WO2023197462A1 - 核电厂机组事故工况监测方法和系统 - Google Patents

核电厂机组事故工况监测方法和系统 Download PDF

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Publication number
WO2023197462A1
WO2023197462A1 PCT/CN2022/103208 CN2022103208W WO2023197462A1 WO 2023197462 A1 WO2023197462 A1 WO 2023197462A1 CN 2022103208 W CN2022103208 W CN 2022103208W WO 2023197462 A1 WO2023197462 A1 WO 2023197462A1
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Prior art keywords
accident
unit
typical
conditions
loss
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PCT/CN2022/103208
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English (en)
French (fr)
Inventor
吴月军
李敏
王泰科
马廷伟
刘志云
孙晨
刘玉华
刘琉
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深圳中广核工程设计有限公司
中广核工程有限公司
中国广核集团有限公司
中国广核电力股份有限公司
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Publication of WO2023197462A1 publication Critical patent/WO2023197462A1/zh

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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D3/00—Control of nuclear power plant
    • G21D3/001—Computer implemented control
    • G21D3/002—Core design; core simulations; core optimisation
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D3/00—Control of nuclear power plant
    • G21D3/008—Man-machine interface, e.g. control room layout
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D3/00—Control of nuclear power plant
    • G21D3/04—Safety arrangements
    • G21D3/06—Safety arrangements responsive to faults within the plant
    • 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
    • Y02E30/30—Nuclear fission reactors

Definitions

  • the invention belongs to the field of accident monitoring of nuclear power plant units. More specifically, the invention relates to a method and system for monitoring accident conditions of nuclear power plant units.
  • a known nuclear power technology adopts a condition-oriented or symptom-oriented accident procedure.
  • the operator mainly performs the accident procedure based on the deterioration of a limited number of unit status parameters that represent the three major safety functional states of the reactor. He lacks understanding of the cause of the accident and cannot It is conducive to the overall understanding of the accident, affects the efficiency of accident handling, and is easy to get lost.
  • the object of the present invention is to provide a nuclear power plant unit accident condition monitoring method and system that can automatically diagnose the initial accident and superimposed accident of the unit and display the cause of the accident.
  • the present invention provides a method for monitoring accident conditions of nuclear power plant units.
  • the method includes the following steps:
  • Each logical computing unit performs logic on the abnormal characteristic parameters, important safety signals of the unit, and the status of special safety facilities related to the logical computing unit according to the preset logic. Calculate and diagnose whether typical accident conditions corresponding to the logical calculation unit occur;
  • the diagnosis results of all typical accident conditions are displayed on the automatic diagnosis screen for unit accident conditions.
  • the typical accident conditions include typical thermal hydraulic accidents, loss of support function accidents and spent fuel pool events; where typical thermal hydraulic accidents include steam Generator heat transfer tube rupture accident, primary circuit breach accident, primary circuit reactive accident and secondary circuit breach accident; accidents of loss of support function include accidents of loss of external power supply, plant-wide power loss accident, electrical panel power loss accident, complete Cold chain loss accidents and waste heat removal system failures; spent fuel pool incidents include spent fuel pool loss of cooling accidents and spent fuel pool water loss accidents;
  • At least one logical calculation unit is set up corresponding to each typical accident condition, and the logical calculation units of different typical accident conditions run in parallel to perform logical calculations on different typical accident conditions.
  • the steam generator heat transfer tube rupture accident logic calculation unit is used to diagnose whether a steam generator heat transfer tube rupture accident occurs, and one of the following situations occurs: Diagnose the steam generator heat transfer tube rupture accident that has occurred:
  • the radioactivity measurement value of the steam pipeline on the secondary side of the steam generator is greater than the high-high threshold or the radioactivity measurement value of the steam generator wastewater sampling is greater than the high-high threshold;
  • the radioactivity measurement value of the steam pipeline on the secondary side of the steam generator is greater than the high threshold or the radioactivity measurement value of the steam generator sewage sampling is greater than the high threshold, and the liquid level of the superimposed primary loop pressure regulator or the capacity control box is low.
  • the primary circuit breach accident is divided into a breach in the containment and a breach in the safety building according to the location of the breach.
  • Two accident logics are used The computing unit diagnoses the two breach accidents respectively;
  • intra-containment breach accident logic calculation unit uses the intra-containment breach accident logic calculation unit to diagnose whether an intra-containment breach accident has occurred. If one of the following conditions occurs, it is diagnosed that an intra-containment breach accident has occurred: (a) The measured value of the dose rate in the containment reaches a high High threshold; (b) The dose rate measurement value in the containment reaches the high threshold value, and the superimposed pressure measurement value in the containment vessel reaches the high threshold value; (c) When the primary circuit is closed, the pressure measurement value in the containment vessel reaches the high threshold value, and there is a Low circuit subcooling or low pressure vessel liquid level signal;
  • the logic for diagnosing that a safe factory breach accident has occurred is: the residual heat exhaust system allowed connection signal has taken effect, and the safe factory pressure measurement value is greater than the high threshold. Or the measured value of the liquid level in the safety workshop pit is greater than the high threshold.
  • the secondary circuit breach accident logical calculation unit is used to diagnose whether a secondary circuit breach accident occurs. If one of the following conditions occurs, the secondary circuit breach accident is diagnosed. It has happened:
  • the automatic isolation signal of the isolation valve of one main steam release system exists, and the automatic isolation signal of the isolation valve of the three main steam release systems does not exist;
  • the internal pressure difference between the two steam generators is high, and the allowed connection signal of the waste heat discharge system does not take effect, and the automatic isolation signal of the isolation valve of the three-row main steam release system is superimposed.
  • the primary circuit reactive accident logic calculation unit is used to diagnose whether a primary circuit reactive accident occurs. If one of the following situations occurs, the primary circuit reactive accident is diagnosed. It has happened:
  • the measured value of total boron concentration in the primary loop is lower than the low threshold
  • the neutron fluence rate in the middle range is high;
  • the neutron fluence rate between sources is high;
  • the shutdown neutron fluence rate is high.
  • the diagnostic logic of the loss of support function accident is:
  • the occurrence of the off-site power supply loss accident is determined by the existence of the corresponding factory AC bus voltage loss alarm;
  • the occurrence of the plant-wide power outage accident can be judged by the existence of corresponding emergency bus and non-emergency bus voltage loss alarms;
  • the occurrence of the electrical panel power loss accident is determined by the existence of the corresponding electrical panel voltage loss alarm
  • the complete loss of cold chain accident can be judged by the existence of all alarms in the equipment cooling water system or important factory water system;
  • the occurrence of the waste heat removal system failure accident is determined by the presence of a waste heat removal system equipment failure alarm.
  • the spent fuel pool water loss accident is characterized by the spent fuel pool liquid level measurement value reaching a low threshold
  • pool loss of cooling accident logic calculation unit uses the pool loss of cooling accident logic calculation unit to diagnose whether a pool loss of cooling accident occurs. If one of the following conditions occurs, it is diagnosed that the pool loss of cooling accident has occurred:
  • the spent fuel pool temperature measurement reaches a high threshold
  • the nuclear power plant unit accident condition monitoring method of the present invention also includes: when the diagnosis result is that a typical accident condition occurs in the nuclear power unit, issuing an audible and visual alarm to require the operator to call the unit accident condition automatic diagnosis screen; The operator obtains the diagnosis results of typical accident conditions of nuclear power plant units through the automatic diagnosis screen of unit accident conditions, and obtains the diagnostic logic decomposition screen of specific typical accident conditions through the corresponding navigation links on the automatic diagnosis screen of accident conditions.
  • the present invention also provides a nuclear power plant unit accident condition monitoring system, which includes:
  • the collection module is used to collect the characteristic parameters of accident conditions, important safety signals of the unit and the status of special safety facilities related to typical accident conditions;
  • the analysis and processing module is used to analyze and process the characteristic parameters of accident conditions and filter out abnormal characteristic parameters that are not within the preset threshold range;
  • the logical calculation module includes multiple parallel logical calculation units; the logical calculation units are respectively used to perform logical calculations according to preset logic on the abnormal characteristic parameters, important safety signals of the unit, and the status of special safety facilities related to the logical calculation unit. , diagnose whether typical accident conditions corresponding to the logical computing unit occur; and
  • the display module is used to display the diagnosis results of all typical accident conditions on the automatic diagnosis screen of unit accident conditions.
  • At least one logical calculation unit is set in the logical calculation module corresponding to each typical accident condition, and the logical calculation units of different typical accident conditions run in parallel. Carry out logical calculations for different typical accident conditions respectively.
  • the logical calculation module includes:
  • the steam generator heat transfer tube rupture accident logic calculation unit is used to diagnose whether the steam generator heat transfer tube rupture accident occurs
  • Containment breach accident logic calculation unit used to diagnose whether there is a containment breach accident
  • Safety factory breach accident logic calculation unit used to diagnose whether a safety factory breach accident occurs
  • the secondary circuit breach accident logic calculation unit is used to diagnose whether a secondary circuit breach accident occurs
  • the primary circuit reactive accident logic calculation unit is used to diagnose whether a primary circuit reactive accident occurs
  • Off-site power supply loss accident logic calculation unit used to diagnose whether an off-site power supply loss accident occurs
  • the plant-wide power outage accident logic calculation unit is used to diagnose whether a plant-wide power outage accident occurs;
  • the electrical panel power outage accident logic calculation unit is used to diagnose whether an electrical panel power outage accident occurs
  • Complete cold chain loss accident logic calculation unit used to diagnose whether a complete cold chain loss accident occurs
  • the waste heat discharge system failure accident logic calculation unit is used to diagnose whether a waste heat discharge system failure accident occurs
  • the logic calculation unit for the water loss accident in the water pool is used to diagnose whether there is a water loss accident in the water pool.
  • the accident condition automatic diagnosis screen is used to display to the user the unit safety signals required for automatic diagnosis of typical accident conditions, the status of special safety facilities, Diagnosis results of typical accident conditions and recommended accident operation procedures.
  • the accident condition automatic diagnosis screen includes a typical accident condition diagnosis display area, a unit status and function monitoring area, an important unit safety signal monitoring area, and important equipment And a dedicated safety facility status monitoring area, unit status diagnosis result display area and recommended operating procedure guidance area; among them, the typical accident condition diagnosis display area is used to display typical thermal and hydraulic accident diagnosis results, loss of support function accident diagnosis results, failure diagnosis results, etc.
  • the diagnosis results of fuel pool events can be displayed in a modular manner to enable the operator to clearly understand the relevant unit status information.
  • the display module is also used to guide the diagnosis logic decomposition screen of specific typical accident conditions through corresponding navigation links on the accident condition automatic diagnosis screen.
  • the nuclear power plant unit accident condition monitoring method and system of the present invention not only monitors the characteristic parameters of the accident condition in real time, but also can monitor important parameter signals of the unit and the status of special safety facilities in real time, and automatically diagnose each origin of the unit in parallel.
  • Accidents or superimposed accidents are displayed through the human-computer interaction interface to assist the operator in judging and handling the unit's accident conditions.
  • Figure 1 is a flow chart of an embodiment of the accident condition monitoring method of a nuclear power plant unit according to the present invention.
  • Figure 2 is the SGTR accident diagnosis logic diagram.
  • Figure 3 shows the logic diagram of LOCA accident diagnosis in the containment vessel.
  • Figure 4 is the logic diagram of safe factory LOCA accident diagnosis.
  • Figure 5 is a logic diagram for secondary circuit breach accident diagnosis.
  • Figure 6 is a logic diagram for primary circuit reactive accident diagnosis.
  • Figure 7 is a diagnostic logic diagram for a cold pool failure accident.
  • Figure 8 is a schematic diagram of the automatic diagnosis screen of accident conditions according to one embodiment of the present invention.
  • Figure 9 is a schematic diagram of a diagnostic logic decomposition screen according to an embodiment of the present invention.
  • Figure 10 is a flow chart of an embodiment of the accident condition monitoring system for nuclear power plant units according to the present invention.
  • the method for monitoring accident conditions of nuclear power plant units according to the present invention includes the following steps:
  • Step 101 Collect the characteristic parameters of accident conditions, important safety signals of the unit and the status of special safety facilities related to typical accident conditions.
  • accident conditions refer to conditions that deviate from normal operation more seriously than the expected operating events of the unit, including design basis accidents and design extended conditions; the monitoring scope of typical accident conditions in the present invention includes design basis accidents and design Extended operating conditions, but does not cover the design extended operating conditions for core damage.
  • the pressurized water reactor nuclear power plant defines 6 state functional parameters (subcriticality, primary circuit water content, waste heat export, steam generator water content, steam generator integrity, containment integrity). These states The combination of functional parameters can characterize the physical state of the reactor. Therefore, the accident conditions and design expansion conditions described in the accident conditions list of the nuclear power plant can be classified according to the different degradation conditions of the six major status functional parameters of the unit caused by the accident conditions.
  • this invention classifies typical accident conditions as follows:
  • Loss of support function accidents including LOOP (loss of off-site power supply) accidents, SBO (plant-wide power outage) accidents, electrical panel power outage accidents, TLOCC (total loss of cold chain) accidents and RIS-RHR (residual heat removal system) Fault;
  • Typical accident conditions of nuclear power units lead to different deterioration of the six major state parameters related to the three major safety functions of the unit.
  • the characteristic parameters that characterize the typical accident conditions can be obtained, such as: A large break accident in the circuit will cause the liquid level of the pressure vessel to drop rapidly and the pressure and radioactivity in the containment to rise.
  • these characteristic parameters need to be continuously measured, such as the characteristic parameters required for a typical thermal hydraulic accident in a pressurized water reactor nuclear power plant: intermediate range neutron flux, primary loop subcooling, pressure vessel liquid level, steam generator Wide range of liquid level, steam generator secondary side steam pipeline radioactivity measurement, steam generator sewage sampling radioactivity measurement, condenser exhaust radioactivity measurement, containment pressure, radioactivity within the containment, etc.
  • the characteristic parameters of the accident conditions can be automatically monitored and obtained through functional parameter measuring instruments of the nuclear power plant.
  • the reactor nuclear power can be measured by the power range detector, intermediate range detector and source range detector of the nuclear instrument system.
  • the core neutron flux, core temperature, and pressure vessel water level can be measured by the self-sufficient neutron detector and thermocouple detector of the core measurement system respectively
  • the primary circuit pressure can be cooled by the reactor arranged in the pressurizer
  • the primary loop temperature can be measured by the narrow range temperature measuring instruments and safety injection system of the reactor coolant system arranged in the cold pipe section and hot pipe section of the primary loop. Measured by the system's wide range temperature measuring instrument.
  • the important safety signals of the unit are obtained by using the protection signals of the normal operation of the nuclear power unit or by designing corresponding logic using the unit status parameters.
  • Special safety facilities refer to safety facilities specially set up by nuclear power plants to ensure the discharge of core heat and the integrity of the containment, limit the development of accidents and reduce the consequences of accidents.
  • the status of special safety facilities can be determined by the special safety facilities for unit accident conditions. NSSS function availability information is monitored.
  • Step 102 Analyze and process the characteristic parameters of the accident conditions and screen out abnormal characteristic parameters that are not within the preset threshold range.
  • analyzing and processing the characteristic parameters of the accident conditions and filtering out the abnormal characteristic parameters that are not within the preset threshold range include: comparing the characteristic parameters of the accident conditions with the preset thresholds, and judging the parameters that are not within the preset threshold range as Abnormal characteristic parameters.
  • the characteristic parameters of the accident conditions are pre-processed, including one or both of the following processes: (1) Preprocessing of the accident conditions according to the conservative value principle To filter the working condition characteristic parameters, the collected multiple columns of redundant data are conservatively selected, such as larger or smaller values, so as to obtain data for preliminary screening to improve the accuracy and effectiveness of data analysis; ( 2) Screen the characteristic parameters of accident conditions for invalidity and delete invalid data to eliminate data that affects the diagnosis results, thereby improving the accuracy and effectiveness of the data.
  • Step 103 Use multiple logical computing units to conduct parallel diagnosis of various typical accident conditions.
  • Each logical computing unit analyzes abnormal characteristic parameters, important safety signals of the unit, and special safety facilities related to the logical computing unit according to preset logic. Perform logical calculations on the status to diagnose whether typical accident conditions corresponding to the logical calculation unit occur.
  • At least one logical calculation unit is set up corresponding to each typical accident condition.
  • the logical calculation units of different typical accident conditions run in parallel and perform logical calculations on different typical accident conditions respectively, thereby diagnosing each typical accident in the shortest time. Whether the accident occurred.
  • the characterization of typical accident conditions of the unit needs to combine the characteristic parameters of the accident conditions with the important signals of the unit and the status of the special safety facilities.
  • the invention adopts parallel monitoring and logical calculation methods, so that the initial accident conditions and the superimposed accident conditions can be monitored simultaneously without interference between the relevant logics.
  • Thermal hydraulic accidents refer to accident conditions that cause direct changes in unit status parameters.
  • the measured radioactivity value of the steam pipeline on the secondary side of the steam generator is greater than the high threshold value (for example, the high threshold value of the pressurized water reactor nuclear power plant is 4.0E+6Bq/m 3 ) or the measured radioactivity value of the steam generator wastewater sampling is greater than the high high threshold value.
  • the threshold value (for example, the high threshold value of the pressurized water reactor nuclear power plant is 2.0E+8Bq/m 3 ) can determine that there is a large radioactive leakage from the primary circuit to the secondary circuit. According to the radioactive source of the pressurized water reactor nuclear power plant, it can be determined that it is due to SGTR The accident caused the heat transfer tube of the steam generator to rupture and radioactive material from the primary circuit leaked into the secondary circuit.
  • the measured radioactivity value of the steam pipeline on the secondary side of the steam generator is greater than the high threshold value (for example, the high threshold value of the pressurized water reactor nuclear power plant is 4.0E+4Bq/m 3 ) or the measured radioactivity value of the steam generator wastewater sampling is greater than the high threshold value (
  • the high threshold value of the pressurized water reactor nuclear power plant is 2.0E+6Bq/m 3
  • the liquid level of the superimposed primary loop voltage regulator or capacity control box is low (if the pressurized water reactor nuclear power plant selects the lowest liquid level of the corresponding equipment, when the signal is triggered Indicates that there is a large amount of water supply in the primary circuit) indicating that an SGTR accident has occurred.
  • the secondary circuit is slightly radioactive, the primary circuit water level fluctuation signal is used to verify whether an SGTR accident has actually occurred.
  • the primary circuit LOCA accident is divided into LOCA in the containment and LOCA in the safety plant according to the location where the LOCA occurs. Two accident logic calculation units need to be used to diagnose the two LOCA accidents respectively.
  • the dose rate measurement value in the containment reaches the high threshold (for example, the high threshold of this pressurized water reactor nuclear power plant is 1.0E+1Gy/h), indicating that the pressure boundary of the primary circuit is damaged and the radioactive material breaks through the second safety barrier.
  • the high threshold for example, the high threshold of this pressurized water reactor nuclear power plant is 1.0E+1Gy/h
  • the dose rate measurement value in the containment reaches the high threshold (for example, the high threshold value of the pressurized water reactor nuclear power plant is 1.0E-2Gy/h), it must be verified by the containment pressure measurement. If the pressure measurement in the containment is superimposed at the same time, If the value reaches the high threshold (for example, the high threshold of this pressurized water reactor nuclear power plant is 1.23 bar.a), it indicates that a LOCA accident in the containment has occurred.
  • the high threshold for example, the high threshold value of the pressurized water reactor nuclear power plant is 1.0E-2Gy/h
  • the logic for diagnosing that a LOCA accident in a safe plant has occurred is: Confirm that the RIS-RHR connection condition has taken effect through the P25 signal (residual heat exhaust system allowed connection signal). After the connection, it is judged that the measured pressure value of the safe plant is greater than the high threshold (such as The high threshold of this pressurized water reactor nuclear power plant is 0.12Mpa.a, indicating that mass and energy release has occurred in the safety plant) or the liquid level measurement value in the pit of the safety plant is greater than the high threshold (for example, the high threshold of this pressurized water reactor nuclear power plant indicates that the pit is about to overflow. Flow level value 0.15m), after RHR connection, radioactive leakage in the primary circuit or large amount of leaked coolant collection indicates the occurrence of LOCA accident in the safe factory.
  • the measured value of the VVP (main steam isolation valve) valve compartment reaches a high threshold (for example, the high threshold of this pressurized water reactor nuclear power plant is 75°C, through a temperature design that is much higher than the ambient temperature and keeps the measuring instrument responsive enough, indicating that Main steam leakage occurs), used to diagnose the main steam pipeline breach accident outside the containment vessel.
  • a high threshold for example, the high threshold of this pressurized water reactor nuclear power plant is 75°C, through a temperature design that is much higher than the ambient temperature and keeps the measuring instrument responsive enough, indicating that Main steam leakage occurs
  • the measurement value of the ARE (main water supply system) pipeline compartment reaches a high threshold (for example, the high threshold of the pressurized water reactor nuclear power plant is 75°C, through a temperature design that is much higher than the ambient temperature and maintains a sufficiently sensitive response of the measuring instrument, indicating that the occurrence of Main water supply pipeline leakage), used to diagnose the main water supply pipeline breach accident outside the containment vessel.
  • a high threshold for example, the high threshold of the pressurized water reactor nuclear power plant is 75°C, through a temperature design that is much higher than the ambient temperature and maintains a sufficiently sensitive response of the measuring instrument, indicating that the occurrence of Main water supply pipeline leakage
  • the primary circuit reactive accident logic calculation unit is used to diagnose whether a primary circuit reactive accident occurs. Please refer to Figure 6. If one of the following situations occurs, it is diagnosed that a primary circuit reactive accident has occurred:
  • the measured value of the total boron concentration in the primary loop is lower than the low threshold (the low pressure threshold of the pressurized water reactor nuclear power plant and the following thresholds in this paragraph are the thresholds required according to the unit technical specifications).
  • Accidents caused by loss of support functions mainly include accident conditions such as loss of power and loss of cooling of the unit.
  • the occurrence of this type of accident can be an initiating accident or a superimposed accident.
  • (1) LOOP accident The off-site power supply loss accident is characterized by the presence of the corresponding factory AC bus voltage loss alarm to determine the occurrence of the accident.
  • the off-site power supply loss accident logic calculation unit is used to diagnose whether an off-site power supply loss accident occurs.
  • TLOCC accident The complete loss of cold chain accident is characterized by the complete loss of alarms in the RRI/SEC system (equipment cooling water system or important factory water system) to determine the occurrence of the accident. Use the complete cold chain loss accident logic calculation unit to diagnose whether a complete cold chain loss accident occurs.
  • RIS-RHR failure accident characterization can be judged by the presence of RIS-RHR equipment failure alarm. Use the RIS-RHR fault accident logic calculation unit to diagnose whether a RIS-RHR fault accident occurs.
  • Spent fuel pool accidents are divided into spent fuel pool loss of cooling accidents and spent fuel pool water loss accidents.
  • Spent pool water loss accident A spent pool water loss accident occurs when the measured value of the spent fuel pool liquid level reaches a low threshold (the low pressure threshold of this pressurized water reactor nuclear power plant is 16.3m. This threshold is to prevent the spent fuel pool liquid level from falling below The alarm liquid level of the cooling column suction port) is characterized. Use the water loss accident logic calculation unit of the water pool to diagnose whether there is a water loss accident in the water pool.
  • Step 104 Display the diagnosis results of all typical accident conditions on the automatic diagnosis screen of unit accident conditions.
  • an audible and visual alarm is issued to require the operator to call the automatic diagnosis screen for the unit accident condition.
  • the operator obtains the diagnosis results of typical accident conditions of nuclear power plant units through the automatic diagnosis screen of unit accident conditions.
  • the sound and light alarm can be issued through the sound and light alarm system in the main control room.
  • FIG. 8 A schematic diagram of a specific embodiment of the automatic diagnosis screen for accident conditions is shown in Figure 8, which is used to display to the user the unit safety signals required for automatic diagnosis of typical accident conditions, the status of special safety facilities, the diagnosis results of typical accident conditions, and Recommended incident operating procedures.
  • the accident condition automatic diagnosis screen includes a typical accident condition diagnosis display area.
  • the typical accident condition diagnosis display area is used to display typical thermal hydraulic accident diagnosis results, loss of support function accident diagnosis results, and spent fuel pool event diagnosis results. The operator can clearly understand the relevant unit status information through modular display of diagnosis results of typical accident conditions.
  • the typical accident condition diagnosis display area is configured with a display indicator light corresponding to the typical accident condition. When the indicator light is on, it indicates that the accident has been triggered.
  • the accident condition automatic diagnosis screen also includes a unit status function monitoring area, an important safety signal monitoring area for the unit, a status monitoring area for important equipment and special safety facilities, a unit status diagnosis result display area, and a recommended operating procedures guidance area.
  • the automatic diagnosis screen for accident conditions not only provides diagnosis results for typical accident conditions of the unit, but also integrates important safety information and status information of special safety facilities on the diagnosis page. The automatic diagnosis results can be verified through relevant information.
  • the automatic diagnosis screen of accident conditions can also be directed to the diagnostic logic decomposition screen of specific typical accident conditions through corresponding navigation links.
  • the operator can obtain specific typical accident conditions through the corresponding navigation links on the automatic diagnosis screen of accident conditions.
  • the diagnostic logic decomposition screen of the situation is shown in Figure 9.
  • the corresponding diagnostic logic decomposition screen is shown in Figure 9.
  • the nuclear power plant unit accident condition monitoring method of the present invention diagnoses various typical accident conditions in parallel without affecting each other, and can diagnose the initiating accident and the superimposed accident at the same time; moreover, the unit safety signal, the dedicated safety signal
  • the facility status can monitor the status of the unit equipment in real time and trigger relevant signals or change the screen icon display when the corresponding status changes.
  • the accident condition monitoring system for nuclear power plant units of the present invention includes a collection module 10 , an analysis and processing module 20 , a logic calculation module 30 and a display module 40 .
  • the acquisition module 10 is used to collect accident condition characteristic parameters, important safety signals of the unit and the status of special safety facilities related to typical accident conditions.
  • this invention classifies typical accident conditions as follows:
  • the characteristic parameters of the accident conditions can be automatically monitored and obtained through functional parameter measuring instruments of the nuclear power plant.
  • the reactor nuclear power can be measured by the power range detector, intermediate range detector and source range detector of the nuclear instrument system.
  • the core neutron flux, core temperature, and pressure vessel water level can be measured by the self-sufficient neutron detector and thermocouple detector of the core measurement system respectively
  • the primary circuit pressure can be cooled by the reactor arranged in the pressurizer
  • the primary loop temperature can be measured by the narrow range temperature measuring instruments and safety injection system of the reactor coolant system arranged in the cold pipe section and hot pipe section of the primary loop. Measured by the system's wide range temperature measuring instrument.
  • the important safety signals of the unit are obtained by using the protection signals of the normal operation of the nuclear power unit or by designing corresponding logic using the unit status parameters.
  • Special safety facilities refer to safety facilities specially set up by nuclear power plants to ensure the discharge of core heat and the integrity of the containment, limit the development of accidents and reduce the consequences of accidents.
  • the status of special safety facilities can be determined by the special safety facilities for unit accident conditions. NSSS function availability information is monitored.
  • the analysis and processing module 20 is used to analyze and process the characteristic parameters of accident conditions and screen out abnormal characteristic parameters that are not within the preset threshold range.
  • the analysis and processing module 20 includes a threshold comparison unit, which is used to compare the characteristic parameters of accident conditions with preset thresholds, and determine parameters that are not within the preset threshold range as abnormal characteristic parameters.
  • the analysis and processing module 20 also includes a preprocessing unit.
  • the preprocessing unit is used to preprocess the accident condition characteristic parameters before comparing the accident condition characteristic parameters with the preset threshold, including one of the following processes: One or two: (1) Screen the characteristic parameters of the accident working conditions according to the conservative value principle, that is, conservatively select the values of the collected multiple columns of redundant data, such as taking a larger or smaller value, so as to obtain preliminary screening data to improve the accuracy and effectiveness of data analysis; (2) filter the invalidity of the characteristic parameters of accident conditions and delete invalid data to eliminate data that affects the diagnosis results, thereby improving the accuracy and effectiveness of the data sex.
  • the logical calculation module 30 includes multiple parallel logical calculation units; the multiple logical calculation units are used to calculate the abnormal characteristic parameters, important safety signals of the unit, and the status of special safety facilities related to the logical calculation unit according to preset logic. Perform logical calculations to diagnose whether typical accident conditions corresponding to the logical calculation unit occur.
  • the logical calculation module 30 is provided with at least one logical calculation unit corresponding to each typical accident condition.
  • the logical calculation units of different typical accident conditions run in parallel to perform logical calculations on different typical accident conditions respectively, so as to achieve the goal in the shortest time. Diagnose the occurrence of each typical accident situation.
  • the logical calculation module 30 includes:
  • SGTR accident logic calculation unit used to diagnose whether an SGTR accident occurs
  • the LOCA accident logic calculation unit within the containment vessel is used to diagnose whether a LOCA accident within the containment vessel occurs;
  • the safe factory LOCA accident logic calculation unit is used to diagnose whether a safe factory LOCA accident occurs;
  • the secondary circuit breach accident logic calculation unit is used to diagnose whether a secondary circuit breach accident occurs
  • the primary circuit reactive accident logic calculation unit is used to diagnose whether a primary circuit reactive accident occurs
  • LOOP accident logic calculation unit used to diagnose whether a LOOP accident occurs
  • the SBO accident logic calculation unit is used to diagnose whether an SBO accident occurs
  • the electrical panel power outage accident logic calculation unit is used to diagnose whether an electrical panel power outage accident occurs
  • TLOCC accident logic calculation unit used to diagnose whether a TLOCC accident occurs
  • RIS-RHR fault accident logic calculation unit used to diagnose whether a RIS-RHR fault accident occurs
  • the logic calculation unit for the water loss accident in the water pool is used to diagnose whether there is a water loss accident in the water pool.
  • each accident logic calculation unit in the logic calculation module 30 is as described in the foregoing method embodiments, and will not be described again here.
  • the display module 40 is used to display the diagnosis results of all typical accident conditions on the automatic diagnosis screen of unit accident conditions.
  • the display module 40 is connected to an alarm module.
  • the alarm module is used to issue an audible and visual alarm to request the operator when the diagnosis result is that a typical accident condition occurs in a nuclear power unit (that is, any one or more logical computing units diagnose a typical accident condition). Call up the automatic diagnosis screen for unit accident conditions.
  • the alarm module can be a sound and light alarm system in the main control room.
  • FIG. 8 A schematic diagram of a specific embodiment of the automatic diagnosis screen for accident conditions is shown in Figure 8, which is used to display to the user the unit safety signals required for automatic diagnosis of typical accident conditions, the status of special safety facilities, the diagnosis results of typical accident conditions, and Recommended incident operating procedures.
  • the accident condition automatic diagnosis screen includes a typical accident condition diagnosis display area.
  • the typical accident condition diagnosis display area is used to display typical thermal hydraulic accident diagnosis results, loss of support function accident diagnosis results, and spent fuel pool event diagnosis results. The operator can clearly understand the relevant unit status information through modular display of diagnosis results of typical accident conditions.
  • the typical accident condition diagnosis display area is configured with a display indicator light corresponding to the typical accident condition. When the indicator light is on, it indicates that the accident has been triggered.
  • the accident condition automatic diagnosis screen also includes a unit status function monitoring area, an important safety signal monitoring area for the unit, a status monitoring area for important equipment and special safety facilities, a unit status diagnosis result display area, and a recommended operating procedures guidance area.
  • the automatic diagnosis screen for accident conditions not only provides diagnosis results for typical accident conditions of the unit, but also integrates important safety information and status information of special safety facilities on the diagnosis page. The automatic diagnosis results can be verified through relevant information.
  • the display module is also used to guide the diagnosis logic decomposition screen of specific typical accident conditions through corresponding navigation links on the automatic diagnosis screen of accident conditions.
  • the corresponding diagnosis logic decomposition screen is shown in Figure 9.
  • the nuclear power plant unit accident condition monitoring method and system of the present invention not only monitor the characteristic parameters of the accident condition in real time, but also can monitor the important parameter signals of the unit and the status of the special safety facilities in real time, and perform automatic parallel diagnosis.
  • Each initial accident or superimposed accident of the unit is displayed through the human-computer interaction interface to assist the operator in judging and handling the accident conditions of the unit.
  • the method and system for monitoring accident conditions of nuclear power plant units of the present invention have at least the following beneficial technical effects:
  • the present invention can realize automatic diagnosis of accident-initiating accidents or superimposed accidents in accident situations, and can help the operator quickly understand the current unit status and evaluate the unit status. It predicts changes, improves the operator's accident response speed, effectively reduces the human failure caused by the operator's control of the unit accident, and can also provide a decision-making basis for the power plant's emergency response organization.
  • the integrated display of important unit signals and the status of special safety facilities used under typical accident conditions of the unit allows the operator to clearly understand the true and reliable system and functional configuration required by the unit, and can clearly Improve the operator's response efficiency to unit accident conditions.
  • RAM random access memory
  • ROM read-only memory
  • electrically programmable ROM electrically erasable programmable ROM
  • registers hard disks, removable disks, CD-ROMs, or anywhere in the field of technology. any other known form of storage media.

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Abstract

一种核电厂机组事故工况监测方法和系统，方法包括：采集与典型事故工况相关的事故工况特征参数、机组重要安全信号和专设安全设施状态（101）；对事故工况特征参数进行分析处理，筛选出不在预设阈值范围内的异常特征参数（102）；利用多个逻辑计算单元对各种典型事故工况进行并行诊断（103）；在机组事故工况自动诊断画面上显示出全部典型事故工况的诊断结果（104）。通过实时监测事故工况特征参数、机组重要安全信号及专设安全设施状态，自动并行诊断机组各个始发事故或叠加事故，并通过人机交互界面予以显示，辅助操纵员对机组事故工况进行判断和处理。

Description

核电厂机组事故工况监测方法和系统 技术领域
本发明属于核电厂机组事故监测领域，更具体地说，本发明涉及一种核电厂机组事故工况监测方法和系统。
背景技术
当核电厂发生故障或事故后，为保证核电厂的安全，保护三道屏障的完整性，限制放射性物质向外释放，需要将机组控制到可控或安全停堆状态。事故运行规程用于故障或事故后指导操纵员控制机组，通过最大限度地利用可用的系统功能，减少异常或限制事故的后果，将机组恢复到一种安全的状态。
目前，一种已知的核电技术采用状态导向或征兆导向事故程序，操纵员主要依据有限几个表征反应堆三大安全功能状态的机组状态参数恶化程度执行事故程序，缺乏对事故起因的了解，不利于对事故的整体把握，影响事故处理效率，且容易迷失。
有鉴于此，确有必要提供一种能够解决上述问题的核电厂机组事故工况监测方法和系统。
发明内容
本发明的目的在于：提供一种能够自动诊断机组的始发事故和叠加事故，且能够显示事故起因的核电厂机组事故工况监测方法和系统。
为了实现上述发明目的，本发明提供了一种核电厂机组事故工况监测方法，所述方法包括以下步骤：
采集与典型事故工况相关的事故工况特征参数、机组重要安全信号和专设安全设施状态；
对事故工况特征参数进行分析处理，筛选出不在预设阈值范围内的异常特征参数；
利用多个逻辑计算单元对各种典型事故工况进行并行诊断，每一逻辑计算单元按预设逻辑对与该逻辑计算单元相关的异常特征参数、机组重要安全信号、专设安全设施状态进行逻辑计算，诊断是否发生与该逻辑计算单元对应的典型事故工况；
在机组事故工况自动诊断画面上显示出全部典型事故工况的诊断结果。
作为本发明核电厂机组事故工况监测方法的一种优选实施方式，所述典型事故工况包括典型热工水力事故、失去支持功能事故和乏燃料水池事件；其中，典型热工水力事故包括蒸汽发生器传热管破裂事故、一回路破口事故、一回路反应性事故和二回路破口事故；失去支持功能事故包括厂外电源失去事故、全厂失电事故、电气盘失电事故、完全丧失冷链事故和余热排出系统故障；乏燃料水池事件包括乏池失冷事故和乏池失水事故；
对应每一典型事故工况设置至少一个逻辑计算单元，不同典型事故工况的逻辑计算单元并行运行，分别对不同典型事故工况进行逻辑计算。
作为本发明核电厂机组事故工况监测方法的一种优选实施方式，利用蒸汽发生器传热管破裂事故逻辑计算单元对是否出现蒸汽发生器传热管破裂事故进行诊断，出现以下情况之一即诊断蒸汽发生器传热管破裂事故已经发生：
一、二回路泄漏率平衡异常信号存在；
蒸汽发生器二次侧蒸汽管道放射性测量值大于高高阈值或蒸汽发生器排污水取样放射性测量值大于高高阈值；
蒸汽发生器二次侧蒸汽管道放射性测量值大于高阈值或蒸汽发生器排污水取样放射性测量值大于高阈值，叠加一回路稳压器或容控箱液位低。
作为本发明核电厂机组事故工况监测方法的一种优选实施方式，所述一回路破口事故根据破口发生的位置不同区分为安全壳内破口和安全厂房破口，利 用两个事故逻辑计算单元对两种破口事故分别进行诊断；
利用安全壳内破口事故逻辑计算单元对是否出现安全壳内破口事故进行诊断，出现以下情况之一即诊断安全壳内破口事故已经发生：(a)安全壳内剂量率测量值达到高高阈值；(b)安全壳内剂量率测量值达到高阈值，叠加安全壳内压力测量值达到高阈值；(c)一回路封闭状态下，安全壳内压力测量值达到高阈值，且存在一回路过冷度低或压力容器液位低信号；
利用安全厂房破口事故逻辑计算单元对是否出现安全厂房破口事故进行诊断，诊断安全厂房破口事故已经发生的逻辑为：余热排出系统允许连接信号已生效，且安全厂房压力测量值大于高阈值或安全厂房地坑液位测量值大于高阈值。
作为本发明核电厂机组事故工况监测方法的一种优选实施方式，利用二回路破口事故逻辑计算单元对是否出现二回路破口事故进行诊断，出现以下情况之一即诊断二回路破口事故已经发生：
主蒸汽隔离阀阀门隔间测量值达到高阈值；
主给水系统管道隔间测量值达到高阈值；
一列主蒸汽释放系统隔离阀自动隔离信号存在，且三列主蒸汽释放系统隔离阀自动隔离信号不存在；
两台蒸汽发生器内压差高，且余热排出系统允许连接信号未生效，叠加三列主蒸汽释放系统隔离阀自动隔离信号存在。
作为本发明核电厂机组事故工况监测方法的一种优选实施方式，利用一回路反应性事故逻辑计算单元对是否出现一回路反应性事故进行诊断，出现以下情况之一即诊断一回路反应性事故已经发生：
一回路总硼浓度测量值低于低阈值；
超功率或超温△T跳堆信号；
功率量程高/低定值中子注量率高；
中间量程中子注量率高；
源间量程中子注量率高；
功率运行R棒棒位低；
停堆中子注量率高。
作为本发明核电厂机组事故工况监测方法的一种优选实施方式，所述失去支持功能事故的诊断逻辑为：
通过相应厂用交流电母线电压失去报警存在来判断所述厂外电源失去事故发生；
通过相应应急母线和非应急母线电压失去报警存在来判断所述全厂失电事故发生；
通过相应电气盘电压失去报警存在来判断所述电气盘失电事故发生；
通过设备冷却水系统或重要厂用水系统全部失去报警存在来判断所述完全丧失冷链事故发生；
通过余热排出系统设备故障报警存在来判断所述余热排出系统故障事故发生。
作为本发明核电厂机组事故工况监测方法的一种优选实施方式，所述乏池失水事故通过乏燃料水池液位测量值达到低阈值来表征；
利用乏池失冷事故逻辑计算单元对是否出现乏池失冷事故进行诊断，出现以下情况之一即诊断诊断乏池失冷事故已经发生：
失去全部乏燃料水池冷却列；
乏燃料水池温度测量值达到高高阈值；
失去部分乏燃料水池冷却列且乏燃料水池温度测量值达到高阈值。
作为本发明核电厂机组事故工况监测方法的一种优选实施方式，还包括：当诊断结果为核电机组发生典型事故工况时，发出声光报警要求操纵员调用机组事故工况自动诊断画面；操纵员通过机组事故工况自动诊断画面获取核电厂 机组典型事故工况诊断结果，通过事故工况自动诊断画面上相应的导航链接获取具体典型事故工况的诊断逻辑分解画面。
为了实现上述发明目的，本发明还提供了一种核电厂机组事故工况监测系统，其包括：
采集模块，用于采集与典型事故工况相关的事故工况特征参数、机组重要安全信号和专设安全设施状态；
分析处理模块，用于对事故工况特征参数进行分析处理，筛选出不在预设阈值范围内的异常特征参数；
逻辑计算模块，包括多个并行的逻辑计算单元；所述逻辑计算单元分别用于按预设逻辑对与该逻辑计算单元相关的异常特征参数、机组重要安全信号、专设安全设施状态进行逻辑计算，诊断是否发生与该逻辑计算单元对应的典型事故工况；和
显示模块，用于在机组事故工况自动诊断画面上显示出全部典型事故工况的诊断结果。
作为本发明核电厂机组事故工况监测系统的一种优选实施方式，所述逻辑计算模块中对应每一典型事故工况设置至少一个逻辑计算单元，不同典型事故工况的逻辑计算单元并行运行，分别对不同典型事故工况进行逻辑计算。
作为本发明核电厂机组事故工况监测系统的一种优选实施方式，所述逻辑计算模块包括：
蒸汽发生器传热管破裂事故逻辑计算单元，用于对是否出现蒸汽发生器传热管破裂事故进行诊断；
安全壳内破口事故逻辑计算单元，用于对是否出现安全壳内破口事故进行诊断；
安全厂房破口事故逻辑计算单元，用于对是否出现安全厂房破口事故进行诊断；
二回路破口事故逻辑计算单元，用于对是否出现二回路破口事故进行诊断；
一回路反应性事故逻辑计算单元，用于对是否出现一回路反应性事故进行诊断；
厂外电源失去事故逻辑计算单元，用于对是否出现厂外电源失去事故进行诊断；
全厂失电事故逻辑计算单元，用于对是否出现全厂失电事故进行诊断；
电气盘失电事故逻辑计算单元，用于对是否出现电气盘失电事故进行诊断；
完全丧失冷链事故逻辑计算单元，用于对是否出现完全丧失冷链事故进行诊断；
余热排出系统故障事故逻辑计算单元，用于对是否出现余热排出系统故障事故进行诊断；
乏池失冷事故逻辑计算单元，用于对是否出现乏池失冷事故进行诊断；和
乏池失水事故逻辑计算单元，用于对是否出现乏池失水事故进行诊断。
作为本发明核电厂机组事故工况监测系统的一种优选实施方式，所述事故工况自动诊断画面用于向用户显示典型事故工况自动诊断所需的机组安全信号、专设安全设施状态、典型事故工况诊断结果以及建议执行的事故运行规程。
作为本发明核电厂机组事故工况监测系统的一种优选实施方式，所述事故工况自动诊断画面包括典型事故工况诊断显示区域、机组状态功能监测区域、机组重要安全信号监测区域、重要设备及专设安全设施状态监测区域、机组状态诊断结果显示区域及建议操作规程指引区域；其中，典型事故工况诊断显示区域用于显示典型热工水力事故诊断结果、失去支持功能事故诊断结果、乏燃料水池事件诊断结果，通过模块化展示典型事故工况诊断结果来使操纵员清晰了解相关机组状态信息。
作为本发明核电厂机组事故工况监测系统的一种优选实施方式，所述显示模块还用于在事故工况自动诊断画面上通过相应的导航链接导向具体典型事故 工况的诊断逻辑分解画面。
与现有技术相比，本发明核电厂机组事故工况监测方法和系统不仅实时监测事故工况特征参数，还能够实时监测机组重要参数信号及专设安全设施状态，自动并行诊断机组各个始发事故或叠加事故，并通过人机交互界面予以显示，辅助操纵员对机组事故工况进行判断和处理。
附图说明
下面结合附图和具体实施方式，对本发明核电厂机组事故工况监测方法和系统进行详细说明。
图1为本发明核电厂机组事故工况监测方法的一个实施例流程图。
图2为SGTR事故诊断逻辑图。
图3为安全壳内LOCA事故诊断逻辑图。
图4为安全厂房LOCA事故诊断逻辑图。
图5为二回路破口事故诊断逻辑图。
图6为一回路反应性事故诊断逻辑图。
图7为乏池失冷事故诊断逻辑图。
图8为本发明一个实施例的事故工况自动诊断画面示意图。
图9为本发明一个实施例的诊断逻辑分解画面示意图。
图10为本发明核电厂机组事故工况监测系统的一个实施例流程图。
具体实施方式
为了使本发明的目的、技术方案及其有益技术效果更加清晰，以下结合附图和具体实施方式，对本发明进行进一步详细说明。应当理解的是，本说明书中描述的具体实施例仅仅是为了解释本发明，并非为了限定本发明。
请参阅图1，本发明核电厂机组事故工况监测方法包括以下步骤：
步骤101，采集与典型事故工况相关的事故工况特征参数、机组重要安全信 号和专设安全设施状态。
在核电领域，事故工况是指比机组预计运行事件更为严重的偏离正常运行的工况，包括设计基准事故，设计扩展工况；本发明典型事故工况的监测范围包括设计基准事故及设计扩展工况，但不覆盖堆芯损坏的设计扩展工况。
具体的，压水堆核电厂定义了6个状态功能参数(次临界度、一回路水装量、余热导出、蒸汽发生器水装量、蒸汽发生器完整性、安全壳完整性)，这些状态功能参数的组合可以表征反应堆所处的物理状态。因此，可以根据事故工况引起机组6大状态功能参数降级情况不同，来对核电厂事故工况清单所述的事故工况和设计拓展工况进行分类。
以某压水堆核电厂为例，本发明对典型事故工况的分类如下：
(一)典型热工水力事故，包括SGTR(蒸汽发生器传热管破裂)事故、一回路LOCA(一回路破口)事故、一回路反应性事故和二回路破口事故；
(二)失去支持功能事故，包括LOOP(厂外电源失去)事故、SBO(全厂失电)事故、电气盘失电事故、TLOCC(完全丧失冷链)事故和RIS-RHR(余热排出系统)故障；
(三)乏燃料水池事件，包括乏池失冷事故和乏池失水事故。
核电机组典型事故工况导致的表征机组3大安全功能相关的六大状态参数恶化情况不同，根据典型事故工况事故演化过程及安全分析结果，可以得到表征该典型事故工况的特征参数，如一回路大破口事故，会导致压力容器液位快速降低，安全壳内压力和放射性升高。在机组运行期间这些特征参数需要连续测量，如某压水堆核电厂典型热工水力事故所需的特征参数：中间量程中子通量、一回路过冷度、压力容器液位、蒸汽发生器宽量程液位、蒸汽发生器二次侧蒸汽管道放射性测量、蒸汽发生器排污水取样放射性测量、冷凝器抽气放射性测量、安全壳压力、安全壳内放射性等。
具体的，所述事故工况特征参数可以通过核电厂功能参数测量仪表自动监 测和获取，例如，反应堆核功率可以由核仪表系统的功率量程探测器、中间量程探测器和源量程探测器测得；堆芯中子通量、堆芯温度、压力容器水位分别可以由堆芯测量系统的自给能中子探测器和热电偶探测器测得；一回路压力可以由布置在稳压器的反应堆冷却剂系统压力测量仪表和布置在一回路热管段的安全注入系统压力测量仪表测得；一回路温度可以由布置在一回路冷管段和热管段的反应堆冷却剂系统的窄量程温度测量仪表和安全注入系统的宽量程温度测量仪表测得。所述机组重要安全信号，通过使用核电机组正常运行的保护信号或通过使用机组状态参数设计相应逻辑获取。专设安全设施是指核电厂为确保堆芯热量的排出和安全壳的完整性，限制事故的发展和减轻事故后果而专门设置的安全设施，专设安全设施状态可以通过机组事故工况专用的NSSS功能可用性信息监测得到。
步骤102，对事故工况特征参数进行分析处理，筛选出不在预设阈值范围内的异常特征参数。
具体的，对事故工况特征参数进行分析处理，筛选出不在预设阈值范围内的异常特征参数包括：将事故工况特征参数与预设阈值进行比较，不在预设阈值范围内的参数判断为异常特征参数。
优选地，在将事故工况特征参数与预设阈值进行比较之前，先对事故工况特征参数进行预处理，包括以下处理中的一种或两种：(1)根据保守取值原则对事故工况特征参数进行筛选，即将所采集到的多列冗余的数据进行保守取值，如取大或者取小，从而获得进行初步筛选的数据，以提高数据分析的准确度和有效性；(2)对事故工况特征参数进行无效性筛选，删除无效数据，以剔除对诊断结果产生影响的数据，从而提高数据的准确性和有效性。
步骤103，利用多个逻辑计算单元对各种典型事故工况进行并行诊断，每一逻辑计算单元按预设逻辑对与该逻辑计算单元相关的异常特征参数、机组重要安全信号、专设安全设施状态进行逻辑计算，诊断是否发生与该逻辑计算单元 对应的典型事故工况。
具体的，对应每一典型事故工况设置至少一个逻辑计算单元，不同典型事故工况的逻辑计算单元并行运行，分别对不同典型事故工况进行逻辑计算，从而在最短时间内诊断出每一典型事故工况是否发生。
机组典型事故工况表征需要结合事故工况特征参数及机组重要信号和专设安全设施状态。本发明采用并联监测和逻辑计算方法，使得始发事故工况和叠加事故工况可以同时监测，相关逻辑之间相不干扰。
以前述压水堆核电厂为例，对各种典型事故工况的表征逻辑进行说明：
(一)典型热工水力事故
热工水力事故是指引起机组状态参数直接变化的事故工况。
(1)SGTR事故
利用SGTR事故逻辑计算单元对是否出现SGTR事故进行诊断。请参阅图2，出现以下情况之一即诊断SGTR事故已经发生：
(a)一、二回路泄漏率平衡异常信号存在。该信号存在表征一回路向二回路存在较大泄漏(如该压水堆核电厂阈值为一回路向二回路泄漏阈值为大于70L/H)，根据压水堆核电机组系统功能联系，可以判断为由于SGTR事故导致该信号触发。
(b)蒸汽发生器二次侧蒸汽管道放射性测量值大于高高阈值(如该压水堆核电厂高阈值为4.0E+6Bq/m 3)或蒸汽发生器排污水取样放射性测量值大于高高阈值(如该压水堆核电厂高高阈值为2.0E+8Bq/m 3)，能够判定出一回路向二回路存在较大放射性泄漏，根据压水堆核电机组放射性来源，可以判断为由于SGTR事故导致蒸汽发生器传热管破裂，一回路放射性物质向二回路泄漏。
(c)蒸汽发生器二次侧蒸汽管道放射性测量值大于高阈值(如该压水堆核电厂高阈值为4.0E+4Bq/m 3)或蒸汽发生器排污水取样放射性测量值大于高阈值(如该压水堆核电厂高阈值为2.0E+6Bq/m 3)，叠加一回路稳压器或容控箱液位 低(如该压水堆核电厂选取对应设备最低液位，信号触发时表明一回路存在大量的水装量补给)表征发生SGTR事故。二回路轻微放射性时，通过一回路水位波动信号来进行验证是否真实发生SGTR事故。
(2)一回路LOCA事故
一回路LOCA事故根据LOCA发生的位置不同区分为安全壳内LOCA和安全厂房LOCA，需要利用两个事故逻辑计算单元对两种LOCA事故分别进行诊断。
利用安全壳内LOCA事故逻辑计算单元对是否出现安全壳内LOCA事故进行诊断。请参阅图3，出现以下情况之一即诊断安全壳内LOCA事故已经发生：
(a)安全壳内剂量率测量值达到高高阈值(如该压水堆核电厂高高阈值为1.0E+1Gy/h)，表明一回路压力边界破损，放射性物质突破第二道安全屏障。
(b)安全壳内剂量率测量值达到高阈值(如该压水堆核电厂高高阈值为1.0E-2Gy/h)时，需通过安全壳压力测量验证，如果同时叠加安全壳内压力测量值达到高阈值(如该压水堆核电厂高阈值为1.23bar.a)，则表征发生安全壳内LOCA事故。
(c)一回路封闭状态下，安全壳内压力测量值达到高阈值时，通过一回路过冷度(△Tsat)低或压力容器液位(LRPV)低信号(如该压水堆核电厂阈值为△Tsat＜-ε或LRPV小于一回路热管管底部，△Tsat＜-ε对应于一回路过热的情形，可以确定堆芯正在逐步失水；LRPV小于一回路热管管底部表明堆芯水装量已严重降级)，表征发生了壳内LOCA事故。
利用安全厂房LOCA事故逻辑计算单元对是否出现安全厂房LOCA事故进行诊断。请参阅图4，诊断安全厂房LOCA事故已经发生的逻辑为：通过P25信号(余热排出系统允许连接信号)已生效确认RIS-RHR连接条件，待连接后判断安全厂房压力测量值大于高阈值(如该压水堆核电厂高阈值为0.12Mpa.a，表明安全厂房出现了质能释放)或安全厂房地坑液位测量值大于高阈值(如该 压水堆核电厂高阈值为地坑即将溢流液位值0.15m)，RHR连接后通过一回路放射性泄漏或泄漏冷却剂收集量多来表征安全厂房内发生LOCA事故。
(3)二回路破口事故
利用二回路破口事故逻辑计算单元对是否出现二回路破口事故进行诊断。请参阅图5，出现以下情况之一即诊断二回路破口事故已经发生：
(a)VVP(主蒸汽隔离阀)阀门隔间测量值达到高阈值(如该压水堆核电厂高阈值为75℃，通过远高于环境温度且保持测量仪表反应足够灵敏的温度设计，表明发生主蒸汽泄漏)，用于诊断安全壳外主蒸汽管道破口事故。
(b)ARE(主给水系统)管道隔间测量值达到高阈值(如该压水堆核电厂高阈值为75℃，通过远高于环境温度且保持测量仪表反应足够灵敏的温度设计，表明发生主给水管道泄漏)，用于诊断安全壳外主给水管道破口事故。
(c)一列VDA(主蒸汽释放系统)隔离阀自动隔离信号存在时，需要排出除该蒸汽发生器以外的故障因素干扰，通过三列VDA隔离阀自动隔离信号不存在来排除。主蒸汽管道压力低会导致VDA隔离阀隔离，单列的主蒸汽管道压力骤降，表明该列主蒸汽管道蒸汽出现了不可控泄漏。
(d)两台蒸汽发生器内压差高(如该压水堆核电厂压差高阈值为10bar)且P25信号未生效时，通过叠加三列VDA隔离阀自动隔离信号存在表征发生二回路破口事故。
(4)一回路反应性事故
利用一回路反应性事故逻辑计算单元对是否出现一回路反应性事故进行诊断。请参阅图6，出现以下情况之一即诊断一回路反应性事故已经发生：
(a)一回路总硼浓度测量值低于低阈值(该压水堆核电厂压低阈值及该段落下述阈值均为根据机组技术规格书所要求的阈值)。
(b)超功率或超温△T跳堆信号。
(c)功率量程高/低定值中子注量率高。
(d)中间量程中子注量率高。
(e)源间量程中子注量率高。
(f)功率运行R棒棒位低。
(g)停堆中子注量率高。
(二)失去支持功能事故
失去支持功能事故主要包络机组失电、失冷等事故工况。该类型事故的发生，可为始发事故，也可为叠加事故。
(1)LOOP事故：厂外电源失去事故表征通过相应厂用交流电母线电压失去报警存在来判断该事故发生。利用厂外电源失去事故逻辑计算单元对是否出现厂外电源失去事故进行诊断。
(2)SBO事故：全厂失电事故表征通过相应应急母线和非应急母线电压失去报警存在来判断该事故发生。利用全厂失电事故逻辑计算单元对是否出现全厂失电事故进行诊断。
(3)电气盘失电事故：单一电气盘失电事故表征通过相应电气盘电压失去报警存在来判断该事故发生。利用电气盘失电事故逻辑计算单元对是否出现电气盘失电事故进行诊断。
(4)TLOCC事故：完全丧失冷链事故表征通过RRI/SEC系统(设备冷却水系统或重要厂用水系统)全部失去报警存在来判断该事故发生。利用完全丧失冷链事故逻辑计算单元对是否出现完全丧失冷链事故进行诊断。
(5)RIS-RHR故障事故：RIS-RHR故障事故表征通过RIS-RHR设备故障报警存在来判断该事故发生。利用RIS-RHR故障事故逻辑计算单元对是否出现RIS-RHR故障事故进行诊断。
(三)乏燃料水池事件
乏燃料水池事故区分为乏池失冷事故和乏池失水事故。
(1)乏池失冷事故
利用乏池失冷事故逻辑计算单元对是否出现乏池失冷事故进行诊断。请参阅图7，出现以下情况之一即诊断乏池失冷事故已经发生：
(a)失去全部乏燃料水池冷却列。
(b)乏燃料水池温度测量值达到高高阈值(该压水堆核电厂压高高阈值为78℃，该阈值为乏燃料水池需要紧急人为干预的温度)。
(c)失去部分乏燃料水池冷却列且乏燃料水池温度测量值达到高阈值(该压水堆核电厂压高阈值为50℃，该阈值为乏燃料水池正常运行所达到的最高温度)。
(2)乏池失水事故：乏池失水事故通过乏燃料水池液位测量值达到低阈值来(该压水堆核电厂压低阈值为16.3m，该阈值为防止乏燃料水池液位低于其冷却列吸入口的报警液位)表征。利用乏池失水事故逻辑计算单元对是否出现乏池失水事故进行诊断。
步骤104，在机组事故工况自动诊断画面上显示出全部典型事故工况的诊断结果。
优选的，当诊断结果为核电机组发生典型事故工况(即任一个或多个逻辑计算单元诊断出典型事故工况已发生)时，发出声光报警要求操纵员调用机组事故工况自动诊断画面，操纵员通过机组事故工况自动诊断画面获取核电厂机组典型事故工况诊断结果。发出声光报警可通过主控室声光报警系统实现。
下面以前述压水堆核电厂为例介绍相关配套画面。
事故工况自动诊断画面的一个具体实施例的示意图如图8所示，用于向用户显示典型事故工况自动诊断所需的机组安全信号、专设安全设施状态、典型事故工况诊断结果以及建议执行的事故运行规程。
具体的，事故工况自动诊断画面包括典型事故工况诊断显示区域，典型事故工况诊断显示区域用于显示典型热工水力事故诊断结果、失去支持功能事故诊断结果、乏燃料水池事件诊断结果，通过模块化展示典型事故工况诊断结果 来使操纵员清晰了解相关机组状态信息。优选的，典型事故工况诊断显示区域通过设置对应典型事故工况后的显示指示灯，在指示灯亮时表示该事故触发。
优选的，事故工况自动诊断画面还包括机组状态功能监测区域、机组重要安全信号监测区域、重要设备及专设安全设施状态监测区域、机组状态诊断结果显示区域及建议操作规程指引区域。事故工况自动诊断画面在给出机组典型事故工况诊断结果的同时，还在诊断页面上将重要安全信息和专设安全设施状态信息进行了集成，可以通过相关信息对自动诊断结果进行验证。
优选的，在事故工况自动诊断画面上也可以通过相应的导航链接导向具体典型事故工况的诊断逻辑分解画面，操纵员可以通过事故工况自动诊断画面上相应的导航链接获取具体典型事故工况的诊断逻辑分解画面，相应的诊断逻辑分解画面如图9所示。通过在诊断逻辑分解画面中，参数监测的集成，配合主画面中机组重要安全信号和专设安全设施状态可以直接验证自动诊断逻辑的触发，避免操作员无序查找，可以减少操纵员工作量，并减少事故响应时间，以快速进入机组事故控制阶段。
通过以上描述可知，本发明核电厂机组事故工况监测方法对各典型事故工况并行诊断，相互之间不产生影响，可以同时诊断始发事故和叠加事故；而且，机组安全信号、专设安全设施状态可以实时监测机组设备状态，在相应状态改变时触发相关信号或改变画面图符显示。
请参阅图10，本发明核电厂机组事故工况监测系统包括采集模块10、分析处理模块20、逻辑计算模块30和显示模块40。
采集模块10用于采集与典型事故工况相关的事故工况特征参数、机组重要安全信号和专设安全设施状态。
以某压水堆核电厂为例，本发明对典型事故工况的分类如下：
(一)典型热工水力事故，包括SGTR事故、一回路LOCA事故、一回路反应性事故和二回路破口事故；
(二)失去支持功能事故，包括LOOP事故、SBO事故、电气盘失电事故、TLOCC事故和RIS-RHR故障；
(三)乏燃料水池事件，包括乏池失冷事故和乏池失水事故。
具体的，所述事故工况特征参数可以通过核电厂功能参数测量仪表自动监测和获取，例如，反应堆核功率可以由核仪表系统的功率量程探测器、中间量程探测器和源量程探测器测得；堆芯中子通量、堆芯温度、压力容器水位分别可以由堆芯测量系统的自给能中子探测器和热电偶探测器测得；一回路压力可以由布置在稳压器的反应堆冷却剂系统压力测量仪表和布置在一回路热管段的安全注入系统压力测量仪表测得；一回路温度可以由布置在一回路冷管段和热管段的反应堆冷却剂系统的窄量程温度测量仪表和安全注入系统的宽量程温度测量仪表测得。所述机组重要安全信号，通过使用核电机组正常运行的保护信号或通过使用机组状态参数设计相应逻辑获取。专设安全设施是指核电厂为确保堆芯热量的排出和安全壳的完整性，限制事故的发展和减轻事故后果而专门设置的安全设施，专设安全设施状态可以通过机组事故工况专用的NSSS功能可用性信息监测得到。
分析处理模块20用于对事故工况特征参数进行分析处理，筛选出不在预设阈值范围内的异常特征参数。
具体的，分析处理模块20包括阈值比较单元，阈值比较单元用于将事故工况特征参数与预设阈值进行比较，不在预设阈值范围内的参数判断为异常特征参数。
优选的，分析处理模块20还包括预处理单元，预处理单元用于在将事故工况特征参数与预设阈值进行比较之前，先对事故工况特征参数进行预处理，包括以下处理中的一种或两种：(1)根据保守取值原则对事故工况特征参数进行筛选，即将所采集到的多列冗余的数据进行保守取值，如取大或者取小，从而获得进行初步筛选的数据，以提高数据分析的准确度和有效性；(2)对事故工 况特征参数进行无效性筛选，删除无效数据，以剔除对诊断结果产生影响的数据，从而提高数据的准确性和有效性。
逻辑计算模块30，包括多个并行的逻辑计算单元；所述多个逻辑计算单元用于分别按预设逻辑对与该逻辑计算单元相关的异常特征参数、机组重要安全信号、专设安全设施状态进行逻辑计算，诊断是否发生与该逻辑计算单元对应的典型事故工况。
具体的，逻辑计算模块30中对应每一典型事故工况设置至少一个逻辑计算单元，不同典型事故工况的逻辑计算单元并行运行，分别对不同典型事故工况进行逻辑计算，从而在最短时间内诊断出每一典型事故工况是否发生。
具体的，逻辑计算模块30包括：
SGTR事故逻辑计算单元，用于对是否出现SGTR事故进行诊断；
安全壳内LOCA事故逻辑计算单元，用于对是否出现安全壳内LOCA事故进行诊断；
安全厂房LOCA事故逻辑计算单元，用于对是否出现安全厂房LOCA事故进行诊断；
二回路破口事故逻辑计算单元，用于对是否出现二回路破口事故进行诊断；
一回路反应性事故逻辑计算单元，用于对是否出现一回路反应性事故进行诊断；
LOOP事故逻辑计算单元，用于对是否出现LOOP事故进行诊断；
SBO事故逻辑计算单元，用于对是否出现SBO事故进行诊断；
电气盘失电事故逻辑计算单元，用于对是否出现电气盘失电事故进行诊断；
TLOCC事故逻辑计算单元，用于对是否出现TLOCC事故进行诊断；
RIS-RHR故障事故逻辑计算单元，用于对是否出现RIS-RHR故障事故进行诊断；
乏池失冷事故逻辑计算单元，用于对是否出现乏池失冷事故进行诊断；和
乏池失水事故逻辑计算单元，用于对是否出现乏池失水事故进行诊断。
逻辑计算模块30中各事故逻辑计算单元的计算逻辑如前述方法实施例中所述，此处不再赘述。
显示模块40，用于在机组事故工况自动诊断画面上显示出全部典型事故工况的诊断结果。
显示模块40连接有报警模块，报警模块用于当诊断结果为核电机组发生典型事故工况(即任一个或多个逻辑计算单元诊断出发生典型事故工况)时，发出声光报警要求操纵员调用机组事故工况自动诊断画面。报警模块可以是主控室声光报警系统。
下面以前述压水堆核电厂为例介绍相关配套画面。
事故工况自动诊断画面的一个具体实施例的示意图如图8所示，用于向用户显示典型事故工况自动诊断所需的机组安全信号、专设安全设施状态、典型事故工况诊断结果以及建议执行的事故运行规程。
具体的，事故工况自动诊断画面包括典型事故工况诊断显示区域，典型事故工况诊断显示区域用于显示典型热工水力事故诊断结果、失去支持功能事故诊断结果、乏燃料水池事件诊断结果，通过模块化展示典型事故工况诊断结果来使操纵员清晰了解相关机组状态信息。优选的，典型事故工况诊断显示区域通过设置对应典型事故工况后的显示指示灯，在指示灯亮时表示该事故触发。
优选的，事故工况自动诊断画面还包括机组状态功能监测区域、机组重要安全信号监测区域、重要设备及专设安全设施状态监测区域、机组状态诊断结果显示区域及建议操作规程指引区域。事故工况自动诊断画面在给出机组典型事故工况诊断结果的同时，还在诊断页面上将重要安全信息和专设安全设施状态信息进行了集成，可以通过相关信息对自动诊断结果进行验证。
优选的，显示模块还用于在事故工况自动诊断画面上通过相应的导航链接导向具体典型事故工况的诊断逻辑分解画面，相应的诊断逻辑分解画面如图9 所示。通过在诊断逻辑分解画面中，参数监测的集成，配合主画面中机组重要安全信号和专设安全设施状态可以直接验证自动诊断逻辑的触发，避免操作员无序查找，可以减少操纵员工作量，并减少事故响应时间，以快速进入机组事故控制阶段。
结合以上对本发明的详细描述可以看出，本发明核电厂机组事故工况监测方法和系统不仅实时监测事故工况特征参数，还能够实时监测机组重要参数信号及专设安全设施状态，自动并行诊断机组各个始发事故或叠加事故，并通过人机交互界面予以显示，辅助操纵员对机组事故工况进行判断和处理。
相对于现有技术，本发明核电厂机组事故工况监测方法和系统至少具有以下有益技术效果：
1)可以对机组状态典型事故工况进行实时、自动监测，通过本发明可以实现在事故情况下的事故始发事故或叠加事故的自动诊断，可帮助操纵员快速了解当前机组状态并对机组状态变化进行预判，提高操纵员事故响应速度，有效减少操纵员控制机组事故过程中产生的人因失效，也可为电厂应急响应组织提供决策依据。
2)通过多个事故逻辑计算单元对各典型事故工况并行诊断，相互之间不产生影响，可以同时诊断始发事故和叠加事故。
3)基于配套画面和逻辑画面的详细设计，集成化显示机组典型事故工况下使用的重要机组信号和专设安全设施状态，让操作员清晰机组需要且真实可靠的系统及功能配置，可以明显提高操纵员对机组事故工况的响应效率。
4)根据事故工况导致机组六大状态参数恶化情况不同的原则，来分别定义出典型事故工况分类，有利于典型事故工况的清晰化展示和区分。
5)在电厂发生事故时自动提示事故的触发原因、用于应对事故的专设安全设施的状态以及建议的操作规程，有助于进一步提高核电厂的智能化和自动化水平。
本说明书中各个实施例采用递进的方式描述，每个实施例重点说明的都是与其他实施例的不同之处，各个实施例之间相同相似部分互相参见即可。对于实施例公开的装置而言，由于其与实施例公开的方法相对应，所以描述的比较简单，相关之处参见方法部分说明即可。
专业人员还可以进一步意识到，结合本文中所公开的实施例描述的各示例的单元及算法步骤，能够以电子硬件、计算机软件或者二者的结合来实现，为了清楚地说明硬件和软件的可互换性，在上述说明中已经按照功能一般性地描述了各示例的组成及步骤。这些功能究竟以硬件还是软件方式来执行，取决于技术方案的特定应用和设计约束条件。专业技术人员可以对每个特定的应用来使用不同方法来实现所描述的功能，但是这种实现不应认为超出本发明的范围。
结合本文中所公开的实施例描述的方法或算法的步骤可以直接用硬件、处理器执行的软件模块，或者二者的结合来实施。软件模块可以置于随机存储器(RAM)、内存、只读存储器(ROM)、电可编程ROM、电可擦除可编程ROM、寄存器、硬盘、可移动磁盘、CD-ROM、或技术领域内所公知的任意其它形式的存储介质中。
以上实施例只为说明本发明的技术构思及特点，其目的在于让熟悉此项技术的人士能够了解本发明的内容并据此实施，并不能限制本发明的保护范围。凡跟本发明权利要求范围所做的均等变化与修饰，均应属于本发明权利要求的涵盖范围。

Claims (15)

  1. 一种核电厂机组事故工况监测方法，其特征在于，所述方法包括以下步骤：
    采集与典型事故工况相关的事故工况特征参数、机组重要安全信号和专设安全设施状态；
    对事故工况特征参数进行分析处理，筛选出不在预设阈值范围内的异常特征参数；
    利用多个逻辑计算单元对各种典型事故工况进行并行诊断，每一逻辑计算单元按预设逻辑对与该逻辑计算单元相关的异常特征参数、机组重要安全信号、专设安全设施状态进行逻辑计算，诊断是否发生与该逻辑计算单元对应的典型事故工况；
    在机组事故工况自动诊断画面上显示出全部典型事故工况的诊断结果。
  2. 根据权利要求1所述的核电厂机组事故工况监测方法，其特征在于，所述典型事故工况包括典型热工水力事故、失去支持功能事故和乏燃料水池事件；其中，典型热工水力事故包括蒸汽发生器传热管破裂事故、一回路破口事故、一回路反应性事故和二回路破口事故；失去支持功能事故包括厂外电源失去事故、全厂失电事故、电气盘失电事故、完全丧失冷链事故和余热排出系统故障；乏燃料水池事件包括乏池失冷事故和乏池失水事故；
    对应每一典型事故工况设置至少一个逻辑计算单元，不同典型事故工况的逻辑计算单元并行运行，分别对不同典型事故工况进行逻辑计算。
  3. 根据权利要求2所述的核电厂机组事故工况监测方法，其特征在于，利用蒸汽发生器传热管破裂事故逻辑计算单元对是否出现蒸汽发生器传热管破裂事故进行诊断，出现以下情况之一即诊断蒸汽发生器传热管破裂事故已经发生：
    一、二回路泄漏率平衡异常信号存在；
    蒸汽发生器二次侧蒸汽管道放射性测量值大于高高阈值或蒸汽发生器排污水取样放射性测量值大于高高阈值；
    蒸汽发生器二次侧蒸汽管道放射性测量值大于高阈值或蒸汽发生器排污水取样放射性测量值大于高阈值，叠加一回路稳压器或容控箱液位低。
  4. 根据权利要求2所述的核电厂机组事故工况监测方法，其特征在于，所述一回路破口事故根据破口发生的位置不同区分为安全壳内破口和安全厂房破口，利用两个事故逻辑计算单元对两种破口事故分别进行诊断；
    利用安全壳内破口事故逻辑计算单元对是否出现安全壳内破口事故进行诊断，出现以下情况之一即诊断安全壳内破口事故已经发生：(a)安全壳内剂量率测量值达到高高阈值；(b)安全壳内剂量率测量值达到高阈值，叠加安全壳内压力测量值达到高阈值；(c)一回路封闭状态下，安全壳内压力测量值达到高阈值，且存在一回路过冷度低或压力容器液位低信号；
    利用安全厂房破口事故逻辑计算单元对是否出现安全厂房破口事故进行诊断，诊断安全厂房破口事故已经发生的逻辑为：余热排出系统允许连接信号已生效，且安全厂房压力测量值大于高阈值或安全厂房地坑液位测量值大于高阈值。
  5. 根据权利要求2所述的核电厂机组事故工况监测方法，其特征在于，利用二回路破口事故逻辑计算单元对是否出现二回路破口事故进行诊断，出现以下情况之一即诊断二回路破口事故已经发生：
    主蒸汽隔离阀阀门隔间测量值达到高阈值；
    主给水系统管道隔间测量值达到高阈值；
    一列主蒸汽释放系统隔离阀自动隔离信号存在，且三列主蒸汽释放系统隔离阀自动隔离信号不存在；
    两台蒸汽发生器内压差高，且余热排出系统允许连接信号未生效，叠加三列主蒸汽释放系统隔离阀自动隔离信号存在。
  6. 根据权利要求2所述的核电厂机组事故工况监测方法，其特征在于，利用一回路反应性事故逻辑计算单元对是否出现一回路反应性事故进行诊断，出现以下情况之一即诊断一回路反应性事故已经发生：
    一回路总硼浓度测量值低于低阈值；
    超功率或超温△T跳堆信号；
    功率量程高/低定值中子注量率高；
    中间量程中子注量率高；
    源间量程中子注量率高；
    功率运行R棒棒位低；
    停堆中子注量率高。
  7. 根据权利要求2所述的核电厂机组事故工况监测方法，其特征在于，所述失去支持功能事故的诊断逻辑为：
    通过相应厂用交流电母线电压失去报警存在来判断所述厂外电源失去事故发生；
    通过相应应急母线和非应急母线电压失去报警存在来判断所述全厂失电事故发生；
    通过相应电气盘电压失去报警存在来判断所述电气盘失电事故发生；
    通过设备冷却水系统或重要厂用水系统全部失去报警存在来判断所述完全丧失冷链事故发生；
    通过余热排出系统设备故障报警存在来判断所述余热排出系统故障事故发生。
  8. 根据权利要求2所述的核电厂机组事故工况监测方法，其特征在于，所述乏池失水事故通过乏燃料水池液位测量值达到低阈值来表征；
    利用乏池失冷事故逻辑计算单元对是否出现乏池失冷事故进行诊断，出现以下情况之一即诊断诊断乏池失冷事故已经发生：
    失去全部乏燃料水池冷却列；
    乏燃料水池温度测量值达到高高阈值；
    失去部分乏燃料水池冷却列且乏燃料水池温度测量值达到高阈值。
  9. 根据权利要求1所述的核电厂机组事故工况监测方法，其特征在于，还包括：当诊断结果为核电机组发生典型事故工况时，发出声光报警要求操纵员调用机组事故工况自动诊断画面；操纵员通过机组事故工况自动诊断画面获取核电厂机组典型事故工况诊断结果，通过事故工况自动诊断画面上相应的导航链接获取具体典型事故工况的诊断逻辑分解画面。
  10. 一种核电厂机组事故工况监测系统，其特征在于，所述系统包括：
    采集模块，用于采集与典型事故工况相关的事故工况特征参数、机组重要安全信号和专设安全设施状态；
    分析处理模块，用于对事故工况特征参数进行分析处理，筛选出不在预设阈值范围内的异常特征参数；
    逻辑计算模块，包括多个并行的逻辑计算单元；所述逻辑计算单元分别用于按预设逻辑对与该逻辑计算单元相关的异常特征参数、机组重要安全信号、专设安全设施状态进行逻辑计算，诊断是否发生与该逻辑计算单元对应的典型事故工况；和
    显示模块，用于在机组事故工况自动诊断画面上显示出全部典型事故工况的诊断结果。
  11. 根据权利要求10所述的核电厂机组事故工况监测系统，其特征在于，所述逻辑计算模块中对应每一典型事故工况设置至少一个逻辑计算单元，不同典型事故工况的逻辑计算单元并行运行，分别对不同典型事故工况进行逻辑计算。
  12. 根据权利要求10所述的核电厂机组事故工况监测系统，其特征在于，所述逻辑计算模块包括：
    蒸汽发生器传热管破裂事故逻辑计算单元，用于对是否出现蒸汽发生器传热管破裂事故进行诊断；
    安全壳内破口事故逻辑计算单元，用于对是否出现安全壳内破口事故进行诊断；
    安全厂房破口事故逻辑计算单元，用于对是否出现安全厂房破口事故进行诊断；
    二回路破口事故逻辑计算单元，用于对是否出现二回路破口事故进行诊断；
    一回路反应性事故逻辑计算单元，用于对是否出现一回路反应性事故进行诊断；
    厂外电源失去事故逻辑计算单元，用于对是否出现厂外电源失去事故进行诊断；
    全厂失电事故逻辑计算单元，用于对是否出现全厂失电事故进行诊断；
    电气盘失电事故逻辑计算单元，用于对是否出现电气盘失电事故进行诊断；
    完全丧失冷链事故逻辑计算单元，用于对是否出现完全丧失冷链事故进行诊断；
    余热排出系统故障事故逻辑计算单元，用于对是否出现余热排出系统故障事故进行诊断；
    乏池失冷事故逻辑计算单元，用于对是否出现乏池失冷事故进行诊断；和
    乏池失水事故逻辑计算单元，用于对是否出现乏池失水事故进行诊断。
  13. 根据权利要求10所述的核电厂机组事故工况监测系统，其特征在于，所述事故工况自动诊断画面用于向用户显示典型事故工况自动诊断所需的机组安全信号、专设安全设施状态、典型事故工况诊断结果以及建议执行的事故运行规程。
  14. 根据权利要求13所述的核电厂机组事故工况监测系统，其特征在于，所述事故工况自动诊断画面包括典型事故工况诊断显示区域、机组状态功能监 测区域、机组重要安全信号监测区域、重要设备及专设安全设施状态监测区域、机组状态诊断结果显示区域及建议操作规程指引区域；其中，典型事故工况诊断显示区域用于显示典型热工水力事故诊断结果、失去支持功能事故诊断结果、乏燃料水池事件诊断结果，通过模块化展示典型事故工况诊断结果来使操纵员清晰了解相关机组状态信息。
  15. 根据权利要求10所述的核电厂机组事故工况监测系统，其特征在于，所述显示模块还用于在事故工况自动诊断画面上通过相应的导航链接导向具体典型事故工况的诊断逻辑分解画面。
PCT/CN2022/103208 2022-04-11 2022-07-01 核电厂机组事故工况监测方法和系统 WO2023197462A1 (zh)

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