies and withdrawn therefrom upwardly, a shroud surrounding the fuel assemblies so as to define the core and having an upper end opening, a shroud head which closes the upper end opening of the shroud and through which the control rods are inserted or withdrawn, a separator standing upward from the shroud head to carry out gas-water separation of steam generated from the core, a fixing pedestal disposed above the separator, a control rod driving mechanism mounted on the fixing pedestal and adapted to drive the control rods, a drier unit including a plurality of drier elements annularly arranged along an upper inner wall surface of the reactor pressure vessel and adapted to dry the steam passing through the steam flow hole of the fixing

Abstract: A reactor core for a boiling water nuclear reactor comprises a plurality of vertical fuel assemblies (40), each one containing a plurality of fuel rods (10, 10a, 10b) with enriched nuclear fuel material, which are arranged between a bottom tie plate (11) and a top tie plate (12) in a surrounding vertical fuel channel (1). Each fuel assembly is designed with an inlet (3) for water for conducting water in through the bottom tie plate, through the vertical fuel channel, and out through the top tie plate. Further, each fuel assembly is arranged with intermediate gaps (37a, 37b) with respect to adjacent fuel assemblies and possibly with a channel (32, 50) arranged internally in the fuel assembly for conducting water through the gaps and through the internal channel (if any) in a vertical direction from below and upwards through the core.

Abstract: A phase-separator assembly for a boiling-water nuclear reactor includes a steam-separator and a dryer. The steam separator and dryer are mechanically coupled so that relative vertical movement is permitted therebetween. The mechanical coupling limits this relative vertical movement. When installed in a reactor, the steam separator and dryer are independently supported. The dryer is supported so as to define a gap between the dryer and steam separator that optimizes their collective performance. When in storage, the dryer is not independently supported. Instead, the dryer is dropped to a minimum height above the steam separator, providing for compact storage.

Abstract: A recirculation system for driving reactor coolant water in a downcomer within a reactor pressure vessel includes a jet pump disposed in the downcomer and having a suction inlet at the top thereof and an outlet at the bottom thereof. A reactor internal pump is joined directly to the pressure vessel at an access nozzle thereof and includes a housing sealingly joined to the access nozzle, and in serial flow communication an inlet channel, an impeller, an outlet channel, and an injector nozzle. The inlet channel carries coolant water from the pressure vessel to the impeller which pressurizes the coolant water and returns it through the outlet channel and the injector nozzle for injection into the jet pump inlet. The water injected from the internal pump into the jet pump drives the jet pump for effecting recirculation flow within the pressure vessel.

Abstract: A fuel assembly having a water cross or water rod arranged between the fuel rods, an inside of the water cross or water rod being divided into a coolant rising passage and a coolant lowering passage, and a control guide tube disposed inside the water cross or water rod and extending along an axial direction of the water cross or water rod. The coolant rising passage has a coolant inlet port formed to a portion above or under a portion at which the fuel rods are supported by the lower tie plate. The control element guide tube has a coolant outlet port formed at that portion so that a coolant introduced into the coolant rising passage flows vertically upwardly, then turns and flows downwardly along the control element guide tube, and flows into an inside thereof through the coolant outlet port. A reactor core comprises a plurality of fuel assemblies each having a structure described above, a control rod and a control element operated in association with the control rod.

Abstract: An expanded reactor core capacity is achieved within the constraints imposed by the diametric extent of the current advanced boiling reactor (ABWR). This enhanced core capacity is achieved through the utilization of a scalloped shroud structure in conjunction with modified control rod designs. The core diameter is expanded toward the reactor vessel interior wall to establish a minimum distance therebetween representing a minimum value thereof avoiding neutron fluence induced vessel embrittlement. The control rods servicing peripheral fuel assemblies are configured having either a T-shaped blade configuration or an L-shaped blade configuration depending upon the parallel orientations of the peripheral fuel assemblies.

Abstract: A conventional low pressure coolant injection system for boiling water reactors is provided. With the modification, the cross tie conduits and associated valves remain open between two LPCI divisions. On the occasion of an LOCA, the LPCI pumps are activated and injection valves for each of the LPCI divisions are opened to commence coolant injection in the recirculation loops in simultaneous fashion. However, the rate of flow of water coolant within each injection system is controlled by a hydraulic resistance, which is selected to achieve reactor core cooling within requisite time and in requisite quantities from one injection path. Thus, even though coolant water may flow through a rupture within one recirculation loops, sufficient water will be injected in the other loop to achieve core cooling.

Abstract: Fuel management in a boiling-water nuclear reactor involves arranging fuel bundles in upper and lower matrices of the reactor core. During a refueling operation, some bundles in the upper matrix are removed and retired, while fresh bundles are inserted in the lower matrix and some bundles originally in the lower matrix are transferred to the upper matrix. In the transfer, fuel bundles are inverted so that included fuel rods in the lower matrix have their plenums oriented downward, while fuel rods in the upper matrix have their plenums oriented upward. This method provides greater flexibility in repositioning fuel bundles for longer burnups and lower high-level waste. In particular, problems with axial spectral variations in neutron flux can be compensated for using the disclosed refueling procedure.