Abstract: An aqueous assembly has a negative coefficient of reactivity with a magnitude. The aqueous assembly includes a vessel and an aqueous solution, with a fissile solute, supported in the vessel. A reactivity stabilizer is disposed within the aqueous solution to reduce the magnitude of the negative coefficient of reactivity of the aqueous assembly during operation of the aqueous assembly.

Abstract: An injection feedwater heater for a steam power generating system includes at least one main heater body and at least one injection nozzle. The main heater body has at least one heat exchange compartment, at least one water inlet, at least one steam inlet, and at least one water outlet formed on the main heater body. The injection nozzle is provided in the main heater body at a position adjacent to the water inlet, wherein a predetermined amount of condensate water is arranged to be pumped into the main heater body through the water inlet. The condensate water passing through the water inlet is arranged to be injected into the heat exchange compartment through the injection nozzle for creating a negative pressure in the heat exchange compartment. The negative pressure drawing a predetermined amount of steam to enter the heat exchange compartment for mixing with the condensate water.

Abstract: A stable startup system includes a reactor vessel containing coolant, a reactor core submerged in the coolant, and a heat exchanger configured to remove heat from the coolant. The stable startup system further includes one or more heaters configured to add heat to the coolant during a startup operation and prior to the reactor core going critical.

Abstract: A nuclear reactor module includes a reactor vessel containing coolant, a reactor core submerged in the coolant, and a heat exchanger configured to remove heat from the coolant. The nuclear reactor module further includes one or more heaters configured to add heat to the coolant during a startup operation and prior to the reactor core going critical.

Abstract: Method of repairing nuclear reactors that include one or more submerged lines welded to one or more support brackets may include: removing a damaged section of one of the one or more submerged lines; and replacing the damaged section of the one of the one or more submerged lines without welding. Methods of operating nuclear reactors that include one or more submerged lines welded to one or more support brackets may include: shutting down the nuclear reactor; repairing damage to at least one of the one or more submerged lines without welding; and starting up the nuclear reactor. Methods of operating nuclear reactors that include one or more submerged lines welded to one or more support brackets may include: cooling down the nuclear reactor; repairing damage to at least one of the one or more submerged lines without welding; and heating up the nuclear reactor.

Abstract: A neutron source rodlet assembly having a separate source capsule assembly that is not encapsulated within the neutron source rodlet assembly. The neutron source rodlet assembly is made up, at least in part, of a neutron source positioning rodlet assembly and the source capsule assembly configured such that assembly together is feasible at a remote site and they can be shipped separately. The source capsule assembly has outer and inner capsules with the outer capsule having a threaded stud at one end that mates with a complimentary threaded recess on the neutron source positioning rodlet assembly. The inner capsule contains a neutron source. The neutron source positioning rodlet assembly and the source capsule assembly are locked together at their interface when the threaded joint is completely tightened. A secondary neutron source material may also be encapsulated within a hollow portion of the neutron source positioning rodlet assembly.

Abstract: In the present invention, a reactor power control apparatus of a natural circulation reactor comprises a power control apparatus for controlling reactor power and a pressure control apparatus for controlling reactor pressure. A power adjustment demand signal S4 is input from the power control apparatus into the pressure control apparatus. The pressure control apparatus controls degree of opening of a inlet port steam control valve provided in a moisture separation heater that is connected to a low-pressure turbine, based on the power adjustment demand signal S4. A reactor power control apparatus of the natural circulation reactor can suppress reactor power variation and obtain stable power.

Abstract: A stable startup system includes a reactor vessel containing coolant, a reactor core submerged in the coolant, and a heat exchanger configured to remove heat from the coolant. The stable startup system further includes one or more heaters configured to add heat to the coolant during a startup operation and prior to the reactor core going critical.

Abstract: The present invention relates to a method of controlling a nuclear reactor during a transient period. The method includes actuating the steam dump system in response to a temperature error signal and a power mismatch signal.

Abstract: In the present invention, a reactor power control apparatus of a natural circulation reactor comprises a reactor pressure vessel which circulates cooling water using the density difference of the coolant inside, a feed water pipe which is connected to the reactor pressure vessel and supplies cooling water, a power control section which controls the reactor power using a control rod. The feed water pipe has an ultrasonic thermometer sensor. Driving of the control rod is controlled by the power control section based on the feed water temperature detected by the feed water thermometer. The reactor power control apparatus can detect the temperature of the feed water and perform drive control of the control rod preferentially, and obtain stable reactor power.

Abstract: In order to stably control a nuclear reactor in a short time, so as not to enter an unstable region that is determined by the relationship between the reactor pressure, the reactor power and the subcooling of the core inlet coolant at start-up time, the nuclear reactor system comprises: an power control apparatus for generating a control rod operation signal for operating a control rod, based on the reactor water temperature change rate; a feed water control apparatus for generating a feed water flow rate signal and a discharge water flow rate signals based on the reactor water level signal; and a process computer for performing overall control of the power control apparatus and the feed water control apparatus, wherein the feed water control apparatus has the reactor water temperature change rate setting section for adjusting the reactor water temperature change rate set value based on the variation of the reactor water level signal.

Abstract: The present invention relates to a method of controlling a nuclear reactor during a transient period. The method includes actuating the steam dump system in response to a temperature error signal and a power mismatch signal.

Abstract: Systems and methods for a method for determining a critical effective k at an off-rated core state of a nuclear power plant includes determining, for the off-rated core state a control rod density, a percent core power, a gadolinium reactivity worth, a doppler reactivity worth, and a xenon reactivity worth responsive to a control rod pattern, a reactor power plan including the off-rated core state, and a reference effective k, calculating a change in an effective k from the reference effective k at the off-rated core state responsive to two or more parameters selected from the group consisting of the control rod density, the percent core power, the gadolinium reactivity worth, the doppler reactivity worth, and the xenon reactivity worth, and generating the critical effective k for the off-rated core state responsive to the change in the effective k from the reference effective k.

Abstract: Method of repairing nuclear reactors that include one or more submerged lines welded to one or more support brackets may include: removing a damaged section of one of the one or more submerged lines; and replacing the damaged section of the one of the one or more submerged lines without welding. Methods of operating nuclear reactors that include one or more submerged lines welded to one or more support brackets may include: shutting down the nuclear reactor; repairing damage to at least one of the one or more submerged lines without welding; and starting up the nuclear reactor. Methods of operating nuclear reactors that include one or more submerged lines welded to one or more support brackets may include: cooling down the nuclear reactor; repairing damage to at least one of the one or more submerged lines without welding; and heating up the nuclear reactor.

Abstract: In a boiling water reactor controlling each control rod hydraulically by driving a solenoid valve, the control system comprises an operation control means 41 having a duplicated data processing unit for generating sequence patterns based on the timing of the driving sequence based on the control information provided manually, a transmission control means 42 for creating a command word corresponding to each control rod being controlled based on said sequence pattern, and mutually communicating each command word between duplicated data processing units and computing the AND logic of said data within a predetermined time difference, and when the computed result coincide, transmitting the selected command word serially; a transmission unit 32 for receiving said command word and performing protocol conversion thereto, and transmitting the same to a plurality of transmission branch units positioned downstream as the control command; and a solenoid valve drive circuit 31 for driving the control rod drive unit corres

Abstract: A safe start-up of a boiling water reactor is achieved when two conditions are adhered to and monitored when the control rods are being moved out of the core: First, the drive of a control rod should be activated only when the directly and diagonally adjacent control rods remain at rest simultaneously, and, second, when all the directly adjacent control rods are still in the fully moved-in initial position.

Abstract: A method and a device are provided for monitoring a power rise during startup of a nuclear reactor (diversitary excursion monitoring). In order to be able to shut down the reactor during startup in the event of prompt critical states, a power band situated in a lower measuring range is prescribed for a signal of power range channels which detects the reactor power in the power range. Startup is continued only when the signal does not exceed an upper limit within a minimum time, after the last occasion of exceeding a lower limit of the power band.

Abstract: A safe start-up of a boiling water reactor is achieved when two conditions are adhered to and monitored when the control rods are being moved out of the core: First, the drive of a control rod should be activated only when the directly and diagonally adjacent control rods remain at rest simultaneously, and, second, when all the directly adjacent control rods are still in the fully moved-in initial position.

Abstract: A method is provided for determining the closeness to criticality of a nuclear reactor during start-up, comprising the steps of completing a control rod withdrawal step, thereby generating a change in an output signal of a neutron detector; measuring the output signal after the completion of the control rod withdrawal step and during a transient portion of the output signal; calculating the effective neutron multiplication factor (Keff) based upon the measured output signal and elapsed time between output signal measurements; and determining the closeness to criticality of the nuclear reactor based upon the calculated value of the effective neutron multiplication factor (Keff). The invention also encompasses apparatus used to perform the above method.

Abstract: An automated method for maintaining pressure within a nuclear power plant primary loop during either startup or shutdown, the method comprises the steps of partially filling a portion of a pressurizer vessel, in fluid communication with the primary loop, with a liquid for maintaining pressure in the primary loop; circulating a primary coolant through the primary loop; automatically injecting an inert gas by a first automated device, operatively connected to the pressurizer, into the pressurizer vessel when the pressure in the pressurizer vessel is less than a first predetermined pressure; and automatically venting the gas by a second automated device, operatively connected to the pressurizer, from the pressurizer vessel when the pressure in the pressurizer vessel is greater than a second predetermined pressure.

Abstract: In a boiling water nuclear reactor, a reactor primary cooling water line is filled with water at the start-up time, and the inside of a pressure vessel is pressurized by a pressurized tank and at the same time control rods are withdrawn to thereby heat cooling water in a state of single-phase flow to high temperature. Succeedingly, the pressurization is released and the pressure of the pressure vessel is gradually approximated to the saturation pressure corresponding to the cooling water temperature to thereby make the cooling water transit into a state of two-phase flow, and thereafter the cooling water is heated by nuclear reaction in the state of two-phase flow to thereby obtain predetermined reactor running temperature and pressure.

Abstract: A gas cooled nuclear reactor may have a stationary pile of spherical operating elements. The reactor may be controlled and shut down by absorber rods displaceable in channels of the side reflector. The neutron sources required for the safe start-up of such pellet pile reactors, which in the case of higher capacity pile reactors are installed in bores of the side reflector, are arranged in the stationary pile, where they are more effective and do not occupy positions in the side reflector. They may be located in one or more graphite elements having the same diameter as the operating elements. These elements may be introduced and removed into and from the core together with the operating elements and remain stationary in operation.

Abstract: A liquid metal nuclear reactor has a sodium tank, having a closed arcuate bottom and upstanding cylindrical wall for containing a reactor core in a pool of liquid metal coolant, which is disposed in a guard tank having a bottom and upwardly extending cylindrical shell, and a support inside the guard tank adjacent the closed bottom of the sodium tank to support the two tanks in spaced relationship to each other. Support legs are also provided about the bottom periphery of the guard tank adapted to rest on a base mat of a reactor cavity. The bottom of the guard tank may be a closed arcuate shape, or the bottom may be a flat bottom wall that rests on the base mat with the cylindrical shell divided into two radially spaced sections, with a pool of lead-bismuth alloy provided on the bottom to seal the spaces between the two shell sections at the lower region thereof.

Abstract: In a liquid metal heat transport system including a source of thaw heat for use in a space reactor power system, the thaw flow throttle or control comprises a fluid passage having forward and reverse flow sections and a partition having a plurality of bleed holes therein to enable fluid flow between the forward and reverse sections. The flow throttle is positioned in the system relatively far from the source of thaw heat.

Abstract: A pressurized water nuclear reactor has a reactor vessel arranged in a pool, which is filled with a neutron absorbing liquid, for example borated water. The reactor vessel is closed except for tubes connecting it with a tray above it. The coolant in the circuit rises from the vessel to the tray, gives up its heat by flashing, and flows back to the bottom of the vessel, driven by natural circulation. The tray is separated from the pool by a vapor-filled bell, which surrounds it. In the bell the vapor gives up its useful heat to a condenser. The relatively low boron content of the cooling circuit, compared to the pool, is achieved by continuous dilution of the condensate from vapor additionally generated out of the pool water. The dilution process is an equilibrium with continuous inflow of the pool water. The inflow is automatically controlled by the pool level, which rises when the pool water is pressed out from below the bell by overproduction of vapor.

Abstract: A nuclear reactor start-up method of starting up a boiling water reactor which is in the cold shut-down state. The method includessetting a subcooled temperature of a coolant in a reactor core inlet to a range smaller than a critical subcooled temperature which is determined by the condition whereby the coolant starts boiling in the reactor core and the condition whereby the instability due to a periodic variation of buoyancy begins,thereafter starting an increase of power of the reactor.

Abstract: A ball-bed (pebble-bed) nuclear reactor, instead of having fuel elements e or less continuously withdrawn and new or reconstituted fuel elements more or less continuously reintroducted, is initially partly filled with fuel balls of which two-thirds have a fissionable material content 12% below and the upper third 24% higher than the average content. This filling meets the requirements of criticality in order to begin operation. Thereafter, fuel balls are added slowly, a few hundred per day, having 150 to 250% of the average fissionable material content of the initial loading, thus preserving the criticality requirments, while keeping the temperature within safe limits until the reactor cavern is filled. Thereafter, the reactor is shut down, cooled off, pressure relieved and emptied, the last step typically from above. An ordered array of the fuel balls in regular layers avoids excess pressure loads on the reflector over the life time of the filling.

Abstract: A secondary source positioning mechanism for a fuel assembly in a nuclear reactor locates an individual one of the secondary sources in a respective guide thimble of the fuel assembly aligned with an opening defined through the upper core plate which overlies the fuel assembly. The positioning mechanism has a tapered locating boss which registers with a chamfer on the lower side of the core plate about one of its openings and a resilient holddown device which resiliently couples the upper end of a rod containing the secondary source to the locating boss in a manner which restrains the upper end of the rod in the lateral direction and positions the secondary source rod in the axial direction.

Abstract: Nuclear reactor cooled by a liquid metal comprising a main vessel sealed by a sealing slab, said vessel containing liquid metal overhung by a covering of inert gas, the reactor core being placed in a primary vessel integrated into the main vessel in such a way that the primary vessel at the outlet from the core constitutes a hot collector for the liquid metal and the space defined between the main vessel and the primary vessel constitutes a cold collector for the liquid metal, at least one heat exchanger supplied by the liquid metal contained in the hot collector under the action of at least primary pumps located in the main vessel, wherein the exchanger is positioned outside the main vessel at a level above that of the sealing slab and communicates with the hot collector by at least one intake pipe and with the cold collector by at least one outlet pipe, it being possible to empty the exchanger and pipes into the main vessel by gravity, the upper part of each exchanger being filled with the inert gas coverin

Abstract: The approach to criticality of a nuclear reactor provided with an artificial neutron source is monitored by two neutron detectors, one which measures the neutron flux in a localized region around the artificial neutron source, and the other located simlarly with respect to the core geometry but removed from the localized region. When the reactor is subcritical, the artificial source dominates and the outputs of the two detectors are noticeably different. As the reactor approaches criticality, the critical mode distribution of neutrons, which is symmetrical with respect to the two detectors, becomes dominant and the detector outputs approach each other in magnitude. The ratio of the output signals therefore, provides an indication of the reactor reactivity state. Either linear or logarithmic ratio signals can be generated, and if desired, either can be compared with an appropriately scaled reference signal representative of a selected reactivity state, such as a multiplication factor of 0.

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 plant having feed water heating means for heating feed water supplied to a nuclear steam supply unit during plant start up and/or shutdown. High pressure heaters are positioned in a feed water delivery line between a feed water pump and a steam generator forming part of the nuclear steam supply unit. Steam is admitted into the high pressure feed water heaters from a main steam delivery line extending from the steam generator during start up to heat the feed water.