KR101813450B1 - Calculation method to protect the core of the pressurized light water reactor protection system - Google Patents

Abstract

The present invention relates to a pressurized light water reactor type reactor protection apparatus and a control method thereof.
The present invention integrates a DNBR stop function and an LPD stop function signal into a programmable field programmable gate array (FPGA) processor, simplifies the parameters input to the core protection signal processor, It is possible to generate a stop signal for core protection directly in the reactor protection system, so that the configuration of the protection system can be simplified.

Description

Technical Field [0001] The present invention relates to a pressurized light-water reactor type reactor protection apparatus and a control method thereof,

The present invention relates to a nuclear reactor protection apparatus for protecting a reactor core by automatically stopping a reactor when an abnormality occurs by measuring the state of devices related to the reactor in a nuclear power plant. More particularly, the present invention relates to a nuclear reactor protection reactor Water Reactor) to optimize the algorithms that are input and processed into the analog signal processor of the core protection computer to shorten the computation time and protect the core by enabling the stop signal to protect the core directly from the reactor protection system The present invention relates to a pressurized light water reactor type reactor protection apparatus and a control method thereof that simplify the operation system configuration and facilitate operation and correction.

Korea Nuclear Power Generation Facility Nuclear power plant with OPR (Optimized Power Reactor) -1000 and 1,4000MW type PWR type nuclear power reactor with APR (Advanced Power Reactor) -1400 as a nuclear reaction of 1,000MW class PWR type Has used the Core Protection Calculator System (CPCS) as part of its efforts to protect the reactor core. The purpose of this system is to protect nuclear reactor core loaded with nuclear fuel in a sound state. If the operation of the nuclear power plant reaches the AOO, Anticipated Operational Occurrence and Departure from Nucleate Boiling (DNB) and Local Power Density (LPD) , The system automatically stops nuclear power generation and protects the reactor from operating beyond its limit.

For the nuclear power plants of Westinghouse type (Kori 1, 2, 3, 4) and Pramatom (Uljin 1 and 2) in Korea, the departure from nucleate boiling ratio (DNBR) and the linear power density LPD, inear power density calculations are performed directly in the reactor protection system. This is possible because the calculation procedure using the parameters of DNBR and LPD is simple in the Westing and Pramatom type reactors. The calculation of DNBR in Westinghouse type nuclear power plants is processed in the OTΔT module installed in the reactor protection system, and the LPD calculation is processed in the OPΔT module. The OT DELTA T module and the OP DELTA T module are each formed in the form of a single analog card. These modules monitor whether the reactor is out of the established core protection limits, taking into account the dynamic calibration results for the temperature deviations of the measured hot and cold tubes, the temperature of the coolant, and the power distribution of the reactor core.

The digital core protection calculation system of OPR-1000 and APR-1400 calculates the three-dimensional output distribution considering both the radial and axial outputs, so the accuracy of the calculation result is high. In particular, DNBR and LPD can be calculated quickly even if the reactor condition changes suddenly due to an accident, so that nuclear power generation can quickly respond to situations where the operation is expected to be in a transient state. The analog type core protection arithmetic system is relatively inferior in stability because the operation speed is slower than the digital type core protection arithmetic system and it can not respond quickly to situations where nuclear power generation reaches a predicted operation transient state.

Table 1 shows the thermal tolerance of the analog type core protection calculation system used in rings 3 and 4 and the digital type core protection calculation system used in OPR type nuclear power generation. As shown in the following table, it can be seen that the thermal margin of the OPR type nuclear power generation is about 30% higher than that of the analog type nuclear power generation. These results indicate that safety is increased because of the high degree of margin from the state where the reactor is stopped due to the accident to the occurrence of the problem in the actual core.

Comparison of thermal tolerance between analog and digital type core protection arithmetic Thermal tolerance Analog type
(Kori 3 and 4) Digital type
(OPR nuclear power generation) DNB 107.9 115.4 LPD 110.1 146.2

The digital core protection arithmetic systems employed in the domestic OPR-1000 and APR-1400 nuclear power plants require complicated arithmetic operations in order to accurately calculate the thermal margin, and thus can not be processed in the reactor protection system. Therefore, after processing the thermal margin using a separate calculator, only the reactor shutdown values by DNBR and LPD are transmitted to the reactor protection system.

The core protection computation system (CPCS) computes the DNBR of the reactor coolant to protect the core from unstable operating conditions and calculates whether the heat transfer from the cladding of the fuel to the coolant is affected by the amount of bubbles generated in the cladding , And functions to protect the fuel rod by calculating the LPD of the fuel pellet located in the covering material.

The reason for protecting the reactor by calculating the DNBR is that when the heat is transferred between the heating surface of the fuel rod and the cooling fluid, the conductivity becomes worse if it goes out of the boiling heat transfer region and the temperature of the fuel rod is increased. In case of extreme case, to be.

1 is a block diagram showing the relationship between a conventional reactor protection system and a core protection computing unit.

Referring to FIG. 1, each of the four channels includes a plurality of sensing means 100 installed in the reactor related apparatus, a core protection arithmetic unit 110 for computing an LPD stop signal, A vise table processor 120 receiving the measured values of the sensing means 100 and the DNBR and LPD stop signals input to the core protection arithmetic unit 110 to generate a reactor stop signal; A local concurrent logic unit 130 for inputting a reactor stop signal output from the vise table processor 120 and performing local concurrent logic for four channels; And a stop start unit 140 for outputting a stop signal of the reactor when the trip signal of two or more channels out of the four channels is generated according to the result of the local simultaneous logic unit 130 to drop the control rod into the reactor core .

2 is a block diagram showing control logic of a conventional core protection computing unit.

More specifically, the parameters input to the computing unit include the reactor coolant system pressure, the error value of the control rod assembly computer, the position information of the control rod assembly, the out-of-neutron flux measurement, the reactor coolant pump speed, Respectively.

The different types of input parameters include various calculations, more specifically, calculation of flowmeter output, calculation of control rod position error, calculation of neutron output, calculation of reactor heat output; Through calculation of the radial peak coefficient, the axial peak coefficient and the reactor output, the result of calculating the nuclear boiling bounce rate and the local maximum output with the axial output value is output. The resulting output value is the reactor shutdown signal generated by the nuclear boiling rejection ratio (DNBR) and the local power density (LPD). If the nuclear boiling rate is less than the stop setpoint of 1.3 or if a transient of 21 kW / ft or more, where the local maximum output is at the stop setpoint, occurs, a control signal is output to stop the reactor to prevent melting of the fuel pellet.

3 shows a program block diagram implementing control logic for obtaining DNBR and LPD in a conventional digital core protection computation system.

Referring to this drawing, a conventional CPCS divides software for implementing control logic into program blocks of FLOW 300, CEAC 310, POWER 320, STATIC 330, UPDATE 340, and TRIPSEQ 350 Respectively. This is applied to the domestic OPR-1000 and APR-1400 domestic nuclear power generation.

Each of the above-mentioned program blocks performs the following functions.

First, the FLOW (300) program calculates the normal flow rate by inputting the RCP speed signal. This flow value is calibrated using a reference value in a 12-hour period for conservative calculation.

The POWER (320) program generates the axial power distribution using the position of the control rod and the external instrument signal used in the UPDATE as input.

STATIC (330) The program uses DNBR and steam quality at the restricted axial node.

The UPDATE 340 program updates the updated value by considering the changed core variables after calculation of this value for the values calculated in detail in the POWER 320 program and the STATIC 330 program.

In the conventional CPCS, the speed of the reactor coolant pump is used to calculate the reactor coolant flow rate, and the radial power density required to calculate the core output is indirectly estimated using the control rod position information. Were measured. The DNBR and LPD calculation programs are divided into a STATIC (330) block and an UPDATE (340) block, which complicates the calculation process.

In the existing CPCS program, the computation for obtaining DNBR and LPD is divided into STATIC (330) and UPDATE (340) because the computer computation speed at the time of development is low and the time requirement of DNBR and LPD calculation Because it can not be accurately calculated. That is, if the STATIC 330 outputs a detailed calculation value every 2 seconds, the UPDATE 340 uses a method of updating it every 50 ms considering the changed core state.

That is, the PLC system outputs the calculation result every 0.45 seconds because the speed of the central processing unit drops. Therefore, in order to meet the system performance requirement to output the calculation result every 50 msec, static calculation is performed every 2 seconds, This result was output by using another input variable after every 50msec.

This method requires a complicated calculation process in order to accurately measure the thermal margin, so it can not be processed by the reactor protection system. Therefore, it is necessary to use an operating system (OS) and a central processing unit After computing using a separate computer, the DNBR and LPD reactor shutdown values are transferred to the plant protection system.

Accordingly, there is a problem in that the computation system is complicated, the operation and correction are inconvenient, and the test maintenance and operator training becomes difficult, in addition to the general errors and security problems that may occur in the hardware and software installed in the computer have.

Conventionally, in the PLC system in which the control logic of the core protection arithmetic unit is implemented, there is an advantage that the operation and maintenance are relatively advantageous by using simple control logic, but it is used for simple process control with a relatively small number of input / output processes per processor, There is a problem in that there is no compatibility between processors and output devices of different models, such as using gateways between different models or limiting transmission / reception data.

KR 10-0848881 B1 (2008. 07. 22.) KR 10-0875467 B1 (Dec. 16, 2008)

 SH. Ahn et al., Several Problems in RCS Flow Rate Measurement, Journal of KNS, Vol. 30, No. 6, 1998

An object of the present invention is to integrate signals of the DNBR stop function and the LPD stop function into a vise table processor to simplify the input parameters of the conventional core protection arithmetic unit analog signal processing apparatus, The present invention provides a reactor protection system capable of generating a signal and a calculation method thereof.

Another object of the present invention is to provide a reactor protection system mounted on a vise table processor by incorporating the function of core protection arithmetic implemented by separate hardware and software for calculation of nuclear boil-off ratio and locally maximum output into stop- And an operation method thereof.

It is a further object of the present invention to provide a system and method for removing a reactor coolant pump speed by removing the speed measuring device installed on the reactor coolant pump shaft using the inlet and outlet pressure differences of the steam generator primary side instead of the reactor coolant pump rate used to calculate the reactor coolant flow rate, And the calculation process is simplified because the flow calculation process and the calibration process are processed in software in the vise table processor. To provide a reactor protection system and a calculation method thereof.

It is still another object of the present invention to provide a radial output table (Look-up Table) proportional to the reactor heat output in a system indirectly estimating the radial power density required for calculating the output of the reactor using conventional control- The present invention is to provide a reactor protection system and its computation method capable of eliminating various hardware for measuring a position of a control rod and processing a signal and an algorithm for processing the signal.

Another object of the present invention is to integrate a DNBR and LPD calculation program divided into an UPDATE block and a STATIC block into one program block and employ a device such as an FPGA (Field Programmable Gate Array) to rapidly perform DNBR and LPD operations To provide a reactor protection system and method of operation that can be performed.

Technical features of the present invention to accomplish the object of the present invention are as follows. Technical features of the pressurized light water reactor type reactor protection system for achieving the object intended by the present invention include a reactor having four channels, A digital reactor protection system that can shut down a reactor safely by outputting a reactor operation stop signal when an abnormal state of the reactor power plant is detected according to a result of a comparison logic for comparing and determining parameters of related devices with a set set value, A plurality of sensing means installed in the reactor related device for sensing the operating state; The reference value of the reactor stop condition parameter of the reactor related apparatus is set to the set value, the signal of the DNBR stop function and the LPD stop function are integrated into the processor, the measured parameter signal from the sensor is inputted, A vise table processor for comparing and judging whether a reactor stop condition parameter has reached a reference value and outputting a trip signal for stopping the reactor when the set value is exceeded; Comparison logic information according to a trip signal output from the vise table processor of the four channels is inputted and comparison logic information inputted is compared for each process variable to determine a process variable in which two or more channels of the four channels are in a trip state , Local simultaneous logic that judges whether trips are made for each plant protection measure in accordance with a combination of process variables in which two or more channels among four channels are in a trip state, and generates a plurality of simultaneous logic information related to the plant protection measure Wealth; And a stop start signal part for performing simultaneous logic on a plurality of simultaneous logical information outputted from the local concurrent logical part and generating a start signal for starting a plant protection measure according to a result of simultaneous logical execution, ) Activates the protection action of the reactor shutdown system according to the generated start signal.

The vise table processor is equipped with a DNBR stop function and an LPD stop function using a programmable field programmable gate array (FPGA) device.

In order to accomplish the object of the present invention, a computation method for protecting a core in a pressurized light-water reactor protection system includes four channels and compares parameters of the reactor-related devices allocated for each channel with set values The method of claim 1, further comprising the steps of: detecting the abnormal state of the nuclear power plant according to the result of the comparison logic to determine whether the nuclear power plant is in operation; ; Performing comparison logic for comparing the process variables input for each channel, the DNBR stop function, and the LPD stop function with the set values, and generating comparison logic information indicating whether or not the process variable is tripped; The comparison logic information generated in the above step is transmitted to the same channel and three other channels to collect comparative logic information generated for each bus, and local concurrent logic is performed for each of the collected comparison logic information, Generating a concurrent logic result value associated with the action; And outputting a stop start signal for starting the protection measure according to the simultaneous logic result value generated in the step.

In the reactor protection calculation method of the present invention, the core protection calculation for signals of the DNBR stop function and the LPD stop function is performed using the inlet and outlet pressure differences of the primary side of the steam generator.

In the reactor protection calculation method of the present invention, the power density in the reactor radial direction for signals of the DNBR stop function and the LPD stop function is calculated using the heat output of the reactor.

Further, in the reactor protection calculation method of the present invention, the signals of the DNBR stop function and the LPD stop function determine the position of the control rod by the reactor output.

Also, in the reactor protection calculation method of the present invention, the average output power of the DNBR stop function and the LPD stop function of the pulse train and the local heat pulse is calculated by first calculating the pulse train speed and the local heat flux, And the value is used to update the result value.

The average output power is calculated by compensating the uncertainty and the output bias uncertainty, and then the updated local power density is obtained by considering the LPD calculation uncertainty.

According to the present invention, the step of generating the comparative logic information in the vise table processor comprises the steps of: comparing the reactor coolant system pressure, the differential pressure across the steam generator, the extraneous neutron flux measurement, the temperature of the hot and cold tubes of the reactor, (S20); The pressure difference between the inlet and the outlet of the steam generator output in the step S20 is inputted to calculate and output the flow rate, and the neutron output is calculated by inputting the measurement value of the extraneous neutron flux, and the temperature of the high- Calculating and outputting the reactor heat output (S21); The flow rate calculation value output in the step S21 is input to calculate and output a radial echo impulse response coefficient, a neutron output calculation value is input to calculate and output an axial direction impulse response coefficient, a reactor heat output is input, A step S22 of calculating; The reactor temperature output in the step S20, the flow rate calculation value output in the step S21, the radial direction rainfall coefficient and the axial direction rainfall coefficient outputted in the step S22 are inputted and the nuclear boiling departure rate is calculated (S23); Calculating (S24) a local maximum output by inputting the flow rate calculation value output in the step (S21), the radial direction rainfall coefficient outputted in the step (S22), the axial rainfall coefficient and the reactor output value; A step (S25) of generating a nucleating boiling release stop signal by comparing the nuclear boiling rejection rate outputted in the step (S23) with a set reactor stopping set value and outputting it to local simultaneous logic; Comparing the local maximum output value outputted in the step S24 with the set reactor stop set value to generate a local maximum output stop signal and outputting it to the local simultaneous logic (S26).

The calculation process of the reactor coolant flow rate using the differential pressure of the steam generator and the calculation results of the reactor heat output is performed by comparing the differential pressure of the primary side of the steam generator with the piping specifications of the reactor coolant (high temperature and low temperature pipes) and the enthalpy of the reactor coolant (S30); A step S31 of calculating the heat output of the reactor coolant using the differential pressure of the primary side of the steam generator inputted in the step S30, the piping specifications of the reactor coolant (high temperature section and low temperature section), and the enthalpy of the reactor coolant ; Calculating (S32) the turbine side heat output; A step (S33) of outputting a calculation result of the core operation monitoring device (COLSS); A step (S34) of comparing the heat output value of the reactor coolant, the turbine side heat output value, and the calculation result of the core operation monitoring apparatus, which are output in the step (S34); And outputting the calibrated reactor coolant heat output value according to the step S34 (S345).

In the present invention, an essential process for calculating DNBR and LPD is to obtain a hot pin, which is the most expensive fuel rod. To select a hot pin in a cylindrical reactor, the radial and axial peak factors must be obtained and multiplied to obtain the composite peak factor. According to the present invention, the process of synthesizing the hot pin output distribution to obtain the DNBR stop function and the LPD stop function signal includes: a step S40 of inputting a reactor output; A step (S41) of generating a control rod position definition table to be compared with a reactor output using a reactor output; A step (S42) of generating a table of the radial direction peak coefficient of the reactor according to the control rod position definition table generated in the step S41; A step (S43) of outputting a signal of an extraneous neutron detector; A step (S44) of inputting a signal of the extraneous neutron detector output in the step S43 to synthesize an axial direction power distribution and calculating a peak coefficient; A step S45 of combining the radial peak coefficient of the reactor output in the step S42 and the output distribution of the hot pin to which the axial power distribution and the peak power coefficient outputted in the step S44 are inputted; A step S46 of outputting the output distribution of the hot pins output in the step S45 to a nucleation boiling departure rate calculating routine and a step S47 of outputting the output distribution to the local maximum output calculating routine.

The reactor protection system according to the present invention includes the function of the core protection arithmetic unit implemented using separate hardware and software for calculation of the nuclear boiling rejection rate and the local maximum output into the stoppable of the reactor shutdown system, And the system configuration is simplified by mounting.

In addition, by eliminating the speed measuring device installed on the reactor coolant pump shaft by using the inlet side inlet and outlet pressure difference of the steam generator instead of the reactor coolant pump rate used to calculate the reactor coolant flow rate, , The flow calculation process and the calibration process are processed in the vise table processor by software, so that the calculation process is simplified.

Further, in a method indirectly estimating the radial power density required for calculating the output of the core using conventional control rod position information, a radial output table (Look-up Table) proportional to the reactor heat output is used to estimate And various hardware and algorithms for signal processing have been eliminated and stability has been secured.

In addition, the program for calculating DNBR and LPD is integrated into one program block and can be operated at a high speed.

According to the present invention, first, the input parameters of the signal processing device are simplified, and the stop signal for core protection can be generated directly in the reactor protection system by a quick calculation process, so that the core can be protected quickly.

Secondly, the functions of core protection arithmetics implemented by separate hardware and software for calculation of nuclear boiling bounce rate and local maximum output are included in stopping variables of reactor stopping system and they are mounted on the vise table processor, so that the system configuration is simplified The system operation and calibration can be facilitated, and the test, maintenance, and operator training can be facilitated.

Third, by eliminating the speed measuring device installed on the reactor coolant pump shaft by using the inlet side and outlet pressure difference of the steam generator primary side instead of the reactor coolant pump rate used to calculate the flow rate of the reactor coolant, , The flow calculation process and the calibration process are processed in software in the vise table processor, so that the calculation process is simplified.

Fourthly, in the method indirectly estimating the radial power density required for calculating the output of the core using the conventional control rod position information, a radial output table proportional to the reactor heat output is used to estimate By measuring the position of the control rod according to the method, various hardware for processing a signal and an algorithm for processing the hardware are eliminated.

Fifth, the DNBR and LPD calculation programs, which are divided into the UPDATE block and the STATIC block, are integrated into one integrated program block, and DNBR and LPD operations are performed using the latest devices such as FPGA (Field Programmable Gate Array) Can be processed accurately and quickly, and the system requirement of 50 msec can be performed.

1 is a block diagram showing the relationship between a conventional reactor protection apparatus and a core protection calculation apparatus
2 is a block diagram showing control logic of a conventional core protection computing unit.
3 is a block diagram of a program for implementing control logic in a conventional digital core protection circuit
4 is a block diagram of a reactor stop system in which the DNBR stop function signal and the LPD stop function signal of the present invention are integrated into a vise table processor
FIG. 5 is a flowchart showing a core protection calculation function integrated in a reactor protection system vise table processor module of the present invention. FIG.
FIG. 6 is a flow chart illustrating a correction process using a coolant flow rate calculation and a heat output calculation result using a differential pressure of a steam generator in the present invention. FIG.
7 is a flowchart illustrating a process of synthesizing a hot pin output distribution using a reactor output in the present invention.

The features and advantages of the present invention will become more apparent from the following detailed description of the present invention when taken in conjunction with the accompanying drawings. Hereinafter, preferred embodiments of the present invention will be described in detail with reference to the accompanying drawings.

FIG. 4 shows a block diagram of a reactor stop system in which a DNBR stop function signal and an LPD stop function signal are integrated into a vise table processor according to the present invention.

4, the PWR type reactor protection system according to the embodiment of the present invention includes a sensing unit 200, a vise table processor 210, a local simultaneous logic unit 220, a stop start unit 230, And a driving unit 240.

The sensing unit 200, the vise table processor 210, the local simultaneous logic unit 220, and the stopping start unit 230 are formed of four channels. In FIG. 1, one channel is shown.

The sensing means 200 includes a plurality of sensors installed in the reactor related apparatus; Amplifying means for amplifying an analog signal sensed by the sensor; And an A / D converter (not shown) for converting the analog signal amplified by the amplifying means into a digital signal to compute the analog signal.

The vise table processor 210 is configured to mount a total of 12 kinds of reactor stop set values and a DNBR and LPD stop software optimally proposed in the present invention in a programmable field programmable gate array (FPGA) do.

The vise table processor 210 determines a trip state by receiving a parameter sensed by a sensor installed in a reactor-related system and comparing the parameter with a preset value.

The local concurrent logic unit 220 determines the trip output of each variable determined by the vise table processor 210 of each channel, together with the result of another channel, whether or not the corresponding channel is tripped. In the local concurrent logic unit 220, based on the trip output result of the vise table processor 210 of the magnetic channel and the trip output of the three-channel vise table processor, the output of the vise table processor 210 of two or more channels among the four channels If it is a trip, it creates a trip for each variable. A variable for stopping the reactor among the outputs of the local simultaneous logic unit 220 is transmitted to the reactor stop start unit 230.

The reactor stopping start unit 230 internally performs a predetermined time delay routine to receive the output from the local simultaneous logic unit 220 to prevent the occurrence of an unnecessary trip signal due to an unexpected disturbance in the protection system, And outputs a final reactor stop signal when the trip state is maintained.

The control rod driving unit 240 causes the control rod to fall freely into the reactor due to gravity by stopping the power supplied to the control rod driving device (not shown) so that the reactor is stopped.

FIG. 5 shows a stop system for performing a core protection calculation function for generating comparison logic information incorporated in the vise table processor 20 in which the DNBR stop function and the LPD stop function are implemented in an FPGA.

The step of generating the comparative logic information in the vise table processor 20 includes the step of outputting the temperature of the reactor coolant system pressure, the differential pressure of the steam generator, the extraneous neutron flux measurement, and the temperatures of the hot and cold tubes of the reactor )Wow; The pressure difference between the inlet and the outlet of the steam generator output in the step S20 is inputted to calculate and output the flow rate, and the neutron output is calculated by inputting the measurement value of the extraneous neutron flux, and the temperature of the high- Calculating and outputting the reactor heat output (S21); The flow rate calculation value output in the step S21 is input to calculate and output a radial echo impulse response coefficient, a neutron output calculation value is input to calculate and output an axial direction impulse response coefficient, a reactor heat output is input, A step S22 of calculating; The reactor temperature output in the step S20, the flow rate calculation value output in the step S21, the radial direction rainfall coefficient and the axial direction rainfall coefficient outputted in the step S22 are inputted and the nuclear boiling departure rate is calculated (S23); Calculating (S24) a local maximum output by inputting the flow rate calculation value output in the step (S21), the radial direction rainfall coefficient outputted in the step (S22), the axial rainfall coefficient and the reactor output value; A step (S25) of generating a nucleating boiling release stop signal by comparing the nuclear boiling rejection rate outputted in the step (S23) with a set reactor stopping set value and outputting it to local simultaneous logic; And a step S26 of comparing the local maximum output value outputted in the step S24 with the set reactor stop set value to generate a local maximum output stop signal and outputting it to the local simultaneous logic.

Among the above input variables, the present invention uses the out-of-neutron velocity measurement, the high-temperature tube temperature and the low-temperature tube temperature information of the reactor coolant pipe, and the reactor coolant pump velocity measurement for calculating the flow rate of the reactor coolant, To be replaced by a differential pressure gauge installed in the main body. The algorithm for generating the comparative logic information is to eliminate the complexity of the conventional CPCS program, the control rod assembly operator error value, the control rod assembly position information, and the reactor coolant pump speed.

5, the signals for protecting the core in the conventional core protection computing unit include the reactor coolant system pressure, the control rod assembly pressure, And the temperature of the low temperature pipe of the reactor. In contrast, in the present invention, the error value of the control rod assembly computer and the control rod assembly value of the control rod assembly, The positional information is deleted and simplified to four input signals, indicating that the front end module for signal processing has been deleted.

The present invention simplifies the input signal to reduce the complexity of the system configuration by optimizing the calculation function of the core protection arithmetic unit to shorten the arithmetic time and employs a high performance arithmetic element such as FPGA which provides a remarkably fast arithmetic operation, Reduce it.

In order to integrate the signals of the DNBR stop function and the LPD stop function into the vise table processor 20, input parameters must be simplified in the algorithm of the CPCS program.

As one method of simplifying the input variables in the CPCS of the present invention, the flowmeter acid of the reactor coolant is calculated by using the pressure difference between the inlet side and the outlet side of the primary side of the steam generator.

The method of calculating the flow rate of the reactor coolant using the pressure difference at both ends of the steam generator as described above is different from the algorithm using the reactor coolant pump speed in the conventional CPCS program.

As an alternative to simplifying the input variables in the CPCS of the present invention, the radial power density required to calculate the core output is estimated using a radial output table (Look-up Table) proportional to the heat output of the reactor .

This method eliminates the process of measuring the position of the control rod and processing the signal, unlike the conventional indirect CPCS program which indirectly estimates the radial power density required to calculate the output of the core using the position information of the control rod.

In addition, the CPCS program of the present invention integrates DNBR and LPD calculation programs divided into an UPDATE block and a STATIC block into one core protection calculation block in the conventional CPCS program. To do this, the calculation algorithm of the STATIC block, which performs static calculation every 2 seconds, and the UPDATE block update algorithm are integrated and simplified to meet the requirement of 50 msec. In the present invention, it is possible to calculate the fault rate and the local heat calculation algorithm which are calculated by the existing STATIC algorithm as they are, but the calculation can be performed within 50 msec by using a device having a high operation speed such as FPGA. In addition, UPDATE incorporates procedures to update using hot-pin heat flux distribution, vapor quality, pressure, and flow values to respond appropriately in transient situations.

Since the vise table processor 20 of the present invention is implemented by an FPGA (Field Programmable Gate Array), integration of the core protection calculation block can be performed at a time faster than 50 msec. Therefore, UPDATE block and STATIC block are integrated It is possible to do. This makes it possible to meet CPCS system performance requirements that require calculation results to be output every 50 msec.

The parameters required to obtain the mass flow rate using the steam generator primary pressure differential are represented by the following formula (1).

Here, Mc : Mass flow rate

Δp : steam generator primary pressure difference

C : Flow rate coefficient of discharge

β : input and output end cross-sectional area ratio of differential pressure gauge

ε : expansion coefficient

ρ : Fluid density

d is the inside diameter of the reactor coolant pipe.

Calculation of the mass flow rate according to the present invention requires that the geometry of the reactor coolant piping be periodically corrected since the uncertainty in the flow measurement results is significant. This geometry is due to the fact that a double elbow is formed in the reactor coolant piping, which does not have a well-developed turbulent shape, which is the optimum flow measurement condition, and the flow rate distribution in the piping changes continuously.

This uncertainty is estimated to be about 1.6% in the proposed "Non-Patent Document 1", so it is required to optimize uncertainty through periodic correction.

For this, as shown in FIG. 5, the calorific value of the reactor coolant is firstly measured using the coolant flow rate, the primary system piping and the coolant piping information of the steam generator, and the enthalpy due to the coolant system temperature.

This measurement result is compared with the calorimetric result of the secondary calorimetry measured every hour and the calorimetry result of the primary calorimetry calculated by the core operation limit monitoring system. With this calibration, the coolant flow rate is measured again and the flow is re-calculated. However, comparison and correction are performed offline every 12 hours rather than in real time.

Equation 2 below is a formula for calculating the corrected coolant flow rate by dividing the calorific value Q calibrated through the comparison of the calorific value calculated in the secondary side and the calorific value calculated in the COLSS system by the difference between the enthalpy h _i and the exit end enthalpy h _o of the steam generator input end .

FIG. 6 shows a correction process using the coolant flow rate calculation using the steam generator differential pressure and the calculation results of the secondary side heat output and the COLSS heat output.

6, the calculation of the reactor coolant flow rate using the differential pressure of the steam generator at both ends and the method of calculating the reactor heat output are based on the difference between the primary pressure of the steam generator and the reactor coolant (high temperature and low temperature tubes) A step (S30) of inputting the piping specification and the enthalpy of the reactor coolant; A step S31 of calculating the heat output of the reactor coolant using the differential pressure of the primary side of the steam generator inputted in the step S30, the piping specifications of the reactor coolant (high temperature section and low temperature section), and the enthalpy of the reactor coolant ; Calculating (S32) the turbine side heat output; A step (S33) of outputting a calculation result of the core operation monitoring device (COLSS); A step (S34) of comparing the heat output value of the reactor coolant, the turbine side heat output value, and the calculation result of the core operation monitoring apparatus, which are output in the step (S34); And outputting the calibrated reactor coolant heat output value according to the step S34 (S35).

In accordance with the present invention, the variables necessary for the calculation of the nuclear boiling point rate and the local maximum output are simplified. This makes it possible to simplify the functions of core protection arithmetic implemented by separate systems (hardware and software) as one of the stopping parameters of the reactor stop system.

Obtain the hottest hot pin to calculate DNBR and LPD.

To select a hot pin in a cylindrical reactor, the radial and axial peak factors must be obtained and multiplied to obtain the composite peak factor.

In the axial direction, 20 nodes in the vertical direction can be selected using the extraneous neutron detector, and the hottest part of them can be obtained. However, the problem is to obtain the radial value. In the conventional OPR-1000 and APR-1400 reactors, it is solved by using a table of the control rod position versus the radial output distribution table prepared in the form of a table, but the hardware configuration is complicated to use the control rod position information in the core protection arithmetic unit .

Therefore, in the present invention, a method using a reactor output instead of the control rod position information is used to eliminate the need for a hardware control module necessary for receiving and processing control rod position information.

FIG. 7 is a flowchart illustrating a process of synthesizing a hot pin output distribution using a reactor output in the present invention.

Referring to this figure, the signals of the DNBR stop function and the LPD stop function include a step S40 of inputting a reactor output; A step (S41) of generating a control rod position definition table to be compared with a reactor output using a reactor output; A step (S42) of generating a table of the radial direction peak coefficient of the reactor according to the control rod position definition table generated in the step S41; A step (S43) of outputting a signal of an extraneous neutron detector; A step (S44) of inputting a signal of the extraneous neutron detector output in the step S43 to synthesize an axial direction power distribution and calculating a peak coefficient; A step S45 of combining the radial peak coefficient of the reactor output in the step S42 and the output distribution of the hot pin to which the axial power distribution and the peak power coefficient outputted in the step S44 are inputted; (S46) of outputting the output distribution of the hot pins output in the step (S45) to the nuclear boiling diversion calculation routine, and (S47) outputting the output distribution to the local maximum output calculation routine.

For example, when 100% of the reactor is output, about 4 cm of the control rod group 5 is inserted and all the control rods are drawn out, so that the control rod position in the normal state can be predicted. However, if the control rod position mismatch occurs under abnormal conditions, this will cause the operator to manually stop the reactor.

In the present invention, since the reactor output is used as the control rod position prediction information, the conventional method for calculating the hot pin is used.

On the other hand, the signals of the DNBR stop function and the LPD stop function are calculated by first calculating the flow rate and the local heat velocity, and then updating the result values using the steam quality, pressure, and flow rate values. Calculate the average power.

In addition, the computed average power is compensated for the uncertainty and the output bias uncertainty to calculate the core mean power, and then the updated local power density is obtained considering the LPD calculation uncertainty.

Hereinafter, the procedure for updating the initially calculated DNBR value to the optimum value is as follows.

Calculating the updated impulse rate using the steam quality, pressure, and flow rate values, calculating the updated local heat flux using the fuel rod peak coefficient and the updated average neutron power, and then calculating the updated flux series and the local flux To update the DNBR value.

The procedure for updating the LPD value that is primarily calculated in the above is to calculate the core average output by compensating the output uncertainty and the output bias uncertainty, and calculate the quadrature output slope unacceptable value and the 3-dimensional peak coefficient After the corrected average output is calculated, the updated local power density is obtained in consideration of the LPD calculation uncertainty.

200: sensing means 210: vise table processor
220: local concurrent logic unit 230: stop start unit
240:

Claims (7)

delete When the abnormality is detected in the reactor power plant according to the result of the comparison logic for comparing and determining the parameters of the reactor related devices allocated to each channel and the set values, the reactor operation stop signal is outputted and the reactor A method for controlling a reactor protection,
Detecting an operation state of the nuclear power plant-related devices on a channel-by-channel basis;
In the above step, the vise table processor integrating the process variables inputted for each channel, the DNBR stop function and the LPD stop function is integrated, and comparison logic comparing with the set value is performed to generate comparison logic information indicating whether or not trip is performed for each process variable ;
The comparison logic information generated in the above step is transmitted to the same channel and three other channels to collect comparative logic information generated for each bus, and local concurrent logic is performed for each of the collected comparison logic information, Generating a concurrent logic result value associated with the action;
And outputting a stop start signal for starting the plant protection measure according to the simultaneous logic result value generated in the step,
The signals of the DNBR stop function and the LPD stop function determine the position of the control rod by the reactor output,
The DNBR stop function and the LPD stop signal are calculated by first calculating the kinetic and local heat fluxes and then updating the result values using the steam quality, pressure and flow values. Wherein the control unit controls the operation of the control unit. 3. The method of claim 2,
The step of generating comparison logic information in the vise table processor comprises:
A step S20 of outputting the pressure of the reactor coolant system, the pressure difference between the two ends of the steam generator, the measurement value of the extraneous neutron flux, the temperature of the hot and cold pipes of the reactor;
The pressure difference between the inlet and the outlet of the steam generator output in the step S20 is inputted to calculate and output the flow rate, and the neutron output is calculated by inputting the measurement value of the extraneous neutron flux, and the temperature of the high- Calculating and outputting the reactor heat output (S21);
The flow rate calculation value output in the step S21 is input to calculate and output a radial echo impulse response coefficient, a neutron output calculation value is input to calculate and output an axial direction impulse response coefficient, a reactor heat output is input, A step S22 of calculating;
The reactor temperature output in the step S20, the flow rate calculation value output in the step S21, the radial direction rainfall coefficient and the axial direction rainfall coefficient outputted in the step S22 are inputted and the nuclear boiling departure rate is calculated (S23);
Calculating (S24) a local maximum output by inputting the flow rate calculation value output in the step (S21), the radial direction rainfall coefficient outputted in the step (S22), the axial rainfall coefficient and the reactor output value;
A step (S25) of generating a nucleating boiling release stop signal by comparing the nuclear boiling rejection rate outputted in the step (S23) with a set reactor stopping set value and outputting it to local simultaneous logic;
And a step S26 of comparing the local maximum output value outputted in the step S24 with the set reactor stop set value to generate a local maximum output stop signal and outputting it to the local simultaneous logic. Control method. 3. The method of claim 2,
The calculation of the reactor coolant flow rate and the heat output of the reactor using the differential pressure of the steam generator
A step (S30) of inputting a differential pressure of the primary side of the steam generator, piping specifications of the reactor coolant (high temperature section and low temperature section), and enthalpy of the reactor coolant;
A step S31 of calculating the heat output of the reactor coolant using the differential pressure of the primary side of the steam generator inputted in the step S30, the piping specifications of the reactor coolant (high temperature section and low temperature section), and the enthalpy of the reactor coolant ;
Calculating (S32) the turbine side heat output;
A step (S33) of outputting a calculation result of the core operation monitoring device (COLSS);
A step (S34) of comparing the heat output value of the reactor coolant, the turbine side heat output value, and the calculation result of the core operation monitoring apparatus, which are output in the step (S34);
And outputting the corrected reactor coolant heat output value according to the step (S34) (S345). 3. The method of claim 2,
The signals of DNBR stop function and LPD stop function,
A step S40 of inputting a reactor output;
A step (S41) of generating a control rod position definition table to be compared with a reactor output using a reactor output;
A step (S42) of generating a table of the radial direction peak coefficient of the reactor according to the control rod position definition table generated in the step S41;
A step (S43) of outputting a signal of an extraneous neutron detector;
A step (S44) of inputting a signal of the extraneous neutron detector output in the step S43 to synthesize an axial direction power distribution and calculating a peak coefficient;
A step S45 of combining the radial peak coefficient of the reactor output in the step S42 and the output distribution of the hot pin to which the axial power distribution and the peak power coefficient outputted in the step S44 are inputted;
(S46) of outputting the output distribution of the hot pins output in the step (S45) to the nuclear boiling diversion calculation routine and outputting the output distribution to the local maximum output calculation routine (S47) Control method of protective device.

delete 3. The method of claim 2,
The procedure for updating the first calculated DNBR value to the optimum value is as follows.
The steam series, pressure and flow values are used to calculate the updated series,
The updated localized heat flux is obtained using the fuel rod peak coefficient and the updated average neutron power,
And the DNBR value is updated using the updated index series and the local heat flux.

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