KR20040099884A - Integrated Thermal On-line Protection System for a Nuclear Power Plant - Google Patents

Abstract

PURPOSE: An integrated thermal on-line protector system is provided to integrate functions of a control element assembly calculator and a core protection calculator. CONSTITUTION: An integrated thermal on-line protector system includes a plurality of integrated thermal on-line protectors(10,11,12,13)(ITOP), a maintenance and test panel(27)(MTP) and an operator module(28)(OM), and a control element assembly position display(29)(CEAPD). The integrated thermal on-line protectors includes a CPU(Central Processing Unit) calculating a departure from nucleate boiling ratio(DNBR) and a local power density(LPD) by integrating operations of a core protection calculator(CPC) and a control element assembly calculator(CEAC), an auxiliary calculator module, and a common and input/output module. The maintenance and test panel receives core driving information through an intra channel network. The operator module enables operators to monitor the core driving information in real time. The control element assembly position display displays the control electrode position information from the maintenance and test panel in graphics.

Description

Integrated Thermal On-line Protection System for a Nuclear Power Plant

The present invention relates to a safety system of a nuclear power plant, and more particularly to an improved nuclear reactor protection system for maintaining the integrity of the nuclear fuel loaded in the reactor.

In general, a nuclear reactor is a device that generates heat by using uranium nuclear fission, and a bundle of nuclear fuel rods containing a concentrated uranium sintered body in a fuel rod is loaded therein. A coolant flow path is formed between the nuclear fuel rods so that heat generated from the fuel rods is transferred to the coolant. The coolant heated by the reactor passes through a once-through steam generator and is circulated back to the reactor by a coolant pump. At this time, the reactor pressure is maintained at a constant level by the pressurizer, and the steam generator generates steam through heat transfer with the reactor cooling water and generates electricity by rotating the steam turbine.

Reactors are divided into light water reactors and heavy water reactors, depending on the type of cooling water. Pressurized water reactors are divided into a cooling type pump, a pressurizer and a steam generator, such as a loop type outside the reactor pressure vessel and an integral type present inside. Many commercialized large-scale nuclear power plants for the production of electricity are loop-type pressurized water reactors, and integrated reactors are small-capacity reactors for the purpose of producing electricity and thermal energy.

Among these, the LWR maintains high temperature and high pressure to maximize the heat transfer from the fuel rods and steam generators and to prevent boiling of the reactor coolant.The heat transfer between the fuel rods and the coolant is convective heat transfer or nuclear It appears as a Nucleate Boiling Heat Transfer phenomenon. According to the heat transfer characteristics between the heating surface and the fluid, the heat transfer efficiency is drastically deteriorated when it is out of the nuclear boiling heat transfer region, so that the local temperature of the heating surface is greatly increased. In other words, if the heat transfer between the fuel rod and the coolant is not good, there may be an accident in which the fuel melting or the fuel rod is locally damaged.

Therefore, in the design of fuels, in order to keep the damage caused by fuel melting or fuel rod damage within the allowable limit, monitoring by setting the limits on the fuel rod local power density (LPD) and the Departure from Nucleate Boiling (DNB) do. The LPD limit is defined as the linear power density of the fuel rod that generates nuclear fuel melting, and the DNB limit is defined as the DNB heat flux or critical heat flux of the fuel rod in which nuclear boiling off occurs. The critical heat flux is calculated by empirical formula using the given local core condition and defines the ratio to actual measured heat flux as the DNB Ratio (DNBR). The reactor thermal protection system is a reactor safety system that monitors the respective design limits for LPD and DNBR using measurement signals for reactor operating conditions and shuts down the reactor in case of violation of the limits.

Reactor thermal protection systems for pressurized water reactors are classified into analog protection systems and digital protection systems. In Korea, Kori Units 1-4 and Yeonggwang 1-2 employ analog protection systems, while Glory 3-6, Uljin 3-6, and Shingori 1-2 use digital protection systems. . Analog reactor protection systems that have been commercialized to date include overpower / overtemperature ?? T (OP ?? T / OT ?? T) protection systems, and digital reactor protection systems provide real-time core protection using computers. Core Protection Calculator System (CPCS) is used. Both OP ?? T / OT ?? T and CPCS are essential reactor protection systems to keep nuclear reactors safe from fuel melting and nuclear boiling off. OP ?? T / OT ?? T, an analog reactor thermal protection system, shows overpower ?? T and overtemperature ?? T from the measured core pressure, core mean temperature and temperature difference (?? T), and upper and lower output deviations of the core, respectively. The trip setting of is calculated using an analog circuit to determine whether the reactor is stopped. The CPCS consists of four independent protection system channels using four core protection calculators (CPCs) and two control element assembly calculators (CEACs). Each CPCS channel calculates the minimum DNBR and maximum LPD in real time using measurement signals for core operating conditions and generates a reactor stop signal when each design limit is violated.

US Patent Nos. 4080251, 4318778, and 4330367 use a digital computer to monitor key fuel design limiting factors LPD and DNBR in real time, and core protection and core monitoring devices and methods for generating reactor stop signals in case of violation of each limit. Describes the invention. Here, the information on the in-position insertion position signal of the individual control rods among the input signals of the core protection calculators (110 to 113; Core Protection Calculator, CPC) that is the core protection device, as shown in Fig. 1, separate control rod calculators 115 and 116; CEAC). Four sets of position signals of the target control rods 117 to 120 of the control element assembly (CEA) groups 121 and 122 are directly input to each of the designated CPC channels 110 to 113. Here, in general, the control rods are defined as the target control rods 117 to 120, each of which is four control rods constitute one group of control rods and transmitted to four CPC channels. Since the core protection device is a safety system to stop the reactor when necessary, it consists of four independent protection system channels. When the reactor stop signal occurs over two channels, the reactor is stopped. In Figure 1, W represents the cooling water pump rotation speed, PR represents the core pressure, TC represents the cold tube temperature, TH represents the hot tube temperature, D represents the nuclear instrument signal.

Therefore, the existing core protection operators (110 to 113) is composed of four independent channels, but the control rod operator (115, 116) is composed of two channels because the installation of the control rod position detector is limited to two control rods. That is, the control rod operators 115 and 116 of the two channels process each individual control rod position signal to transfer the information according to the control rod position deviation to all four channels of the core protection calculators 110 to 113. If the information of one channel of the control rod operator 115, 116 is a false signal due to a failure of the signal connection system or is unavailable, all four channels of the core protection operators 110 to 113 are affected to generate an unnecessary reactor stop signal. Most likely.

Thus, CPCS, a conventional digital reactor thermal protection system, has four channels of core protection operator (CPC) and two channels of control rod operator (CEAC) in hardware. All of the core protection operators in the system had the disadvantage of causing unnecessary reactor shutdowns.

In addition, a number of nuclear reactors have been unexpectedly stopped due to the shortcomings of the existing digital core protection device. In addition, a periodic inspection test for the control rod operators 115 and 116 is required, and since only one channel is in operation during the inspection, a single failure of the control rod operator system 115 or 116 system may cause a reactor shutdown.

In addition, the existing digital core protection device adopts a very simplified calculation model for real-time DNBR calculation, which is too conservative and has a large calculation error. This is due to the fact that digital computers were developed in the early 1970s when the computational performance was low. Increasing DNBR calculation errors and conservative calculation methods reduce the thermal margin of the reactor, reducing the margin to prevent reactor shutdowns in the event of anticipated transient operations or postulated accidents. As a result, power plant utilization is reduced. Therefore, in order to not only improve power plant utilization, but also to secure nuclear fuel design margin and increase reactor output, efforts to increase the thermal margin of the reactor are important. At present, the computational performance of the computer is very good, so it is possible to increase the core thermal margin by applying the more accurate DNBR calculation methodology.

Conventional digital core protection device synthesizes core axial output distribution using off-grid nuclear instrument signal. The existing synthesis method predicts the output distribution shape according to the size of the instrument signal and uses a predetermined synthesis coefficient in each case. In this method, the relationship between the instrument signal and the output distribution is not simple, and in some cases, very incorrect core axial output distribution is synthesized. In other words, the existing core protection system has a problem that the reactor thermal margin is reduced due to the use of a very simple DNBR calculation method and the axial power distribution synthesis method.

On the other hand, the digital core protection device has the function of generating the reactor stop signal for DNBR and LPD, which are the main fuel safety factors, as well as the auxiliary stop signal generation function. In other words, the measured core operating conditions such as core temperature, core pressure, axial output distribution, etc., such as the variable overpower stop function, the asymmetric steam generator transient state prevention function, and the permissible operating range violation function to prevent the sudden increase of the reactor output, It includes a function to generate a reactor stop signal when it is outside the allowable range. The existing core protection device has a preliminary stop alarm function for DNBR and LPD, but it does not have a preliminary stop alarm function for auxiliary stop signals, which prevents the operator from taking measures to prevent reactor shutdown.

The present invention has been proposed to develop an integrated digital reactor thermal protection system applicable to domestic and international loop-type pressurized water reactor nuclear power plant, the four-core core protection operator and the two-channel control rod operator is hardware-connected control rod operator Unlike the conventional digital reactor thermal protection system CPCS, where a failure of one channel affects all core protection operators of four channels, causing unnecessary reactor shutdown, unnecessary reactor shutdown due to a failure of the control rod operator and signal connection equipment. The first purpose is to reduce the

The second object of the present invention is to propose an optimal DNBR calculation method and a reactor axial power distribution synthesis method for increasing the thermal margin of the reactor for securing fuel design margin and increasing reactor power.

The third object of the present invention is to improve power plant performance by improving the function of the reactor thermal protection system to minimize unnecessary reactor shutdowns and reduce operator error and burden.

1 is a block diagram schematically illustrating a conventional digital core protection operator system (CPCS).

Figure 2 is a block diagram schematically showing an integrated reactor thermal protection system according to the present invention.

3 is a detailed view of the integrated real time thermal protector (ITOP) shown in FIG.

Figure 4 is a block diagram schematically showing a control module unit of the integrated reactor thermal protection system according to the present invention.

5A is a diagram illustrating a detailed 4-group model of an optimal DNBR calculation module (Static DNBR, SDNBR) according to an embodiment of the present invention.

FIG. 5B is a quarter detail view of the hot fuel bundle shown in FIG. 5A;

Fig. 6A is a graph showing a speed transfer model in detail 4-group channels.

6B is a graph depicting the pressure transfer model in detail 4-group waterways.

6C is a graph depicting an energy transfer model with detailed 4-group channels.

Explanation of symbols on the main parts of the drawings

10 to 13: integrated real time thermal protector

15: CPU 21: reactor coolant flow rate calculation module

22: control rod calculation function module 23: reactor output distribution synthesis module

24: DNBR calculation module 25: complex function module

26: reactor stop signal processing module

In order to achieve this object, the present invention uses the measurement signal for the core operating conditions according to the information on the control rod position deviation processing the individual control rod position signal, the minimum nuclear boiling rate (DNBR) and the maximum fuel rod linear power density In a real-time reactor protection system for a loop pressurized water reactor using a digital computer that calculates (LPD) in real time and generates a reactor stop signal when each design limit is violated, the core protection operator (CPC) and the control rod operator (CEAC), a CPU that calculates the core core operating factors such as DNBR and LPD in real time and a sub-computation unit that performs calculations for operator linkage functions such as core flow and core output correction and core DNBR operation limit monitoring. Modules and instrument input signals are processed and major real-time calculations are performed on the maintenance test bench (MTP), operator module (OM), A plurality of integrated real-time thermal protectors (ITOP) comprising a common and input / output module (31) for transmitting to the burnout alarm system; Maintenance test bench (MTP) and operator module (OM) that receive the core operation information calculated in real time in each channel of the integrated real-time thermal protector through the intra channel network so that the operator can observe the core operation information in real time. ; It provides integrated real-time reactor thermal protection system that consists of control rod position indicator (CEAPD) that displays the control rod position information received from the maintenance test bench to the operator as pictorial information indicating the state of the rod insertion.

Hereinafter, an integrated real-time reactor thermal protection system according to a preferred embodiment of the present invention will be described in detail with reference to the accompanying drawings.

The present invention is an integrated digital real-time reactor thermal protection system for pressurized water reactors, which is an improvement of the existing digital reactor thermal protection system CPCS, and consists of an operating module containing hardware using a digital computer and software for operating the same. As shown in FIG. 2, the hardware removes a separate control rod operator for controlling the control rod position signal and transfers the function to the integrated thermal on-line protector (ITOP). It consists of an integrated nuclear reactor thermal protection system.

That is, the thermal protection system of the present invention directly connects two sets of control rod position signals RSPT1 & RSPT2 to the respective ITOPs 10, 11, 12, and 13 using the CEA input / output module 32 and the ITOP (10). Each ITOP 10, 11, 12, 13 performs the function of the control rod operator using the software 11, 12, 13). The core information calculated in real time on each channel of ITOP (10, 11, 12, 13) is maintained in the maintenance and test panel (MTP) and the operator module (28; operator) through the optical fiber intra channel network, respectively. Module, OM) enables the operator to observe the core operation information in real time. In addition, the control rod position information from the MTP 27 is transmitted to the control rod position display device 29 (CEA Position Display, CEAPD) to show the operator in the furnace insert information of the control rod as a picture. Here, the signal isolation device is used to maintain the independence between the control rod position signal and each ITOP (10, 11, 12, 13) channel. In order to achieve this, the ITOPs 10, 11, 12, and 13 use a high-performance central processing unit 15 (CPU) as shown in FIG.

Therefore, the main core operation factors (DNBR and LPD, etc.) are calculated in real time using the CPU 15, and the auxiliary operation unit module 30 monitors the operator linkage function (ie core flow rate and core output correction, and core DNBR operation limit monitoring). In order to facilitate the operation), a predetermined calculation is performed, such as automatically calculating the correction rate constant of the core flow rate of the reactor coolant flow rate calculation module 21 and the core output of the combined function module 25. At this time, the correction constant is calculated by comparing the flow rate calculated in the cooling water flow rate calculation module 21 with the reference flow rate for the core flow rate correction. The auxiliary operation module 30 also includes a monitoring device configured to automatically perform a comparison of the minimum DNBR calculated in real time in the composite function module 25 and the preset DNBR operating limit value to monitor the DNBR operating limit value.

Meanwhile, the common module (COM.) And the input / output module (I / O) process the input signal of the instrument as shown by reference numeral 31, and the main real-time calculation results are obtained by MTP 27 and OM ( 28) and transmit to the plant alarm system.

3 is a detailed view of ITOP channel A 10 and shows an input signal and an output signal. Compared to the input signal of the existing core protection operators (110 to 113) of the channel A (110) of FIG. 1, the individual control rod position signal (Reed Switch Position Transmitter, instead of the control rod group position signal) It can be seen that RSPT1 & RSPT2) are used and supply power frequency (??) and current signal (I) are used instead of the cooling water pump rotation speed.

The functional module portion of the integrated reactor thermal protection system is composed of five calculation function modules and one reactor stop signal processing module as shown in FIG. The reactor coolant flow calculation module 21 calculates the core flow rate using the measured coolant pump supply power frequency and the input current signal, and transmits the core flow rate to the reactor output distribution synthesis module 23 and the DNBR calculation module 24. . In addition, the minimum DNBR and the maximum LPD value are input from the combined function module 25 to perform correction calculation according to the core flow rate change, and transmit the modified minimum DNBR and the maximum LPD value to the reactor stop signal processing module 26. The control rod calculation function module 22 (CEA Deviation, CRDEV) processes the individual control rod position signals to calculate the DNBR and LPD penalty factors according to the position deviation. DNBR and LPD penalty factors due to the control rod position deviation are transmitted to the composite function module 25 (Update Variable, UPDVAR), and the target control rod position information is transmitted to the reactor output distribution synthesis module 23. The composite function module 25 calculates core output using various instrument input signals and core output distribution related information, and calculates DNBR and LPD at short time intervals to reflect the reactor operating conditions that change frequently. DNBR and LPD modified by UPDVAR are transmitted to the reactor stop signal processing module 26. The reactor power distribution synthesis module (23; Core Power Distribution, CPWRD) synthesizes the core axial and radial output distributions using the core flow rate calculated from CFLOW (21) and off-core nuclear instrument signals transmitted from UPDVAR (25). . The core axial and radial output distributions synthesized in real time are sent to the DNBR calculation module 24 and the composite function module 25. Static DNBR and SDNBR perform a detailed DNBR calculation using the core flow rate calculated in CFLOW 21 and the core output distribution synthesized in CPWRD 23, and send them to UPDVAR 25. Finally, the reactor stop signal processing module 26 (Trip Generation, TRPGEN) determines whether to generate the reactor stop signal and the auxiliary stop signal of DNBR and LPD. It also processes preliminary stop signals from DNBR and LPD to generate an alarm if necessary.

Here, the reactor coolant flow rate calculation module 21 calculates the core flow rate using the frequency and three-phase current signals of the supply power instead of the rotational speed signal of the conventional coolant pump. That is, the core flow rate uses a principle that is proportional to the output of the pump supply power. The change of the input signal makes it possible to recognize the coolant pump shaft fixation accident and the shaft break accident more quickly. Therefore, it is possible to reduce the thermal margin required in case of an accident and the use of the three-phase current signal multiplexes the input signal. There is also.

The control rod operation function module 22 (CEA Deviation, CRDEV) is an improvement of the function module of the integrated thermal protector (ITOP) in the present invention to perform the function of the existing control rod operation. The individual control rod position signals are processed to calculate DNBR and LPD penalty factors according to position deviations, and the group control rod position signals are also determined and provided to the output distribution synthesis module. If one of the two sets of control rod position input signals is unavailable, only the signals that are available for a certain period of time are used. If the control rod position signals are all disabled, the function of the CRDEV (22) module is automatically bypassed and the DNBR and LPD penalty factors use the previous values.

The reactor output distribution synthesis module 23 synthesizes the core axial and radial output distributions. The axial output distribution refines the existing third-order spline synthesis method to refine the output distribution characteristics and use the optimal synthesis coefficient. The radial output distribution is determined using the group control rod positions provided by the CRDEV module 22 and synthesizes a hypothetical hot-pin output distribution for conservative LPD and DNBR calculations.

DNBR calculation module 24 obtains the numerical solution of the reactor thermal flow governing equation and calculates the core minimum DNBR using a detailed four-channel analysis model as shown in FIG. 5A. That is, the flow path between the nuclear fuel rod bundles in the reactor where the coolant flows is divided into a core average channel 41 and a high temperature fuel bundle channel 42, and the inside of the channel 42 is a high temperature channel 44 assuming that a minimum DNBR is generated. And the high temperature waterway and the surrounding waterway 43. The detailed four-channel channel was further modeled by the boundary channel 42 'and the boundary channel 42 "to more accurately and accurately calculate the momentum and energy transfer between the group channel 43 and the channel 42. In this case, the channel 42 is an average channel representing channels without any marking except for the part indicated by 44, 43, 42 'or 42 "in FIG. 5B, and the detailed 4-channel is a boundary channel. The addition of 42 'and the boundary channel 42 "means that the modeling can be further refined to allow more detailed and accurate calculation of the momentum and energy transfer throughout the channel.

The momentum and energy transfer between the group channels is further defined in real time according to the reactor operating conditions by further defining the boundary channels between the group channels. The existing simplified 4-channel analysis method does not define the boundary channel but uses a transfer coefficient with a constant value determined through coordination with the more detailed DNBR analysis method.

6A-6C schematically show the flow rate (velocity), pressure and enthalpy exchange between adjacent channels for solving axial and transverse momentum and energy equations, respectively, and the pressure, velocity and enthalpy transfer coefficients between adjacent channels. N _P , N _U , N _H ) is defined as in Equation 1 below.

Where P, U, and H represent the mean pressures, velocities, and enthalpies of I and J in any adjacent integral water, and p, u, and h represent the pressure, velocity, and enthalpy of the boundary water between I and J as integrated water. In the detailed four-channel analysis model used in the present invention, each of the transfer coefficients defined above is mainly used to exchange pressure, velocity, and enthalpy between the hot fuel bundle channel 42 and the peripheral channel 43 of the high temperature channel. It is defined to represent. The transfer coefficients N _P and N _U for pressure and velocity use pre-calculated optimal constant values, and only the enthalpy transfer coefficient (N _H ) is obtained using the enthalpy of the boundary channel (42 'and 42 "). Calculate it together.

UPDVAR (25; Update Variable), a multi-functional module, calculates modified DNBR by reflecting reactor output and operating conditions that change frequently. The UPDVAR 25 calculates and compares the respective auxiliary stop factors for variable overpower stop, asymmetric steam generator transient protection and allowable operating range violation, and compares them with the respective reactor stop limits. Existing core protection device has a disadvantage that the operator can not take action to prevent the reactor stop because there is no pre-stop alarm function for the auxiliary stop signal, in the present invention, the pre-stop alarm according to all the auxiliary stop factors The module is additionally installed in the UPDVAR to transmit the result of comparing the auxiliary stop factor and the reactor stop limit to the reactor stop signal processing module 26.

The reactor stop signal processing module 26 determines whether to generate the reactor stop signal and the auxiliary stop signal of the DNBR and LPD. Furthermore, in the present invention, the operator can process the preliminary stop signal of all reactor stop variables to generate an alarm if necessary. Take necessary reactor shutdown precautions.

The integrated reactor thermal protection system configured in this way provides the operator with the reactor power and coolant flow rate calculated in real time to automate the calibration of reactor power and coolant flow to improve operator connectivity. In addition, monitoring of DNBR normal operating limits can be automated if the core monitoring system is inoperable.

Therefore, according to the digital reactor thermal protection system according to the present invention, by combining the existing control rod operator (CEAC) and the core protection operator (CPC) can simplify the system and the possibility of reactor shutdown due to a single failure of the connection system It can be greatly reduced.

In addition, by optimizing the DNBR calculation method, which is a major operating variable of the reactor, it is possible to increase the thermal margin of the reactor, thereby increasing the reactor margin and nuclear fuel design margin, and securing a foundation for increasing the reactor output. In addition, by adding a preliminary stop alarm function of all auxiliary stop factors, the reactor operator can take measures to prevent unnecessary reactor shutdowns.Automated linkage with the reactor operator reduces the operator's mistakes and burdens. Will be able to improve.

Although the present invention has been shown and described with respect to specific embodiments, it will be apparent to those skilled in the art that various modifications and variations can be made without departing from the spirit and scope of the invention as set forth in the claims. It will be easy to see.

Claims (6)

According to the information on the control rod position deviation processing individual control rod position signals, the minimum nuclear breakaway rate (DNBR) and the maximum fuel rod linear power density (LPD) are calculated in real time using the measurement signals for the core operating conditions. In a real-time reactor protection system for a loop-type pressurized water reactor using a digital computer that generates a reactor stop signal when a design limit is violated, CPU 15, which calculates the core operating parameters such as DNBR and LPD in real time by integrating the functions of core protection operator (CPC) and control rod operator (CEAC) into one, core flow and core output correction, and core DNBR operation limit monitoring Auxiliary operation unit module 30, which performs calculations for operator linkage functions, processes instrument input signals, and sends the main real-time calculation results to maintenance test benches (27; MTP), operator modules (28; OM), and power plant alarm systems. A plurality of integrated real-time thermal protectors (10, 11, 12, 13; ITOP) comprising a common and input / output module (31) for transmitting; Maintenance test bench that allows the operator to observe the core operation information in real time by receiving the core operation information calculated in real time in each channel of the integrated real-time thermal protector (10, 11, 12, 13) through the intra channel network An MTP 27 and an operator module 28 OM; A control rod position display device (CEAPD) 29 for displaying the control rod position information received from the maintenance test bench 27 to the operator as picture information indicating an insertion state of the control rod; Integrated real-time reactor thermal protection system, characterized in that consisting of. According to claim 1, Each of the integrated real-time thermal protector (10, 11, 12, 13) calculates the core flow rate using the measured cooling water pump supply power frequency and input current signal to the reactor output distribution synthesis module (23) and DNBR calculation module (24) Transmits the modified minimum DNBR and the maximum LPD value to the reactor stop signal processing module 26 by performing the correction calculation according to the core flow rate change by receiving the minimum DNBR and the maximum LPD value from the composite function module 25. A reactor coolant flow calculation module 21 (Core Flow, CFLOW); Calculate DNBR and LPD penalty factors according to position deviations by processing individual control rod position signals, transfer DNBR and LPD penalty factors due to control rod position deviations to the composite function module 25, and distribute the target control rod position information to the reactor output distribution. A control rod arithmetic function module 22 (CEA Deviation, CRDEV) which is transmitted to the synthesis module 23; Core axial and radial directions synthesized by synthesizing the core axial and radial output distributions using the core flow rate calculated by the reactor coolant flow rate calculation module 21 and the core nuclear instrument signal transmitted from the composite function module 25. Reactor output distribution synthesis module 23 for delivering the output distribution to the DNBR calculation module 24 and the composite function module 25 in real time (Core Power Distribution, CPWRD); DNBR calculation module for performing detailed DNBR calculation using the core flow rate calculated in the reactor coolant flow rate calculation module 21 and the core output distribution synthesized in the reactor output distribution synthesis module 23 to transfer to the composite function module 25. (24; Static DNBR, SDNBR); A complex function module (25; Update Variable, UPDVAR) that calculates core output using various instrument input signals and core output distribution related information and calculates DNBR and LPD at short time intervals to reflect the reactor operating conditions that change frequently; And A reactor stop signal processing module 26 (Trip Generation, TRPGEN) is configured to determine whether to generate the reactor stop signal and the auxiliary stop signal of the DNBR and LPD and to generate an alarm when necessary by processing the preliminary stop signal of the DNBR and LPD. Integrated real-time reactor thermal protection system comprising a control module. The method of claim 2, The DNBR calculation module 24 and SDNBR perform real-time optimal DNBR calculation using a detailed 4-channel analysis model in which boundary channels 2 'and 2 "are set between the group channel 43 and the channel 42. Integrated real-time reactor thermal protection system, characterized in that it can be performed by. The method of claim 2, The composite function module 25 is configured to issue a preliminary stop alarm by calculating the relevant auxiliary stop factors for variable overpower stop, asymmetric steam generator transient protection and permissible operating range violation, and comparing them with the respective reactor stop limits. Digital real-time reactor thermal protection system, characterized in that it comprises an alarm module. According to claim 1, The auxiliary operation module 30 is configured to automatically calculate the correction constant of the core flow rate of the reactor coolant flow rate calculation module 21 and the core output of the composite function module 25, the digital real-time reactor thermal protection system. According to claim 1, The auxiliary operation module 30 is provided with a monitoring device that can automatically perform comparison of the minimum DNBR calculated in real time in the composite function module 25 and the preset DNBR operating limit value in order to monitor the DNBR operating limit value. Digital real-time reactor thermal protection system, characterized in that.