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 valve control apparatus for generating an anticipatory output response for controlling a fluid flow in a fluid flow line includes a valve provided inside the fluid flow line, at least one sensor for providing at least one input signal, said at least one input signal responsive to properties of said fluid flow in advance of said valve, and a letdown back-pressure controller receiving said at least one input signal and providing a tunable output signal for selectively opening and closing said valve to vary flow in the flow line responsive to said at least one input signal.

Abstract: A method of controlling the system pressure in a power generating system, having a turbine-generator and a BWR, that modulates the core thermal power of the reactor while maintaining the main turbine control valves in a constant steady position is described. The core thermal power may be adjusted by adjusting the control rod density within the reactor core or by adjusting the flow rate through the reactor which may be accomplished by modulating the speed of variable frequency recirculation pumps or by modulating recirculation flow control valves. The method includes transferring the power generation system from normal turbine control valve modulation pressure control to core thermal power modulation pressure control. Additionally the method includes modifying the bypass valve closure bias and the power control bias to accommodate the variances from core power modulation pressure control over normal pressure control.

Abstract: A feedwater control system and method for a pressurized water reactor steam generating system having first and second output signals. The first output signal is determined by first and second input signals, and, when combined with a third input signal automatically controls at least one feedwater pump and first designated valves which regulate water flow from the one or more feedwater pumps to a steam generator when a steam generator steam load and reactor are operating at a first predetermined power level. The first input signal is determined by a downcomer feedwater flow differential pressure. The second input signal is determined by a reactor power level. The third input signal is determined by a steam generator level. The second output signal, determined by a steam generator water level, automatically controls at least one startup feedwater control valve when the steam generator steam load and the reactor are operating at a second predetermined power level.

Abstract: The operator of a nuclear steam supply system manually selects a lineup of either one, two, or three main feedwater pumps for normal reactor operation to generate power. This selection sets or resets a status latch for each pump, representative of intended pump operation.

Abstract: A power supply system for driving reactor coolant recirculation pumps within an advanced boiling water reactor is configured in such a manner that: at least one normal-operation busbar branches off from either a main on-site power line or from a power line for auxiliary power generation that branches off from the main on-site power line, with a house transformer therebetween; at least one MG set is connected to one of these normal-operation busbars; one static adjustable-frequency power supply device is connected to each MG set; and a plurality of recirculation pumps is connected to each static adjustable-frequency power supply device.

Abstract: A chimney which can be reconfigured or removed during refueling to allow vertical removal of the fuel assemblies. The chimney is designed to be collapsed or dismantled. Collapse or dismantlement of the chimney reduces the volume required for chimney storage during the refueling operation. Alternatively, the chimney has movable parts which allow reconfiguration of its structure. In a first configuration suitable for normal reactor operation, the chimney is radially constricted such that the chimney obstructs vertical removal of the fuel assemblies. In a second configuration suitable for refueling or maintenance of the fuel core, the parts of the chimney which obstruct access to the fuel assemblies are moved radially outward to positions whereat access to the fuel assemblies is not obstructed.

Abstract: A natural circulation reactor, especially a boiling water reactor, includes a reactor pressure vessel, a reactor core and a core jacket disposed within the reactor pressure vessel, surrounding the reactor core and having at least one inlet port and one outlet port for core coolant. A flow path for the core coolant, which is formed between the core jacket and the reactor pressure vessel, leads through the inlet port into the core jacket and has a flow cross section. At least one throttle element is disposed in the flow path for varying the flow cross section in order to regulate the core coolant flow rate. The flow cross section can be both increased and reduced in size by varying the cross sectional area of the throttle element, so that a regulation of the power of the natural circulation reactor is carried out in a simple way and, in an incident, a reduction in power to approximately 40% of the instantaneous reactor power is ensured.

Abstract: A reactor core coolant flow rate control system for a BWR type nuclear power plant in which electric power is generated by a turbine-driven generator driven by a steam turbine rotating under the work of a main steam generated from a cooling water recirculated through a boiling water reactor includes an internal pump driving unit comprised of an electric motor driven by an electric power generated by the turbine-driven generator, a fluid coupling for transmitting an output torque of the electric motor and a variable-frequency generator whose rotation number is changed in accordance with the torque transmitted by way of the fluid coupling, a fluid coupling control signal generating unit for generating a fluid coupling control signal for the fluid coupling in response to at least one of deviation in a load and a rotation speed of the turbine-driven generator and deviation in a rotation speed of the internal pumps, and a plurality of internal pumps for recirculating the cooling water.

Abstract: A nuclear reactor, in particular a pressurized water reactor, has a containment, a containment shell surrounding the containment and a concrete construction of a reactor building surrounding the containment shell. A heat dissipation system for the nuclear reactor includes a sump volume disposed in a lower region of the containment shell for receiving coolant. A sump cooler is disposed inside the sump volume, has cooling tubes with a primary side and a secondary side and has feed and return lines. The primary side of the cooling tubes is covered at least when the sump volume is largely filled with coolant. An intermediate cooler has a tertiary side and is connected through the feed and return lines of the sump cooler to the secondary side of the cooling tubes. A heat sink is disposed outside the reactor building and is connected to the intermediate cooler on the tertiary side.

Abstract: A method and an apparatus for in situ ultrasonic inspection of the core spray T-box to thermal sleeve attachment weld and the core spray T-box to cover plate attachment weld in BWRs. The apparatus incorporates a stationary clamping mechanism and sliding and rotational mechanisms which are operated remotely to scan the transducer over the T-box welds and heat-affected zones thereof.

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: A control system for avoiding coupled neutronic-thermal hydraulic instabilities during maneuvering of a boiling water nuclear reactor that determines the average elevation in the reactor core of the coolant boiling boundary, and avoids instabilities by maintaining the core average coolant boiling boundary elevation above a target value, which is dependent on reactor design specifics but may be about four feet, in core power-flow operating regions which are subject to instabilities. The system may be implemented without physical modification of existing reactors or reactor instrumentation and control systems through use of the reactor core monitoring computer, which typically has available all core parameters necessary for the computations required in the invention.

Abstract: The reactor shut-down margin and the thermal margin, as well as the output per unit volume of the core is increased. When A is the total cross sectional area of the non-boiling water areas in the channel boxes each surrounding a fuel assembly and non-boiling water areas outside the channel boxes and B is the cross sectional areas of boiling water areas in the channel boxes, the relation of 1>B/(A+B).gtoreq.0.76 is set. A control unit adjusts a rotational speed of a recirculation internal pump which controls the cooling water flow rate per unit area in the core section to 3.0.times.10.sup.3 t/h/m.sup.2 at the beginning of the operation cycle and controls the cooling water flow rate per the unit area in the core cross section to 3.3.times.10.sup.3 t/h/m.sup.2 at the end of an operation cycle.

Abstract: A fuel assembly (16) for a boiling-water nuclear reactor arranged in groups with four fuel assemblies in each group and with a centrally placed control rod (17) forming a supercell. Each fuel assembly (16) includes a plurality of vertical fuel rods (12) with enriched nuclear fuel material (24), said rods being arranged between a bottom tie plate (13) and a top tie plate (14) in a surrounding vertical fuel channel (1) which is connected to a transition section (2). The transition section is provided with holes (20a-d) for by-pass flow outside the fuel assembly. The fuel assembly is characterized in that the by-pass holes are arranged in such a way that when turning the fuel assembly (16), including the transition section (2), 90.degree., 180.degree. or 270.degree. around the longitudinal axis of the fuel assembly, said holes are adapted to direct the by-pass flow away from the control rod (17).

Abstract: A method and a device for minimizing electrostatically enhanced deposition of charged particulates in the jet pump nozzles of operating BWR plants. The jet pump nozzle has an inner conducting surface which is electrically isolated from the main body of the nozzle. An electrical circuit is electrically coupled to the inner conducting surface of the nozzle. The electrical circuit is designed to electrically minimize charged-particle deposition at the nozzle surface exposed to a free-stream electrical potential, using a DC circuit with active element feedback to adjust the surface potential to minimize the net current to the inner conducting surface, thereby minimizing deposition.

Abstract: A locking piston drive system and valve assembly is provided for shutdown of a nuclear reactor. "Chatter" in a check valve assembly which occurs during charging and/or re-charging of control fluid within a control line is eliminated by providing a valve head and a resisting portion defined on the valve head to resist movement of the valve head when it is urged from sealing engagement on its seat during the charging or re-charging procedure.

Abstract: An object of the present invention is to provide a boiling water reactor wherein two phase flow instability in a fuel assembly occurs scarcely.A bottom flow path is formed beneath a partition plate which is supported by control rod driving mechanism housings in a reactor pressure vessel, and coolant guide tubes which open beneath the partition plate stands perpendicularly on the partition plate. An upper end portion of the coolant guide tube is restrained at outer periphery by a fixing portion of a square grid, and a coolant entrance of a fuel assembly is connected through to the coolant guide tube.Pressure loss of the coolant decreases and flow instability of the coolant in the fuel assembly occurs scarcely, lowering of a critical power ratio at a pump trip can be suppressed.

Abstract: Decreasing the speed of a pump supplying cooling water to the core in the boiling water type nuclear reactor based on a trip signal that is generated when a pump supplying feed water is tripped. This performs a first operation to decrease the nuclear reactor power to a first fixed value. A second operation lowers the nuclear reactor power to a second fixed value that is lower than the first fixed value and that prevents run-out of other first pumps in operation. The first fixed value is lower than the reactor power at the moment when the first pump is tripped. The steam flow is decreased substantially simultaneously with the decrease of the feed water flow. Thereby, the decrease of the steam flow is carried out in an early stage when the feed water pump is tripped and the possibility of nuclear reactor scram is reduced.

Abstract: A main steam isolation valve of a reactor power plant comprises a valve body provided with inlet and outlet portions through which a steam flows, a cylindrical valve disk accommodated in the valve body to be reciprocatingly movable therein along an inner peripheral surface of the valve body for opening and closing the steam flow inlet portion and a driving mechanism secured to the valve body and operatively connected to the valve disk for reciprocatingly moving the valve disk in the valve body. A coupling member is applied to an end opening of the valve body for holding the valve disk when the valve disk is shifted to a position fully opening the inlet portion. A tubular wall member is integrally formed with the valve body so as to surround the valve disk with a gap therebetween when the valve disk is fully opened. A guide rib is further disposed to the valve body for causing asymmetric steam flow in the inlet portion of the steam flow of the valve body.

Abstract: A nuclear reactor is joined to a steam turbine by a main steamline for discharging steam thereto. A plurality of flow control valves regulate flow to the turbine, and a bypass valve selectively bypasses a portion of the steam around the turbine to its condenser. A pressure regulator and turbine controller are operatively joined to the control valves and the bypass valve for controlling steamflow to the turbine. An apparatus for detecting failure of one of the control valves is operatively connected to the bypass valve, and upon failure of one of the control valves to channel sufficient flowrate, the bypass valve is automatically opened to reduce reactor pressure rise. The failure detecting apparatus also provides a reduction demand signal for reducing reactor power for allowing the bypass valve to close.

Abstract: An improved guide plate of the type used in the guide tubes of a nuclear reactor and having a central orifice for conducting coolant, and circular guide holes interconnected by slots for guidingly conducting the movement of control rodlets and control rod vanes through the guide tube is disclosed. The improvement comprises a plurality of vent openings in the form of vent holes uniformly disposed between the slots of the plate and vent gaps disposed between the outer periphery of the guide plate and the inner wall of the guide tube for reducing turbulence in the flow of coolant through the central orifice, thereby advantageously reducing both fretting and frictional engagement between the control rodlets and the guide holes. The resulting turbulence reduction further advantageously reduces the rodlet drop time through the guide tube, thereby enhancing reactor safety.

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: Method and apparatus for improving coolant flow in a nuclear reactor during accident as well as nominal conditions. The reactor has a plurality of fuel elements in sleeves and a plenum above the fuel and through which the sleeves penetrate. Holes are provided in the sleeve so that coolant from the plenum can enter the sleeve and cool the fuel. The number and size of the holes are varied from sleeve to sleeve with the number and size of holes being greater for sleeves toward the center of the core and less for sleeves toward the periphery of the core. Preferably the holes are all the same diameter and arranged in rows and columns, the rows starting from the bottom of every sleeve and fewer rows in peripheral sleeves and more rows in the central sleeves.

Abstract: The fuel assembly has fuel arrangement of a square lattice of 10 lines by 10 rows. Four water rods having a large diameter are arranged in central region of horizontal cross section wherein arrangement of 12 fuel rods is possible. The four water rods having a large diameter are so arranged adjacently in a circle as to form an internal between each other. First coolant flow path which is extended toward axial direction is formed by surrounded with the water rods having a large diameter. The first coolant flow path leads to third coolant flow path which is formed around the fuel rods through second coolant flow path which is the interval between the water rods having a large diameter.The fuel assembly is able to optimize the moderator to fuel atom number density ratio and to reduce the pressure loss because a portion to be an excessively moderated region is utilized as the first coolant flow path.

Abstract: A natural circulation boiling water reactor system using free-surface steam separation incorporates a series of tubes near the inner wall of its reaction vessel extending from a downcomer to well-above the downcomer. These tubes allow steam bubbles which would otherwise be trapped within the downcomer to escape from the recirculating water toward the steam exit of the vessel. This lowers the volume of steam in the downcomer, lowering the water velocity required to maintain a given volume of water recirculation. The reduced water velocity results in less carryunder which results in smaller voids in the reactor core. The smaller voids provide longer fuel burnups and enhance reactor stability.

Abstract: A fuel bundle of square cross section and standard vertical dimension is disclosed. The fuel bundle includes a matrix of upstanding vertical rods, a lower tie plate for supporting the matrix of vertical rods and permitting the inflow of liquid moderator coolant (preferably light water), and upper tie plate for holding the matrix of vertical fuel rods upright and permitting the outflow of liquid and vapor moderator coolant (water and steam), and a preferably square sectioned channel extending between the tie plate for confining the flow path between the tie plate around the fuel bundle. The novel characteristics of the invention is the introduction of a corresponding matrix of water rods (cylindrical rods with liquid phase water flow) positioned between the fuel rods, as opposed to the current practice of using or displacing a subset of fuel rods with large central water rod(s) for this purpose.

Abstract: A method and apparatus for improving the operating efficiency of a boiling water reactor is disclosed. Instead of coolant water being injected simultaneously at the bottom of the fuel rod assemblies and core bypass regions, the two flows then proceeding in parallel to the top of the reactor, the top of the bypass regions is sealed off and bypass coolant water is directed at the bottom of the sealing cover. This forces the bypass coolant water to flow down through the bypass regions. At the bottom of the bypass regions, provision is made to allow the bypass water to flow into the bottom of the fuel rod assemblies where it is entrained by the fuel rod coolant water flow. By converting flow through the core bypass region into a single series flow, increased operating efficiency is achieved.

Abstract: A fuel bundle for a natural-circulation boiling-water reactor includes at least one coolant bypass tube which is at least partially open at both its top and bottom and extends about 2/3 of the bundle height. Water within the bypass tube remains liquid. As it exits the top of the bypass tubes, it merges with and "cools" the flow exterior to the tube. This arrangement reduces the pressure drop across the core, increasing coolant flow. In addition, the merging coolant helps improve heat transfer at the maximum heat flux levels within the core. Lateral holes through the bypass tube can further enhance with heat transfer distribution in the core. The reduced pressure drop in the core and the more uniform heat flux distribution both permit a reactor to operate at higher power ratings.

Abstract: In a system for detecting abnormal movement of a part of a gas-cooled nuclear reactor, for example abnormal tilting of a boiler unit caused by failure of a boiler support, a tracer material is released into the gas coolant path in response to the movement. The releasing of the material may be effected by causing a gas canister to be pierced as a result of the movement. The presence of the tracer material in the coolant is detected by the coolant monitoring system.

Abstract: A forced-circulation boiling-water reactor includes bypass check valves between a downcomer and a core inlet plenum. When the recirculation pumps are operating at full capacity, there is a maximum pressure differential from the downcomer to the core inlet plenum. This pressure differential keeps the valves closed so that recirculating fluid is constrained to flow through the pumps. When the pumps are not operating, a driving water head in the downcomer forces the valves open, augmenting the flow cross section between the downcomer and the core inlet plenum, enhancing natural circulation. The enhanced natural circulation provides greater core stability during pump shutdown. The valves are selected or adjusted so that they open when the pressure differential falls through a predetermined range to augment diminished pumping capacity with a higher natural circulation flow rate.

Abstract: In an apparatus including a fluid circulation unit for allowing a fluid in a container to be circulated along a passage in the container, a flow rate measuring apparatus includes a rotational speed transducer for detecting the rotational speed of a motor for driving the fluid circulation unit and a pipe network model calculator for receiving a rotational speed signal from the rotational speed transducer. The model calculator has an initially programmed pipe network model corresponding to flows in the container and calculates the flow rate at a corresponding location of the model with the rotational speed of the circulation unit replaced by constants of the model and hence calculates a flow rate of the fluid at that location in a real passage.

Abstract: The present invention discloses an improved methodology for achieving the desirable feature of recirculation flow control in a sparger-type natural-circulation BWR, enabling reactors utilzing the method to perform in load-following and/or spectral shift modes of power operations. In its broadest aspects, the invention is based on locating the feedwater sparger at an elevational level which is about water level established during normal operation of the BWR. One or more of load following or spectral shift can be performed by controlling the feedwater fed to said BWR through the feedwater inlet and into the sparger to vary the water level to be above, at or below the elevational level of such sparger.

Abstract: A natural circulation boiling-water reactor system using free-surface steam separation incorporates a steam separator near the inner wall of its reaction vessel and extending into a downcomer. Steam so collected supplements the main steam flow from the reactor vessel to drive a turbine, which in turn drives a generator. A flow controller opens and closes a valve to adjust the flow of steam from the steam collector to the turbine. The flow controller can be programmed to implement a predetermined power output level or to adjust power output as a function of load or time. When this level is exceeded, as indicated by feedback from the generator, the flow controller valve is constricted. The auxiliary steam flow to the turbine is directly decreased, reducing power output. Concomitantly, less steam is removed from the steam collector, and thus from the downcomer. As a result there is more carryunder steam, which reduces the power generated at the reactor core.

Abstract: High temperature reactor with residual-heat transfer system comprises a cooling gas intake at the bottom and cooling gas outlet at the top so that a cooling gas can flow from the bottom to the top through the reactor core. In order to assure reliable heat transfer a bypass duct is provided with a lower end communicating with the cooling gas intake and the upper end communicating with the cooling gas outlet. The bypass duct is arranged parallel to the reactor core and passing a partial flow of cooling gas from the bottom to the top. This partial flow of cooling gas heats up only trivially. This partial flow of cooling gas is further cooled by the cooler. The upward flow of the comparatively cold cooling gas in the bypass duct stops and by itself reverses because the cooling gas in the bypass duct is drawn toward the reactor core on account of the natural convection.

Abstract: A steam generating system including a heat energy source, a feedwater heating section including at least one heat exchange unit (2) having a feedwater inlet, a device (10) for supplying feedwater to the feedwater inlet of the heat exchange unit (2), and a steam generating section composed of at least two steam generating heat exchange units (4) each having a feedwater inlet connected for receiving heated feedwater from the feedwater heating section and each connected to the heat energy source for transferring heat from the energy source to the feedwater.

Abstract: A fuel assembly has a water rod which consists of a coolant ascending path and a coolant descending path, the coolant ascending path opening below a fuel support of a lower tie plate, the coolant descending path opening above the fuel support and adapted to guide downwardly the coolant that has flowed up the coolant ascending path to the upper portion of the water rod. The lateral cross section of the coolant ascending path is more than 25 times that of the coolant descending path.The nuclear reactor with these fuel assemblies loaded in the core is operated as follows. After startup of the reactor when the reactor power is low, all the control rods inserted in the core are withdrawn completely. Then, the reactor power is controlled by regulating a liquid level formed in the water rods. This reactor operation method prolongs the lifetime of the control rods, thereby reducing the number of times they have to be replaced.

Abstract: An apparatus and process for the accurate measurement and calibration of core flow in a boiling water reactor is disclosed. The reactor includes a pressure vessel having internal forced circulation through the core. An annular downcomer region establishes reactor coolant flow downwardly in the periphery of the reactor vessel and finally upwardly and centrally into the core across a core plate. A pressure sensor detecting the pressure differential across the core plate is utilized. This pressure sensor has its measurement enhanced by input from local power range monitors in the core to utilize both the sensed pressure differential and the power to predict more accurately flow in the reactor. An algorithm is developed for utilizing the pressure differential and the real time readings from the local power range monitor to accurately gauge overall reactor coolant flow. To ensure accurate calibration, two calibration standards are itilized at steady states of reactor power output and coolant flow.

Abstract: In a nuclear electric power generating plant connected to an electric power system, an output of a nuclear reactor of the plant is reduced during a nighttime operation, increased to a predetermined level during a daytime operation and reduced during a lunch time operation in which a poisonous effect of fission products, i.e., X.sub.e.sup.135 produced during a decreased output running during a nighttime operation becomes significant.

Abstract: A natural circulation reactor includes a reactor pressure vessel, a cylindrical shroud disposed within the reactor pressure vessel to surround a reactor core and to extend to a position above the reactor core, and a coolant descending passage defined between the reactor pressure vessel and the shroud for allowing the coolant discharged from the top of the reactor core to be again supplied to the bottom of the reactor core. In such a natural circulation reactor, a liquid level of coolant in the coolant descending passage is adjusted on the basis of a reactor power request signal at a position below a liquid level of coolant in the shroud, and thus reactor power is controlled so that it may reach a predetermined reactor power by adjusting the liquid level of coolant in the coolant descending passage.

Abstract: This invention provides a method and an apparatus for controlling a polymerization reaction. The temperature difference between a target temperature, calculated from the degradation rate of a polymerization initiator, which varies as a function of temperature, and a polymerization reactor temperature is calculated. The difference in polymerization rate between a target polymerization rate previously set and a polymerization rate calculated from a measured heat release value at a certain time in the polymerization process is calculated. The temperature inside the reactor is finely adjusted on the basis of the temperature difference and the difference in polymerization rate.

Abstract: A control apparatus and method for restricting liquid flow in a liquid moving pump, usually of the centrifugal type, to prevent pump cavitation and pump prime mover overloading. The control apparatus includes sensors to detect liquid temperature and pressure at the inlet of the pump. It may further include a device such as a current transformer to develop a signal indicative of power consumed by the prime mover of the pump where the prime mover is an electrical motor. The liquid pressure and temperature indications are used to generate a specific indication of the subcooling of the liquid. The temperature indication is used to derive an indication of the instantaneous required subcooling of the pump. The subcooling indication and the required subcooling indication are introduced to a comparator. Should the subcooling of the liquid fail to exceed the required subcooling, a first control signal is generated.

Abstract: In the method of shutting down a high temperature nuclear reactor having a negative temperature coefficient of reactivity, such as a gas cooled pebble-bed nuclear reactor, while the core is operating in the critical state with the core having a predetermined critical average core temperature and with the heat generated in the critical state being removed by a coolant, effecting the shut down by reducing or discontinuing the removal of heat from the reactor core and increasing the average core temperature by an amount above the critical average core temperature for rendering the core hot sub-critical due to the negative temperature coefficient of reactivity. The core can be maintained in the hot sub-critical state by a controlled removal of after shut-down heat from the core.

Abstract: A power control apparatus of a boiling water atomic reactor for transferring the atomic reactor to a low power stand-by operation when a load of a generator is rejected. The necessity of insertion of control rods into a core is decided in accordance with the state of the atomic reactor when a load rejection occurs. When the insertion of control rods is required selected control rods are inserted into the core and a core flow rate is reduced to transfer the atomic reactor to the low power stand-by operation. When the insertion of control rods is not necessary only the reduction of core flow rate is performed.

Abstract: A nuclear reactor is supplied with feed water through a feed water pump system. A primary steam flow produced from the reactor is controlled by regulating a recirculated flow of feed water. The feed water pump system comprises two main pumps each of 55%-capacity and two auxiliary pumps each of 27.5%-capacity. Normally, the two main pumps are operated. Upon occurrence of abnormal condition of at least one main pump, the auxiliary pumps are started to supply feed water. At that time, the recirculated flow is controlled for a predetermined time to a reduced rate which is smaller as compared with that of the primary steam flow decreased rapidly due to the shutdown of the main pump. Subsequently, the recirculated flow is so controlled that the primary steam flow rate is slightly smaller as compared with the feed water flow which is determined by the available capacity of the pumps.

Abstract: A cold trap including a section for cooling inflowing liquid sodium, a section for trapping impurities in the sodium, and a double-walled cylindrical structure disposed between the cooling section and the trapping section. The double-walled structure is filled with a heat-insulating gas, and is closed at its upper end but open at its lower end. An expanded portion designed for storing the heat-insulating gas is formed at the open lower end or in the vicinity thereof of the double-walled structure.

Abstract: The pressure P.sub.R in the reactor pressure vessel, the flow rate W.sub.c of the cooling water flowing through the core in the reactor pressure vessel, and the flow rate W.sub.STM of the steam supplied from the reactor pressure vessel to the turbine are detected. The enthalpy H.sub.f of the saturated cooling water is determined on the basis of the pressure P.sub.R. The values of the cooling water flow rate W.sub.c, the steam flow rate W.sub.STM, the enthalpy H.sub.f of the saturated cooling water and a predetermined enthalpy H.sub.IN of the cooling water at the core inlet are substituted into the equation below thereby to obtain the enthalpy H.sub.fw of the feed water. ##EQU1## An error is determined between the feed-water enthalpy H.sub.fw thus obtained and the enthalpy T.sub.

Abstract: A power change operating method of a boiling water reactor, in which an operation allowable range is defined by the relation between xenon and iodine concentrations, the present concentrations is compared with the allowable range, and the core flow is controlled so that the concentrations after the change can be included within the allowable range.

Abstract: A nuclear reactor power monitoring system for monitoring the power level of a reactor and preventing an excessive rise thereof attributable to a transient increase in the core coolant flow rate before the reactor is scrammed. The system include an operating region monitor (ORM) for blocking the increase in the core coolant flow rate or running-back the flow rate when the power level exceeds a predetermined coolant block threshold power level which is a function of the core coolant flow rate.

Abstract: A control apparatus for a boiling water reactor includes a monitoring section for monitoring power distribution in the reactor, a core flow control section for controlling the core flow and an arithmetic operational section for calculating change rate of the linear heat generation rate from the output of the monitoring section to thereby cause the core flow to be increased or decreased through the core flow control section when the calculated change rate exceeds a predetermined limit, whereby the increasing rate of the linear heat generation rate of fuel rods is maintained at a value not greater than the limit value. The power distribution monitoring section serves to generalize the current power distribution in the whole core in trms of typical monitoring quantities at selected locations of in-core neutron detectors.