Abstract: For a soft error of an electronic device, a technique capable of ensuring high reliability because of a low soft error rate (SER) is provided. By using building data including information of a structural object of a building and facility data including information of a plurality of facilities including an electronic device arranged in the building, a SER calculating device calculates a model including an attenuation index value representing a degree of attenuation of radiation entering the building attenuated by the structural object of the building until the radiation reaches a position of the facility arranged in the building, calculates a SER at each position of the facility arranged in the building by using the model including the attenuation index value, and outputs information including the SER at each position of the facility.

Abstract: A critical heat flux prediction device, a critical heat flux prediction method, a safety evaluation system, and a core monitoring system using the safety evaluation system can predict critical heat flux in a core of a reactor with a high degree of accuracy by obtaining a correlation plot distribution representing a relation of critical heat flux on a thermal equilibrium quality based on experimental data, approximating a correlation plot distribution through a logistic function that is a model function in which critical heat flux is expressed by a function of a thermal equilibrium quality, and obtaining a critical heat flux correlation of critical heat flux and a thermal equilibrium quality.

Abstract: An inspection apparatus for a penetration pipe of a nuclear reactor head comprising: a body; a probe module installed at the body and having a probe which is inserted in the penetration pipe to inspect damage of the penetration pipe; a fixing module installed along a longitudinal direction of the body and having an expanding cylinder which is inserted in the penetration pipe to support an inner diameter of the penetration pipe; and a rotating module installed in the longitudinal direction of the body and having an expanding cylinder which is inserted in the penetration pipe to support the inner diameter of the penetration pipe.

Abstract: Energy savings can be obtained by masking a computationally-intensive task from a processor performance management system which selects the processor performance state based on the load on the processor (CPU). By preventing the PPM system from reacting to the computational load the application places on the processor, the time to complete execution of the application increases but the energy used by the application may be greatly reduced and thermal stress on the CPU is also reduced (preventing noisy fans from operating). This approach makes it convenient to run a computationally intensive task as a background task. The masking can be achieved by running the task in tiny bursts, with micro-sleeps in between them, so that the average CPU load is low over a time period that the PPM system uses for measuring CPU activity/load.

Abstract: This invention relates generally to ultrasonic inspection of welds and more particularly, to apparatus and methods for ultrasonic inspection of welds on access hole covers found in boiling water reactors having jet pumps. The apparatus includes a base, a center frame coupled to the base and projecting vertically relative to the access hole cover, a radial arm structured to rotate on the center frame and having attached thereto a first pneumatic linear thruster and a second pneumatic linear thruster, and a skew motor assembly and transducer are attached to the first pneumatic linear thruster for scanning the access hole cover weld. The skew motor assembly is structured to control the angle of the transducer and the first pneumatic linear thruster is structured to raise and lower the transducer.

Abstract: A method for electrodepositing a metal layer from an electrolytic bath is described, wherein the concentrations of at least two components of the electrolytic bath are monitored by carrying out the following steps: (a) a sample is taken from the electrolytic bath; (b) the sample is supplied to a column for solid phase extraction, which contains a solid sorbent material; (c) the column is subjected to a washing procedure with a first eluent, wherein the at least two components remain on the column, and undesired components are eluted from the column; (d) the at least two components are eluted from the column by a second eluent; (e) the concentrations of the at least two components in the eluate obtained in step (d) are determined without separating the components from each other. In this method the use of poly(divinyl benzene-co-N-vinyl pyrrolidone) for solid phase extraction is particularly advantageous.

Abstract: This document's object is to provide an axial power distribution control method in which only the control of an axial power distribution in a nuclear reactor with a simple operation with a clear operational target keeps the control of a xenon oscillation, thereby suppressing the xenon oscillation to an extremely small magnitude in advance at the same time.

Abstract: Electrochemical corrosion potential (ECP) probe assemblies may be used to monitor ECP of materials due to coolant chemistry in an operating nuclear reactor. Example embodiment assemblies include at least one ECP probe that detects ECP of potentially several different materials, a structural body providing a fluid flow path for the coolant over the ECP probes, and a signal transmitter that transmits or carries ECP data to an external receiver. The ECP probes may be of any number and/or type, so as to detect ECP for different component materials, including stainless steel, a zirconium alloys, etc. The ECP probes may further detect ECP due to ion concentration, pH, etc. The ECP data may be transmitted through wired or wireless signal transmitters. Example methods include installing and using example embodiment ECP probe assemblies in nuclear reactors and facilities.

Abstract: A neutron detector system and method for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. Comparison of the observed neutron count distribution with a Poisson distribution is performed to distinguish fissile material from non-fissile material.

Abstract: An axial power distribution control device includes an axial offset calculation unit 52, a parameter calculation unit 53, and an axial offset determining unit 55. The axial offset determining unit 55 predicts whether a core axial offset of the power distribution is increased or decreased after a current time, based on a major axis of an ellipse drawn by the xenon parameter and the iodine parameter calculated by the parameter calculation unit 53 and the xenon parameter and the iodine parameter at the current time. This makes it possible to predict a change of the axial offset of the power distribution of a reactor for suppressing a xenon oscillation in the reactor.

Abstract: A jet pump beam (hereinafter, referred to as a beam) is fitted into a pair of projecting portions installed to a transition piece, and after the beam is arched, a beam bolt engaged with the beam is tightened. An end of the beam bolt comes in contact with a top surface of an insert member fitted into an elbow disposed between the pair of projecting portions. An ultrasonic sensor head is fixed to the beam bolt, and ultrasonic waves are sent to the beam bolt from the ultrasonic sensor in the ultrasonic sensor head. An ultrasonic measuring apparatus obtains echo intensity of each of reflected waves generated at the end of the beam bolt and a bottom surface of the insert member, and based on the echo intensities, an echo intensity ratio R is calculated. The installed state of the beam is checked using the echo intensity ratio R.

Abstract: A real time monitoring system of a spent fuel pool includes a detection unit configured to detect condition information using a sensor installed in the spent fuel pool; an input storage unit configured to receive and store configuration information of a spent fuel, history information related to burnup, and a normal value and a limit value of current condition information; an operation and determination unit configured to calculate the current condition information of the spent fuel by using the condition information detected by the detection unit and the configuration information and the history information stored in the input storage unit and configured to determine a risk level by comparing the current condition information with the limit value corresponding to the risk level; and a display unit configured to display the current condition information calculated by the operation and determination unit and configured to display the determined risk level.

Abstract: A nuclear reactor comprises a pressure vessel, a nuclear reactor core comprising fissile material disposed in the pressure vessel, and a core basket disposed in the pressure vessel and containing the nuclear reactor core. An incore instrument is disposed in a guide tube of the nuclear reactor core. The incore instrument has a cable extending out of the bottom of the nuclear reactor core and making a 180° turn. A bottom support element is attached to the bottom of the core basket and defines or includes a routing tube that routes the cable of the incore instrument through the 180° turn. In some embodiments, the pressure vessel includes upper and lower vessel portions, the nuclear reactor core and core basket are disposed in the lower vessel portion, and the vessel penetration is through the lower vessel portion or through a mid-flange connecting the upper and lower vessel portions.

Abstract: The invention relates to a device for inspecting a fuel assembly in the pool of a nuclear plant, that comprises an image sensor with an observation field, and further comprising a boom with at least one removable fastener to the assembly, a reference graduation extending along an axis parallel to a longitudinal axis of the assembly, so that the image sensor can monitor within its field both the boom and the assembly. The device is designed so that the boom can be removably attached to the fuel assembly when the latter is suspended outside the pool by the hook of the machine for handling the fuel assemblies. The invention also relates to a corresponding method.

Abstract: A system and method for predicting acoustic loads expected on a boiling water reactor (BWR) may include a BWR scale model, a test fixture for generating air flow in the scale model, and one or measurement devices for monitoring system behavior to predict how acoustic loads may affect plant operation for the BWR being evaluated.

Abstract: Systems and methods of monitoring rod control systems of nuclear power plants, including: measuring a particular signal applied to at least one coil of a rod movement mechanism at one or more predetermined times relative to an initial energizing of a mechanism coil; comparing the measured signal to a reference signal parameter, and determining if the measured signal deviates from the reference signal parameter by a predetermined amount to indicate degradation of the rod control system.

Abstract: Systems and methods of monitoring step movements of control rods of a nuclear power plant, including measuring output signals of a plurality of rod movement coils during a step movement sequence of one or more control rods, analyzing the output signals to determine a direction of the step movement sequence, and comparing one or more output signals to reference output sequence to verify that a step of the control rod has occurred.

Abstract: A nuclear power distribution measurement assembly that is sized to fit within an instrumentation thimble of a nuclear fuel assembly, that employs a spaced tandem arrangement of thermo-acoustic engines, each of which has a heat source side that is insulated from the reactor coolant traversing the nuclear core in which the fuel assembly is to be placed and a cold side housing a resonator chamber with enhanced thermal conductance to the coolant. The resonator chamber of each of the thermo-acoustic engines is of a different length to generate a different frequency whose amplitude is proportional to the neutron activity at the axial and radial position of the thermo-acoustic engine. The frequency identifies the measurement assembly's position. Acoustic telemetry is employed to monitor the acoustic waves generated by the individual thermo-acoustic engines to provide a remote reading of the axial and radial power distribution of a reactor core.

Abstract: A high resolution digital rod position indication (high resolution DRPI) system having improved resolution. The high resolution DRPI system monitors the rod control cluster and provides an indication of the rod position with precision to a single step. In addition, the high resolution DRPI system is capable of producing a rod position output compatible with existing rod control systems. The improved resolution of the high resolution DRPI system allows the actual position of the control rods to be continuously monitored and eliminates the need for or reduces the frequency of offline re-initialization of the step counters.

Abstract: A method for determining a leaking nuclear fuel rod that remotely measures the internal pressure in the plenum region of the fuel rod by remotely measuring the thermal response of the rod when a localized heating is imposed on the outside of the rod in the plenum region. The temperature of the cladding at two symmetrically spaced points on either side of the location where the heating is imposed is remotely monitored as a function of time and compared to a standard to determine the internal pressure of the plenum.

Abstract: A neutron detector system and method for discriminating fissile material from non-fissile material wherein a digital data acquisition unit collects data at high rate, and in real-time processes large volumes of data directly into information that a first responder can use to discriminate materials. The system comprises counting neutrons from the unknown source and detecting excess grouped neutrons to identify fission in the unknown source. Comparison of the observed neutron count distribution with a Poisson distribution is performed to distinguish fissile material from non-fissile material.

Abstract: A device for measuring and correcting a parallelism error of an upper plug end of a nuclear fuel rod comprising mechanisms for measuring a parallelism error and for correcting said error, and a mechanism for positioning said device on the fuel rod and cooperating with a rack on which the fuel rod is stored, said correction means being arranged opposite the measuring mechanism relative to the location of the fuel rod, in order to allow a measurement of the parallelism error during correction of said error.

Abstract: A nuclear power distribution measurement assembly that is sized to fit within an instrumentation thimble of a nuclear fuel assembly, that employs a spaced tandem arrangement of thermo-acoustic engines, each of which has a heat source side that is insulated from the reactor coolant traversing the nuclear core in which the fuel assembly is to be placed and a cold side housing a resonator chamber with enhanced thermal conductance to the coolant. The resonator chamber of each of the thermo-acoustic engines is of a different length to generate a different frequency whose amplitude is proportional to the neutron activity at the axial and radial position of the thermo-acoustic engine. The frequency identifies the measurement assembly's position. Acoustic telemetry is employed to monitor the acoustic waves generated by the individual thermo-acoustic engines to provide a remote reading of the axial and radial power distribution of a reactor core.

Abstract: A reactor core coolant temperature measuring apparatus for measuring the temperature of the coolant flowing in the reactor core of a nuclear reactor comprises a ?-ray thermometer having a temperature measuring section arranged below the bottom of the reactor core and a ?-ray heat generation detecting section arranged between the bottom and the top of the reactor core, a cable for transmitting the signal output from the temperature measuring section and an output signal processing means for receiving the signal and computationally determining the local temperature of the coolant at the position of placement of the temperature measuring section.

Abstract: A method for measuring the neutron flux in the core of a nuclear reactor, the method including several steps recurrently performed at instants separated by a period, the method comprising at each given instant the following steps: acquiring a total signal by a cobalt neutron detector placed inside the core of the reactor; assessing a calibration factor representative of the delayed component of the total signal due to the presence of cobalt 60 in the neutron detector; assessing a corrected signal representative of the neutron flux at the detector from the total signal and from the calibration factor; assessing a slope representative of the time-dependent change of the calibration factor between the preceding instant and the given instant; the calibration factor at the given instant being assessed as a function of the calibration factor assessed at the preceding instant, of the slope, and of the time period separating the given instant from the preceding instant.

Abstract: Systems and methods of monitoring position information of a control rod in a nuclear reactor, including detecting an output signal induced by movement of a control rod with respect to at least one detection coil of the control rod, the output signal having identifiable characteristics representative of a plurality of positions of the control rod within the at least one detection coil, and processing the identifiable characteristics to derive the positions of the control rod within the at least one detection coil.

Abstract: According to an embodiment, a pressurized water reactor plant has a primary system which includes: a reactor vessel for housing a reactor core which is cooled by a primary coolant, a single steam generator, a hot leg pipe for connecting the reactor vessel and the steam generator, cold leg pipes, at least two primary coolant pumps, and a pressurizer for pressurizing the primary coolant pressure boundary in which the primary coolant flows. The plant also has: a passive cooling and depressurization system which is a primary depressurization means for equalizing the primary system pressure to the secondary system pressure at the time of a tube rupture accident of the steam generator, and a reactor containment vessel containing the primary system and cooling the primary system by air cooling. Thus, a compact pressurized water rector with high economic efficiency, safety, and reliability can be provided.

Abstract: A method to determine the effective masses of deposits of fissile matter, including a first measuring step to measure the count rate of deposits of fissile matter of known effective masses to form a matrix [C]0, a second measuring step to measure the count rate of deposits of fissile matter to form a matrix [C], and a computing step to calculate the effective masses to be determined.

Abstract: The present invention relates to a nuclear power plant cable including: conductors; at least one or more insulation layers adapted to correspondingly surround the conductors; a sheath adapted to surround the insulation layers; and monitoring means adapted to monitor the states of the insulation layers or the state of the sheath in real time.

Abstract: A device for detecting steel is provided. The steel is used in a nuclear plant. The device has a detecting circuit of electrochemical corrosion potential (ECP) and alternative current (AC) impedance. An environment in a boiling water reactor (BWR) is simulated under hydrogen water chemistry (HWC). The environment is used for detecting intergranular stress corrosion cracking (IGSCC) of components coated with different precious metals. Thus, effect of coating different precious metals on steels can be evaluated for HWC.

Abstract: A nuclear reactor vibration surveillance system has a first ultrasonic transducer for transmission, an ultrasonic transmitter, a second ultrasonic transducer for reception, an ultrasonic receiver, a signal processor, and a display unit. The first ultrasonic transducer for transmission is arranged on the outer surface of a reactor pressure vessel and is configured to convert a transmission signal into an ultrasonic pulse signal and allow the ultrasonic pulse to be transmitted to a reactor internal component. The second ultrasonic transducer for reception is arranged on the outer surface of the reactor pressure vessel and is configured to receive a reflected ultrasonic pulse reflected by the reactor internal component and convert the received reflected ultrasonic pulse into a reception signal.

Abstract: In one embodiment, the in-bundle force measurement device includes an elongated main body, and a sensing structure disposed at a distal end of the main body. The sensing structure is configured to output a measurement indicative of force applied to the sensing structure.

Abstract: An in-core neutron monitor that employs vacuum microelectronic devices to configure an in-core instrument thimble assembly that monitors and wirelessly transmits a number of reactor parameters directly from the core of a nuclear reactor without the use of external cabling. The in-core instrument thimble assembly is substantially wholly contained within an instrument guide tube within a nuclear fuel assembly.

Abstract: The method comprises the steps of: partitioning the core in cubes to constitute nodes of a grid for computer implemented calculation, splitting the cubes into a first category and a second category, each cube of the first category being adjacent only to cubes from the second category so that the first category and second category of cubes constitute a checkerboard-like pattern, ordering the cubes of the first category and then the cubes of the second category, calculating neutron flux and/or thermohydraulics parameters by using an iterative solving procedure of at least one linear system and/or an eigensystem, the components of an iterant of the linear system and/or the eigensystem constituting the neutron flux and/or thermohydraulics parameters to be calculated, wherein, during the iterative solving procedure, calculations are conducted on the cubes of the first category then on the cubes of the second category.

Abstract: A computer implemented method for modelizing a nuclear reactor core, including the steps of: partitioning the core in cubes to constitute nodes of a grid for computer implemented calculation, calculating neutron flux by using an iterative solving procedure of at least one eigensystem, the components of an iterant of the eigensystem corresponding either to a neutron flux, to a neutron outcurrent or to a neutron incurrent, for a respective cube to be calculated. A control parameter is varied to impact a neutron eigenvalue ? through a perturbed interface current equation and drive the neutron eigenvalue ? towards 1.

Abstract: An object of the invention is to embody a small and inexpensive passage selector, which can be applied by commonly using an index device even if the number of detector passages is changed, and which is easy to make inspection and maintenance. The passage selector of a reactor in-core nuclear-measuring apparatus of the invention includes: a drive motor; an index device that is driven by the drive motor and that makes a rotary output of a predetermined index number; a central rotating shaft that is driven to rotate by the index device and that causes a passage selecting guide tube to be located in opposition to any detector passage; and a speed-increasing and decreasing device that is interposed between an output shaft of the index device and the central rotating shaft, and that adjusts the index number of the central rotating shaft.

Abstract: A surface conditioning scheduling process is used to mitigate susceptibility to crack initiation or crack growth in a boiling water nuclear reactor using a plurality of working crews. A plurality of working zones are defined in an annulus region of a reactor vessel flange, and a plurality working zones are defined in a core region of the reactor vessel. One of the working crews is positioned in each of the annulus region working zones, and one of the working crews is positioned in each of the core region working zones. The working crews perform surface conditioning in areas of the reactor vessel accessible from their respective working zones such that the surface conditioning process can be completed in no more than thirty days.

Abstract: The invention concerns a method comprising measurement on a fuel channel (14) of fuel assemblies (8) for nuclear boiling water reactors. The method comprises that: the measurement is done by the use of a non-destructive inductive eddy current measurement method, the measurement is done on a fuel channel (14) which has been used at least a certain time during operation in the core of a nuclear boiling water reactor, the measurement is done when the fuel channel (14) is located in water, the measurement is done on different places on the fuel channel (14), wherein through the method at least the hydride content of the fuel channel (14) at said places is determined. The method can be used in order to find out how shadow corrosion from a neighboring control rod influences the properties of the fuel channel (14).

Abstract: The invention relates to a method of determining at least one technological uncertainty factor in respect of nuclear fuel elements (23) as a function of variations in the production parameters of the elements (23) in relation to nominal values. The inventive method comprises a step involving the use, for at least one production parameter, of a collective variation in said parameter in relation to the nominal value within a batch of produced elements (23). The invention can be used, for example, to design, produce and check pellets for light water reactors.

Abstract: A method and apparatus for monitoring a parameter in an irradiated environment and communicating a signal representative of the monitored parameter to a less caustic environment that employs a wireless transmitter that is powered by the irradiated environment. The power for the wireless transmitter is derived from a self-powered radiation detector disposed within the radioactive environment.

Abstract: The present invention relates to an irradiation device for material test using a gamma ray radiated from a spent nuclear fuel assembly and provides the irradiation device for material test using a gamma ray radiated from a spent nuclear fuel assembly wherein an irradiation device for material test to achieve a radiation effect evaluation is manufactured to be movable upward, downward and horizontally in order to study the hardening phenomenon of the frail materials to the radiation among the atomic power facilities using a gamma ray radiated from a spent nuclear fuel assembly, thereby it is possible to adjust a position of the spent nuclear fuel used for material test using a gamma ray radiated from a spent nuclear fuel and a test material, identify a distance between the spent nuclear fuel and the test material easily with a scale and evaluate the radiation effects on the materials used at facilities handling a spent nuclear fuel under the same situation as they are really exposed.

Abstract: A pair of linear arrays of gamma thermometer (GT) sensors arranged in a nuclear reactor core including: a first linear array of GT sensors, wherein the GT sensors are arranged asymmetrically along a length of the first linear array; a second linear array of GT sensors, wherein the GT sensors are arranged asymmetrically along the second linear array and wherein the second linear array of GT sensors is asymmetrical with respect to the first linear array of GT sensors, and the first linear array positioned in the reactor core at a first core location and the second instrument housing positioned at a second core location, wherein a line of symmetry of the core extends through a center of the core and the first core location is the same horizontal distance from the line of symmetry as the second core location.

Abstract: In an underwater remote surface inspection method for a reactor constituting member, in order to improve the precision of an operation of inspecting a surface shape of the reactor constituting member, an underwater remote surface inspection apparatus includes a replica picking head, an ultrasonic vibrator, and a replica agent cartridge. The replica picking head is pressed against a surface of core internal structure as an inspection target. A replica agent is supplied from the replica agent cartridge into the replica agent supply region formed inside the replica picking head and contacting with the surface of the core internal structure. After the operation of supplying the replica agent ends, an ultrasonic wave is transmitted from an ultrasonic vibrator to the replica agent inside the replica agent supply region. Accordingly, gas bubbles or liquid bubbles existing inside the replica agent supply region rise up so as to be discharged to the outside of the replica picking head through an air extracting hole.

Abstract: A method to assess and predict pressurized water stress corrosion cracking in operational nuclear power plants and the effect of adding zinc compounds into a reactor coolant system of the nuclear power plant.

Abstract: An object of the present invention is to provide an inspection apparatus for inspecting weld zones in a reactor pressure vessel, the inspection apparatus comprising: an ultrasonic probe 6 for emitting an ultrasonic wave; a probe holding unit 60 for holding the ultrasonic probe 6 such that a ultrasonic wave transmitting surface of the ultrasonic probe 6 is kept in direct contact with or at a constant distance from the outer surface of the reactor pressure vessel 1; a pressing unit 50 for pressing the probe holding unit 60 parallel to a central axis of a control rod drive housing 8 against the reactor pressure vessel; and a rotator 40 for rotating the probe holding unit 60 and the pressing unit 50 about the central axis of the control rod drive housing 8.

Abstract: To provide a rod block monitor in which irrespective of the selection or non-selection of a control rod, such a process is repeatedly executed that gain adjustment is performed at fixed intervals at all times to match a local area average value of signals of neutron detectors surrounding the control rod with a nuclear reactor average power and a block level appropriate to the local area average value of the neutron detector signals after the gain adjustment is set.

Abstract: A method of monitoring a condition of a nuclear reactor pressure vessel disposed in a radioactive environment is provided. The method includes the steps of sensing a condition of the reactor pressure vessel with an instrument, transmitting a signal indicative of the condition of the reactor pressure vessel from the instrument to a powered wireless transmitting modem disposed in the radioactive environment, wirelessly transmitting a signal indicative of the condition of the reactor pressure vessel from the transmitting modem to a receiving modem in the line of sight of the transmitting modem, transmitting a signal indicative of the condition of the reactor pressure vessel from the receiving modem to a signal processing unit, and determining the condition of the reactor pressure vessel from the wirelessly transmitted signal.

Abstract: A nuclear power plant control system (1) includes a control button (21) which receives an operation for controlling a control target device (40), a notification lamp (12) which notifies that a control signal corresponding to the operation received by the control button (21) arrives at a predetermined position on a path connected from a control panel (20) to the control target device (40), and a control signal inhibition unit (33) which inhibits a control signal from arriving at the control target device (40) in midstream between the predetermined position on the path and the control target device (40) in response to an operation received by a test permission button (11).

Abstract: The present invention relates to a method of detecting and/or measuring defects in at least one part. The method includes providing a neutron source that produces a neutron flux, arranging the neutron source such that the neutron flux at least partly penetrates said part, providing a detection device for detecting neutrons, providing a hydrogen containing substance on a surface of said part such that the hydrogen containing substance penetrates into defects in said part, arranging the detection device to detect neutrons from the neutron flux that have been reflected by said substance, and detecting and/or measuring at least one possible defect in said part using the detection device to detect defects by detecting said reflected neutrons.

Abstract: A method for monitoring and acoustic damping a boiling water reactor which includes: a reactor pressure vessel; a steam pipe for transporting steam out from a steam dome of the reactor pressure vessel; a high pressure turbine driven by the steam; a feedwater heater which heats feedwater supplied to the reactor pressure vessel using bleed steam from the high pressure turbine; a bleeding valve which adjusts a flow rate of the bleed steam; and a pressure sensor provided in a main steam line including the steam dome and the steam pipe. The method includes monitoring a fluctuating pressure in the main steam line and controlling an opening degree of the bleeding valve based on a magnitude of the fluctuating pressure.