CN114091237B - Method for identifying key sensitive equipment of digital reactor protection system - Google Patents

Method for identifying key sensitive equipment of digital reactor protection system Download PDF

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CN114091237B
CN114091237B CN202111222715.8A CN202111222715A CN114091237B CN 114091237 B CN114091237 B CN 114091237B CN 202111222715 A CN202111222715 A CN 202111222715A CN 114091237 B CN114091237 B CN 114091237B
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protection system
reactor protection
fault tree
spv
equipment
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CN114091237A (en
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黄艇博
祁军
尚宪和
任诚
韩雨
于渭清
吴剑
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CNNC Nuclear Power Operation Management Co Ltd
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    • G—PHYSICS
    • G06—COMPUTING OR CALCULATING; COUNTING
    • G06F—ELECTRIC DIGITAL DATA PROCESSING
    • G06F30/00—Computer-aided design [CAD]
    • G06F30/20—Design optimisation, verification or simulation
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D3/00—Control of nuclear power plant
    • G21D3/001—Computer implemented control
    • G—PHYSICS
    • G06—COMPUTING OR CALCULATING; COUNTING
    • G06F—ELECTRIC DIGITAL DATA PROCESSING
    • G06F2119/00—Details relating to the type or aim of the analysis or the optimisation
    • G06F2119/02—Reliability analysis or reliability optimisation; Failure analysis, e.g. worst case scenario performance, failure mode and effects analysis [FMEA]
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin

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  • Physics & Mathematics (AREA)
  • General Engineering & Computer Science (AREA)
  • Theoretical Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Computer Hardware Design (AREA)
  • Evolutionary Computation (AREA)
  • Geometry (AREA)
  • General Physics & Mathematics (AREA)
  • Testing And Monitoring For Control Systems (AREA)

Abstract

The invention belongs to the field of nuclear power, and particularly relates to a method for identifying key sensitive equipment of a digital reactor protection system of a nuclear power plant. The relevance among all clamping pieces of the digital reactor protection system is strong and has more complex logic, and the influence of the fault of the clamping piece on the whole system is difficult to comprehensively analyze by an FMEA analysis method. The method is complementary with the traditional SPV identification method, and provides a rapid and reliable temporary SPV equipment identification method for the digital reactor protection system for the nuclear power plant. The method mainly comprises the steps of establishing a fault tree model of a reactor protection system in a zero maintenance test, wherein the obtained first-order cut set related equipment is SPV equipment; establishing boundary conditions of corresponding tests according to periodic test rules of a reactor protection system; and (3) establishing a temporary boundary condition, and comparing the temporary boundary condition with a first-order cut set obtained by a fault tree model of a zero maintenance zero repair test, wherein equipment corresponding to the newly added first-order cut set is the temporary SPV of the system in the current state.

Description

Method for identifying key sensitive equipment of digital reactor protection system
Technical Field
The invention belongs to the field of nuclear power, and particularly relates to a method for identifying key sensitive equipment of a digital reactor protection system.
Background
With the advancement of the reform of the domestic electric power market system, higher requirements are put on the competitiveness of nuclear power plants in the electric power market. Therefore, the reliability level of the nuclear power plant system equipment is improved, and the reduction of unplanned shutdown and shutdown of the unit becomes important content of each power plant. An important content of the device reliability management system is device classification, and the highest importance in device classification management is SPV (critical sensitivity) devices, and the method for identifying SPV devices mainly adopts FMEA (failure mode and impact analysis) and operation experience at present. However, as the digitalized reactor protection system is widely adopted in newly built nuclear power plants in China, as the relativity among clamping members of the digitalized reactor protection system is strong and has more complex logic, the influence of clamping member faults on the whole system is difficult to comprehensively analyze by an FMEA analysis method, so that the invention establishes a logic relationship between the clamping member fault mode and misoperation of the reactor protection system by a fault tree method and solves the defects of the FMEA analysis method.
Disclosure of Invention
1. The purpose is as follows:
A new method is provided for the identification of SPV equipment (key sensitive equipment) of a digital reactor protection system of a nuclear power plant, and the method is complementary with the traditional SPV identification method; a rapid and reliable temporary SPV equipment identification method for a digital reactor protection system is provided for a nuclear power plant.
2. The technical scheme is as follows:
The method for identifying key sensitive equipment of the digital reactor protection system specifically comprises the following steps: step one: determining the analysis boundary of a malfunction fault tree of the digital reactor protection system; step two: establishing a fault tree model taking the power failure of a control rod caused by tripping of a shutdown circuit breaker as a top event; step three: establishing a boundary condition of a corresponding fault tree for periodic test of a reactor protection system; step four: step two and step three establish the digital reactor protection system real-time fault tree model together; step five: calculating a first-order cut set basic event corresponding device obtained by the fault tree model in the second step, and determining SPV (specific surface wave) equipment; step six: establishing boundary conditions of temporary SPV calculation of a digital reactor protection system; step seven: and step five and step six, calculating together to obtain the temporary SPV equipment under the current system configuration.
The first step is as follows: determining the analysis boundary of a malfunction fault tree of a digital reactor protection system, which specifically comprises the following steps: the outer boundary of a typical digital reactor protection system comprises a shutdown breaker as an actuating mechanism, a digital reactor protection system computer, a signal processing cabinet and an in-situ transmitter; the inner boundary is a functional clamping piece in each cabinet.
The second step is as follows: the method for establishing the fault tree model by taking the event that the shutdown circuit breaker trips to cause the control rod to lose electricity as a top event specifically comprises the following steps: drawing a system diagram within the analysis range of the fault tree; a fault tree model is built that takes a shutdown circuit breaker trip causing a control rod to lose power as a top event.
And step three: establishing a boundary condition of a corresponding fault tree for periodic test of a reactor protection system, which specifically comprises the following steps: and determining a configuration state and setting boundary conditions according to the test procedure.
And step four: step two and step three establish the real-time fault tree model of the digital reactor protection system together, include specifically: and embedding the boundary conditions set in the third step into the zero-maintenance zero-repair test fault tree model in the second step, and selecting corresponding boundary conditions according to the real-time ongoing test of the power plant to obtain the real-time fault tree model of the digital reactor protection system.
Step five, the said step: calculating a first-order cut set basic event corresponding device obtained by the fault tree model in the second step, and determining SPV equipment, wherein the method specifically comprises the following steps: calculating the top event of the fault tree obtained in the second step, and setting first-order cut-off during calculation to obtain N (N is more than or equal to 1) first-order cut sets, wherein the equipment corresponding to the N first-order cut sets basic events is the SPV equipment of the digital reactor protection system.
Step six, the step is as follows: establishing boundary conditions of temporary SPV calculation of a digital reactor protection system, which specifically comprises the following steps: setting all the basic events in the first order cut set calculated in the fifth step as FALSE, namely, considering that the basic events do not occur when calculating the temporary SPV, calculating the fault tree top event under the boundary condition, and setting no first order cut set.
Step seven, the method comprises the following steps: step five and step six calculate together and get the temporary SPV apparatus under the present system configuration, include specifically: and selecting corresponding boundary conditions according to the current test operation of the system, and simultaneously selecting temporary SPV calculation boundary conditions, wherein equipment corresponding to the first-order cut set obtained by calculating the fault tree top event is temporary SPV equipment under the current system configuration.
3. The effect is as follows:
And establishing a logic relation between a fault mode of each clamping piece or equipment in the reactor protection system and the power failure of the control rod caused by tripping of the shutdown circuit breaker through a fault tree, and obtaining equipment corresponding to the first-order cutting set, namely SPV equipment, through cutting set calculation on the fault tree.
The configuration state of the reactor protection system caused by defects, maintenance, experiments and the like is changeable, if the temporary SPV equipment is identified by adopting the FMEA method, the comprehensive analysis needs to be carried out once under each system configuration state, but the configuration state is not exhaustive due to the diversity of the system configuration states; by adopting the analysis method of the fault tree model, only the failure mode of the equipment which causes the system configuration state to change is analyzed, and the influence of the failure of the control rod caused by tripping of the shutdown circuit breaker can be transmitted through the established logic relationship in the fault tree without re-analysis, so that the time for identifying the temporary SPV equipment can be greatly shortened.
Drawings
FIG. 1 is a flow chart for identifying key sensitive equipment of a digital reactor of a nuclear power plant;
FIG. 2 illustrates a typical connection of a shutdown circuit breaker for a nuclear power plant.
Detailed Description
The invention mainly comprises the following steps:
Establishing a zero maintenance zero repair test reactor protection system fault tree model taking a control rod power failure caused by tripping of a shutdown circuit breaker as a top event;
Establishing boundary conditions of corresponding tests according to periodic test rules of a reactor protection system;
Establishing a temporary boundary condition for system configuration state change caused by reactor protection system defects or defect maintenance;
calculating a first-order cut set related device obtained by the zero maintenance zero test fault tree model to obtain SPV equipment;
according to the current test and the existing defect or the maintenance aiming at the defect of the reactor protection system, a corresponding boundary condition is selected to establish a boundary condition set, a fault tree under the boundary condition is calculated to obtain a new first-order cut set, the new first-order cut set is compared with the first-order cut set obtained by the fault tree model of the zero-maintenance test, and equipment corresponding to the new first-order cut set is the temporary SPV of the system in the current state.
The following description is made with reference to specific examples and drawings:
Firstly, determining the analysis boundary of a malfunction fault tree of a digital reactor protection system: taking a typical digital reactor protection system as an example, the outer boundary of the digital reactor protection system comprises a shutdown breaker serving as an executing mechanism, a digital reactor protection system computer, a signal processing cabinet and an in-situ transmitter; the inner boundary is a functional clamping piece in each cabinet.
The second step is to build a fault tree model taking the event that the shutdown breaker trips to cause the control rod to lose electricity as a top event, and the specific process is as follows:
1. A system diagram is drawn within the scope of fault tree analysis, taking a typical digital reactor protection system as an example, comprising: eight shutdown breakers serving as an actuating mechanism are divided into four channels, two breakers of each channel are used for realizing channel four-out and two-shutdown logic in a special connection mode, and the logic is shown in a figure II; the instrument control part comprises an in-situ transmitter, an analog signal adjusting module SAA1, a platinum resistance signal converting module STT1, a standard signal expanding module SNV1, an overvoltage protecting module SOB1, a switching value signal adjusting module SBC1, an analog signal input module SAI1, a switching value input module SDI1, a CPU module SVE2, a communication module SL22, a photoelectric converting module SLM2, a switching value output module SDO1 and a relay module SRB1, and the logic calculating part of each channel is provided with two independent subgroups; the four channels are communicated through optical fibers, and signal transmission is carried out only in the corresponding sub-group; each channel will perform 2/4 logic or 2/3 logic (related to the number of in-situ transmitter channels) of the shutdown signal.
2. A fault tree model is built with the control rod losing electricity caused by tripping of the shutdown circuit breaker as a top event, and shutdown signals are taken as examples of low main pump rotation speed (a subgroup a, 3 channels of a field transmitter) and main pump tripping (a subgroup b, 4 channels of a field contact).
Thirdly, calculating the top event of the fault tree obtained in the second step, and setting first-order cut-off during calculation to obtain a plurality of first-order cut sets, wherein the equipment corresponding to the basic event of the first-order cut sets is the SPV equipment of the digital reactor protection system.
And fourthly, establishing a boundary condition of temporary SPV calculation of the digital reactor protection system, setting all basic events in the first-order cut set obtained by the third step as FALSE, namely, considering that the basic events do not occur when the temporary SPV is calculated, calculating a fault tree top event under the boundary condition, and leaving no first-order cut set.
Fifth, a reactor protection system periodic test is established corresponding fault tree boundary conditions for rapid system state configuration during these periodic tests. Taking the following general configuration during the test of the reactor protection system as an example, corresponding boundary conditions are established:
1. according to the test rules, the reactor protection system test is to send a trip signal of a single shutdown breaker one by one through an engineer station to verify the action condition of the shutdown breaker, so the test is divided into eight configuration states, namely eight trip states of the shutdown breaker;
2. When a trip test of the No. 1 shutdown breaker is carried out, two gates of a Q1 trip signal output by a processor module of the 1-YFTB-A1-SVE2-A-Q1 shutdown protection system A1 channel diversity a and a Q1 trip signal output by a processor module of the 1-YFTB-A1-SVE2-B-Q1 shutdown protection system A1 channel diversity B in the corresponding fault tree are true, so that the boundary conditions are set as follows:
1-YFTB-A1-SVE2-A-Q1 TRUE
1-YFTB-A1-SVE2-B-Q1 TRUE
similarly, when the trip test of the No.2 shutdown breaker is carried out, the boundary conditions are set as follows:
1-YFTB-A1-SVE2-A-Q2 TRUE
1-YFTB-A1-SVE2-B-Q2 TRUE
when the trip test of the 3# shutdown breaker is carried out, the boundary conditions are set as follows:
1-YFTB-A2-SVE2-A-Q3 TRUE
1-YFTB-A2-SVE2-B-Q3 TRUE
when the trip test of the No. 4 shutdown circuit breaker is carried out, the boundary conditions are set as follows:
1-YFTB-A2-SVE2-A-Q4 TRUE
1-YFTB-A2-SVE2-B-Q4 TRUE
When the trip test of the No. 5 shutdown breaker is carried out, the boundary conditions are set as follows:
1-YFTB-B1-SVE2-A-Q5 TRUE
1-YFTB-B1-SVE2-B-Q5 TRUE
When the trip test of the No. 6 shutdown breaker is carried out, the boundary conditions are set as follows:
1-YFTB-B1-SVE2-A-Q6 TRUE
1-YFTB-B1-SVE2-B-Q6 TRUE
When the trip test of the 7# shutdown breaker is carried out, the boundary conditions are set as follows:
1-YFTB-B2-SVE2-A-Q7 TRUE
1-YFTB-B2-SVE2-B-Q7 TRUE
when the trip test of the 8# shutdown breaker is carried out, the boundary conditions are set as follows:
1-YFTB-B2-SVE2-A-Q8 TRUE
1-YFTB-B2-SVE2-B-Q8 TRUE
And sixthly, calculating the temporary SPV equipment under the current system configuration. And selecting corresponding boundary conditions according to the current test operation of the system, and simultaneously selecting temporary SPV calculation boundary conditions, wherein equipment corresponding to the first-order cut set obtained by calculating the fault tree top event is temporary SPV equipment under the current system configuration.

Claims (3)

1. A method for identifying key sensitive equipment of a digital reactor protection system is characterized by comprising the following steps of:
step one: determining the analysis boundary of a malfunction fault tree of the digital reactor protection system;
step two: establishing a zero maintenance zero repair test fault tree model taking a control rod power failure caused by tripping of a shutdown circuit breaker as a top event;
the method specifically comprises the following steps: drawing a system diagram within the analysis range of the fault tree; establishing a fault tree model taking the power failure of a control rod caused by tripping of a shutdown circuit breaker as a top event;
Step three: establishing a boundary condition of a corresponding fault tree for periodic test of a reactor protection system;
Step four: step two and step three establish the digital reactor protection system real-time fault tree model together;
The method specifically comprises the following steps: embedding the boundary conditions set in the third step into the zero-maintenance zero-repair test fault tree model in the second step, and selecting corresponding boundary conditions according to the real-time ongoing test of the power plant to obtain a digital reactor protection system real-time fault tree model;
step five: calculating a first-order cut set basic event corresponding device obtained by the fault tree model in the second step, and determining SPV (specific surface wave) equipment;
The method specifically comprises the following steps: calculating the top event of the fault tree obtained in the second step, and setting first-order cut-off during calculation to obtain N (N is more than or equal to 1) first-order cut sets, wherein equipment corresponding to the N first-order cut set basic events is SPV equipment of the digital reactor protection system;
step six: establishing boundary conditions of temporary SPV calculation of a digital reactor protection system;
The method specifically comprises the following steps: setting all basic events in the first-order cut set obtained by calculation in the step five as FALSEs, namely, considering that the basic events cannot occur when calculating the temporary SPV, calculating fault tree top events under the boundary condition, and leaving no first-order cut set;
Step seven: step five and step six calculate and get the temporary SPV apparatus under the present system configuration together;
The method specifically comprises the following steps: and selecting corresponding boundary conditions according to the current test operation of the system, and simultaneously selecting temporary SPV calculation boundary conditions, wherein equipment corresponding to the first-order cut set obtained by calculating the fault tree top event is temporary SPV equipment under the current system configuration.
2. The method for identifying key sensitive equipment of a digital reactor protection system according to claim 1, wherein the method comprises the following steps: the first step is as follows: determining the analysis boundary of a malfunction fault tree of a digital reactor protection system, which specifically comprises the following steps: the outer boundary of a typical digital reactor protection system comprises a shutdown breaker as an actuating mechanism, a digital reactor protection system computer, a signal processing cabinet and an in-situ transmitter; the inner boundary is a functional clamping piece in each cabinet.
3. The method for identifying key sensitive equipment of a digital reactor protection system according to claim 1, wherein the method comprises the following steps: and step three: establishing a boundary condition of a corresponding fault tree for periodic test of a reactor protection system, which specifically comprises the following steps: and determining a configuration state and setting boundary conditions according to the test procedure.
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CN109358583A (en) * 2018-10-23 2019-02-19 中核核电运行管理有限公司 A method of preventing the failure of nuclear power unit key work center method

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KR100848881B1 (en) * 2006-08-07 2008-07-29 삼창기업 주식회사 Digital reactor protection system
JP5480033B2 (en) * 2010-06-23 2014-04-23 行政院原子能委員會核能研究所 Quantitative risk assessment method using computer-aided top logic for nuclear power plants.
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CN109358583A (en) * 2018-10-23 2019-02-19 中核核电运行管理有限公司 A method of preventing the failure of nuclear power unit key work center method

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