CN110853785B - Method for analyzing output capacity fault of nuclear power pressurized water reactor unit - Google Patents
Method for analyzing output capacity fault of nuclear power pressurized water reactor unit Download PDFInfo
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Abstract
The invention discloses a method for analyzing the output capacity fault of a nuclear power pressurized water reactor unit, which comprises the following steps: decomposing the thermodynamic cycle of the nuclear power pressurized water reactor unit into a hot end, a cold end and a middle part, performing fault analysis and judgment on the hot end by adopting a parameter correlation analysis method, performing fault analysis and judgment on the cold end, and performing fault analysis and judgment on the middle part, wherein the fault analysis and judgment on the middle part comprises the steps of integrating the middle part and analyzing the split part; the system and the equipment comprise a steam generator and a thermal power measurement thereof, a steam extraction pipeline of a steam-driven auxiliary water feeding pump, a main steam pipeline, a drainage system, a main steam bypass discharge system, a new steam extraction pipeline of a steam-water separation reheating system, a high-pressure steam inlet valve of a steam turbine and a steam guide pipe; the cold end comprises a condenser, a circulating water system, a vacuumizing system and a vacuum boundary; the middle part comprises a steam turbine body, a steam-water separation reheating system and a backheating heating system.
Description
Technical Field
The invention relates to the field of nuclear power generation, in particular to a method for analyzing output capacity faults of a nuclear power pressurized water reactor unit.
Background
Generally, in the process of tracking the output of the unit, when the deviation of the corrected electric power and an examination test value (or a reference value) is found to exceed 2MW and above, the output capacity of the unit is considered to be abnormal, and then the reason searching is started. In the search, the fault tree shown in fig. 1 is generally used for item-by-item elimination. The searching mode can be used for parallelly classifying the faults affecting the output capacity of the unit, is not strong enough in hierarchy, is time-consuming and labor-consuming to apply, and cannot visually play the function of the statistical data of the output of the unit.
Disclosure of Invention
In view of the above, there is a need to provide a method for analyzing the failure of the output capability of a nuclear pressurized water reactor unit, which fully exerts the function of the output statistical data of the daily unit and improves the efficiency of the failure analysis of the output capability of the unit, and the technical scheme is as follows:
the invention provides a method for analyzing the output capacity fault of a nuclear power pressurized water reactor unit, which comprises the following steps: decomposing the thermodynamic cycle of the nuclear power pressurized water reactor unit into a hot end, a cold end and a middle part, and performing fault analysis and judgment on the hot end, then performing fault analysis and judgment on the cold end and then performing fault analysis and judgment on the middle part by adopting a parameter correlation analysis method, wherein the step of performing fault analysis and judgment on the middle part comprises the steps of analyzing the whole middle part and then analyzing the split parts of the middle part;
the hot end comprises a system and equipment for flowing new steam, wherein the system and equipment for flowing new steam comprise one or more of a steam generator and a thermal power measurement thereof, a steam extraction pipeline of a steam-driven auxiliary water feeding pump, a main steam pipeline and a drain pipe, a main steam bypass discharge system, a new steam extraction pipeline of a steam-water separation reheating system, a high-pressure steam inlet valve of a steam turbine and a steam guide pipe;
the cold end comprises a condenser, a circulating water system, a vacuumizing system and a vacuum boundary;
the middle part comprises a steam turbine body, a steam-water separation reheating system and a backheating heating system.
Further, the step of performing fault analysis and judgment on the hot end includes:
s11, monitoring a deviation value of the output according to the daily output tracking table, and executing S12 if the deviation value is more than 2 MW;
s12, comparing the accumulated value of output force within a certain time with the generator watt-hour meter to determine whether the indicated value of the electric power meter is accurate, if so, executing S13, otherwise, executing S13 after calibrating the electric power meter;
s13, collating output tracking data in a preset recent time period to update the following relation curve;
s14, updating a relation curve of the thermal power of the reactor, the front pressure of the high-pressure cylinder, the first-stage steam extraction pressure, the temperature of main feed water and the equivalent flow of the steam turbine;
s15, updating a relation curve of the temperature of the main water supply and the first-stage steam extraction pressure;
s16, judging whether the positions of the mark points in the recent time period are abnormal or not according to the updated relation curve, if so, judging that the hot end is in fault, otherwise, executing S17;
s17, updating a relation curve of the temperature of the condensate water and the temperature of the seawater;
and S18, judging whether the position of the mark point in the recent time period is abnormal, if so, performing fault analysis and judgment on the cold end, otherwise, performing fault analysis and judgment on the middle part.
Further, after the hot-end fault is determined in step S16, the following measures are further taken:
s19, updating the relation curves of the thermal power of the reactor and the main steam flow, the main water supply flow and the water supply pump flow so as to judge the stability of the thermal power measuring system of the reactor;
and S110, carrying out general survey on system valves related to the main steam, and/or carrying out reliability analysis on an evaporator blowdown flow instrument, and/or checking a reactor thermal power measuring system.
Further, the step S18 of performing fault analysis and judgment on the cold end further includes:
s21, looking up the efficiency test data of the condenser at all times;
and S22, updating the temperature of the condensed water, the oxygen of the condensed water, the temperature of the seawater, the temperature rise of the circulating water and the change trend of the seawater tide level along with the time, and analyzing the change trend.
Further, if the analysis result in step S22 is that the trend of change is abnormal, the following measures are performed:
and detecting whether the air leakage of the vacuum boundary is increased or not, and/or detecting whether the vacuum pump is cavitated or not, and/or detecting whether the cleanliness of the titanium pipe is reduced or not, and/or detecting whether the circulating water flow is reduced or not, and/or detecting whether the area of the condenser is reduced or not.
Further, the step S18 of performing fault analysis and judgment on the intermediate portion further includes: the steam turbine and reheat/reheat system are analyzed.
Further, the analyzing the steam turbine and the reheating/regenerative system comprises:
s31, calculating electric power corrected to the rated main steam flow and the rated back pressure, and comparing the electric power with a reference value to determine whether the performance of the middle part is changed;
s32, looking up the operation log and the alarm record, and judging whether the important valve is abnormal;
s33, carrying out abnormity detection on the pressure ratio of each stage of the high-pressure cylinder and the low-pressure cylinder of the steam turbine body;
s34, carrying out abnormity detection on the heating steam flow, the circulating steam temperature rise, the steam sweeping flow, the drainage flow and the reheating pressure drop of the reheating system;
s35, carrying out abnormity detection on the temperature rise of water supply, the opening degree of a drainage regulating valve, the drainage flow, the upper/lower end difference and the steam extraction pressure loss of the regenerative system;
s36, carrying out abnormity detection on the opening of the adjusting valve of the shaft seal system and the temperature rise of the water side of the shaft seal heater;
and S37, generally checking valves of the thermodynamic system.
Further, in steps S16 and S18, if the deviation value of the updated relation curve from the empirical reference curve exceeds a preset threshold, it is determined that the position of the identification point in the recent period is abnormal.
The invention has the following advantages: according to the output tracking data of the daily unit, the steam turbine principle correlation theory is applied to perform key parameter correlation analysis, thermal failure range definition is realized according to three-section division, and then a conventional method for analyzing the performance of thermal equipment is used for achieving rapid positioning of output capacity loss. The method avoids a large amount of repeated calculation, and is quicker and more intuitive and higher in analysis efficiency compared with the traditional method.
Drawings
In order to more clearly illustrate the technical solutions in the embodiments of the present invention, the drawings needed to be used in the description of the embodiments will be briefly introduced below, and it is obvious that the drawings in the following description are only some embodiments of the present invention, and it is obvious for those skilled in the art to obtain other drawings based on these drawings without creative efforts.
FIG. 1 is a schematic diagram of a prior art power plant capacity analysis fault tree;
FIG. 2 is a flow chart of a method for analyzing a failure of the output capacity of a nuclear power pressurized water reactor unit, which is provided by the embodiment of the invention;
FIG. 3 is a flow chart of a method for analyzing faults at a hot end of a nuclear pressurized water reactor unit according to an embodiment of the invention;
FIG. 4 is a flow chart of a method for analyzing a fault of a nuclear pressurized water reactor unit at a cold end according to an embodiment of the invention;
FIG. 5 is a flow chart of a method for analyzing faults in the middle of a nuclear pressurized water reactor unit according to an embodiment of the invention;
FIG. 6 is a graph of the pre-high pressure of the high pressure cylinder in relation to the thermal power of the reactor in an example application provided by an embodiment of the present invention;
FIG. 7 is a graph of a segment extraction pressure versus reactor thermal power for an example application provided by an embodiment of the present invention;
FIG. 8 is a graph of principal feedwater temperature versus reactor thermal power in an example application provided by an embodiment of the present invention;
FIG. 9 is a graph of principal feedwater temperature versus first stage extraction pressure for an application example provided by an embodiment of the present invention;
fig. 10 is a graph of the temperature of the condensate water (hot well outlet) versus the temperature of the seawater in an application example provided by the embodiment of the present invention.
Detailed Description
In order to make the technical solutions of the present invention better understood, the technical solutions in the embodiments of the present invention will be clearly and completely described below with reference to the drawings in the embodiments of the present invention, and it is obvious that the described embodiments are only a part of the embodiments of the present invention, and not all of the embodiments. All other embodiments, which can be derived by a person skilled in the art from the embodiments given herein without making any creative effort, shall fall within the protection scope of the present invention.
It should be noted that the terms "first," "second," and the like in the description and claims of the present invention and in the drawings described above are used for distinguishing between similar elements and not necessarily for describing a particular sequential or chronological order. It is to be understood that the data so used is interchangeable under appropriate circumstances such that the embodiments of the invention described herein are capable of operation in sequences other than those illustrated or described herein. Furthermore, the terms "comprises," "comprising," and "having," and any variations thereof, are intended to cover a non-exclusive inclusion, such that a process, method, apparatus, article, or device that comprises a list of steps or elements is not necessarily limited to those steps or elements expressly listed, but may include other steps or elements not expressly listed or inherent to such process, method, article, or device.
The key points of the technology of the invention are as follows: the method is characterized in that the knowledge of performance characteristics, common faults, coupling relations among equipment, influence degree of the overall performance and the like of the pressurized water reactor nuclear power unit thermodynamic system, the equipment, the system and the equipment is skillfully mastered, the overall thermodynamic system is decomposed into three sections, namely a hot end, a middle end and a cold end, namely a three-section theory of the pressurized water reactor nuclear power unit thermodynamic cycle, correlation analysis is carried out on key parameters according to daily unit output tracking statistical data, preliminary judgment of thermodynamic faults is given, and then the root cause of output capacity loss is determined by combining information such as operation logs, maintenance conditions, parameter trends and field general investigation, so that the direction is indicated for subsequent operation adjustment and maintenance management.
The main idea of the invention is as follows: the thermal cycle of the nuclear power pressurized water reactor unit is decomposed into three sections, namely a hot end section, a middle section and a cold end section. In combination with common thermal faults of the current domestic nuclear power units, the number of hot ends and cold ends is large, and the number of middle parts is relatively small. Therefore, based on daily unit output tracking data, a parameter correlation analysis method is adopted, the hot end is judged firstly, then the cold end is confirmed, finally, the middle part is analyzed in a mode of 'integral first and split later', and the whole analysis flow is shown in fig. 2.
In an embodiment of the present invention, a method for analyzing a nuclear power pressurized water reactor unit output capacity fault is provided, and referring to fig. 2, the method for analyzing the nuclear power pressurized water reactor unit output capacity fault includes: decomposing the thermodynamic cycle of the nuclear power pressurized water reactor unit into a hot end, a cold end and a middle part, and performing fault analysis and judgment on the hot end, then performing fault analysis and judgment on the cold end and then performing fault analysis and judgment on the middle part by adopting a parameter correlation analysis method, wherein the step of performing fault analysis and judgment on the middle part comprises the steps of analyzing the whole middle part and then analyzing the split parts of the middle part;
the hot end comprises a system and equipment for flowing new steam, wherein the system and equipment for flowing new steam comprise one or more of a steam generator and a thermal power measurement thereof, a steam extraction pipeline of a steam-driven auxiliary water feeding pump, a main steam pipeline and a drain pipe, a main steam bypass discharge system, a new steam extraction pipeline of a steam-water separation reheating system, a high-pressure steam inlet valve of a steam turbine and a steam guide pipe;
the cold end comprises a condenser, a circulating water system, a vacuumizing system and a vacuum boundary;
the middle part comprises a steam turbine body, a steam-water separation reheating system and a backheating heating system.
For the hot end portion: whether the hot end breaks down or not can be basically judged through the relation trend of the thermal power of the reactor, the front pressure of the high-pressure cylinder, the temperature of the main water supply and the pressure of the first-stage steam extraction and the relation curve of the temperature of the main water supply and the pressure of the first-stage steam extraction. And then, whether the thermal power measuring system drifts or not can be distinguished by combining the relation curves of the thermal power of the reactor, the main steam flow, the main water supply flow and the water supply pump flow and the inspection and maintenance conditions of relevant elements and instruments of the thermal power measuring system of the reactor during the recent overhaul. In the interests of stability, whether the thermal power measurement system is abnormal or not, the system valve related to the main steam needs to be generally checked to determine whether the main steam leaks or whether the internal leakage degree is deteriorated or not. Referring to fig. 3, the step of performing fault analysis and judgment on the hot end includes:
s11, monitoring a deviation value of the output according to the daily output tracking table, and executing S12 if the deviation value is more than 2 MW;
s12, the accumulated value of the output force within a certain time (such as 24 hours) is taken and compared with the generator watt-hour meter to confirm whether the indicated value of the electric power meter is accurate, if so, S13 is executed, otherwise, S13 is executed after the electric power meter is calibrated;
s13, collating the output tracking data in a preset recent time period (such as 3-5 days) to update the following relation curve;
s14, updating a relation curve of the thermal power of the reactor, the front pressure of the high-pressure cylinder, the first-stage steam extraction pressure, the temperature of main feed water and the equivalent flow of the steam turbine;
s15, updating a relation curve of the temperature of the main water supply and the first-stage steam extraction pressure;
and S16, judging whether the positions of the mark points in the recent time period are abnormal or not according to the updated relation curve, if so, judging that the hot end is in fault, otherwise, executing S17.
This is because: there are two main types of thermal failures common to the hot side: 1. deviation occurs in thermal power measurement; 2. the main steam leaks, and the new steam leaks to a drainage flash tank (or flash evaporation tank). The present invention does not preclude the presence of other hot-side fault types.
S17, updating a relation curve of the temperature of the condensate water and the temperature of the seawater;
and S18, judging whether the position of the mark point in the recent time period is abnormal, if so, performing fault analysis and judgment on the cold end, otherwise, performing fault analysis and judgment on the middle part.
Further, after the hot-end fault is determined in step S16, the following measures are further taken:
s19, updating the relation curves of the thermal power of the reactor and the main steam flow, the main water supply flow and the water supply pump flow so as to judge the stability of the thermal power measuring system of the reactor;
and S110, carrying out general survey on system valves related to the main steam, and/or carrying out reliability analysis on an evaporator blowdown flow instrument, and/or checking a reactor thermal power measuring system.
For the cold end section: the condenser vacuum measurement is affected by the adhesion and deposition of condensed water in the pressure leading pipeline and is not accurate enough, the exhaust temperature of the low-pressure cylinder is not representative, but the judgment of the normal performance of the whole cold end can be given out by a relation curve of seawater temperature-condensed water temperature and a condensed water-soluble oxygen trend curve. If the cold end is not normal, then the main factors affecting the condenser vacuum are checked one by one according to the sequence of easy before difficult, so that the fault analysis and judgment of the cold end in step S18 is shown in fig. 4, and further includes:
s21, looking up the past condenser efficiency test data, namely looking up the condenser efficiency test data once or many times in the past year;
and S22, updating the temperature of the condensed water, the oxygen of the condensed water, the temperature of the seawater, the temperature rise of the circulating water and the change trend of the seawater tide level along with the time, and analyzing the change trend.
The condenser pressure is determined by three parameters of seawater temperature, circulating water temperature rise and condenser end difference, and the condenser end difference is related to four parameters of circulating water temperature rise, circulating water flow, condenser area and overall heat transfer coefficient. And (3) cold end faults need to be analyzed from factors influencing four parameters of circulating water flow, circulating water temperature rise, condenser area and overall heat transfer coefficient.
Further, if the analysis result in step S22 is that the trend of change is abnormal, the following measures are performed:
and detecting whether the air leakage of the vacuum boundary is increased or not, and/or detecting whether the vacuum pump is cavitated or not, and/or detecting whether the cleanliness of the titanium pipe is reduced or not, and/or detecting whether the circulating water flow is reduced or not, and/or detecting whether the area of the condenser is reduced or not.
For the middle part: compared with a thermal power turbine with the same capacity, the nuclear power turbine and the thermodynamic system have the characteristics of low initial parameter, large flow, through-flow of wet steam, multiple system hydrophobic points and the like. Under general conditions, the performance of equipment such as turbine through-flow, backheat heater, catch water reheater, oxygen-eliminating device is comparatively stable, and the trouble often takes place: the state of the thermodynamic system valve changes (such as emergency drainage, starting steam exhaust, accident exhaust and the like), the drain valve is not tightly sealed due to frequent actions, the valve sealing surface is blown and corroded in the starting process and the like. For the faults, firstly, whether the performance of the middle part has obvious change is measured by correcting the index of the electric power under the two conditions of the rated main steam flow and the rated back pressure; secondly, looking up an operation log and an alarm record, checking whether the states of important valves in the thermodynamic system are abnormal or not, and performing trend analysis on key parameters of the steam turbine body, reheating equipment and reheating equipment to confirm the performance stability of the steam turbine body, the reheating equipment and the reheating equipment; finally, the valves influencing the unit output in the thermodynamic system are subjected to comprehensive general investigation, so that the leakage point is determined. Therefore, the step S18 of performing the fault analysis and determination on the middle portion is as shown in fig. 5, and further includes: the steam turbine and reheat/reheat system are analyzed.
Common thermal failures in the middle section are: 1. the steam pipeline drains water and leaks; 2. draining water in emergency, starting steam exhaust, discharging in accident and other valve inner leakage; 3. a heater failure; 4. a moisture separator reheater fault; 5. and through-flow failure of the steam turbine.
Further, the method for analyzing the steam turbine and the reheating/regenerative system is shown in fig. 5, and includes:
s31, calculating electric power corrected to the rated main steam flow and the rated back pressure, and comparing the electric power with a reference value to determine whether the performance of the middle part is changed;
s32, looking up the operation log and the alarm record, and judging whether the important valve is abnormal;
s33, carrying out abnormity detection on the pressure ratio of each stage of the high-pressure cylinder and the low-pressure cylinder of the steam turbine body;
s34, carrying out abnormity detection on the heating steam flow, the circulating steam temperature rise, the steam sweeping flow, the drainage flow and the reheating pressure drop of the reheating system;
s35, carrying out abnormity detection on the temperature rise of water supply, the opening degree of a drainage regulating valve, the drainage flow, the upper/lower end difference and the steam extraction pressure loss of the regenerative system;
s36, carrying out abnormity detection on the opening of the adjusting valve of the shaft seal system and the temperature rise of the water side of the shaft seal heater;
and S37, generally checking valves of the thermodynamic system.
The step S31 represents the "whole first" abnormality determination of the middle portion, and the steps S32 to S37 represents the "split last" abnormality determination of the middle portion.
Further, in steps S16 and S18, if the deviation value of the updated relation curve from the empirical reference curve exceeds a preset threshold, it is determined that the position of the identification point in the recent period is abnormal.
The above method will be described by taking the H2 machine set as an example. The daily crew capacity tracking found that the current capacity of the H2 crew was reduced by about 4MW compared to the first fuel cycle (numbered H2C 01). For this reason, cause analysis and troubleshooting are initiated. Firstly, comparing the electric power meter used for output tracking with the generator watt-hour meter, and confirming that the electric power meter operates normally and the deviation is within an allowable range. Secondly, screening out the data of full power of the reactor and shutdown of the steam conversion system for the plant from the operation data of 5 fuel cycles, and making a correlation trend between the following parameters:
as shown in fig. 6, 7 and 8, in the graphs, H2C01_ F indicates the initial stage of the 1 st fuel cycle of the H2 plant, H2C01_ R indicates the later stage of the 1 st fuel cycle of the H2 plant, H2C02 indicates the 2 nd fuel cycle, H2C03 indicates the 3 rd fuel cycle, H2C04 indicates the 4 th fuel cycle, and H2C05 indicates the 5 th fuel cycle. The trend of the correlation of the first stage extraction pressure and the principal feedwater temperature, as shown in fig. 9, is substantially linear.
Comprehensively analyzing the associated trend graph to give the following judgment:
under the condition of the same reactor thermal power, the pressure before the high pressure cylinder stage is still 0.4-0.5% lower than that of the initial production stage after H204 overhaul (namely H2C 05). On the premise that the flow area of the steam turbine is constant, the front pressure of the high-pressure cylinder stage is lower by 0.4% -0.5%, and the corresponding electric power reduction is 4-5 MW.
Under the condition of the same reactor thermal power, the steam turbine has low steam inlet flow, which is caused by main steam leakage or illusion caused by virtual high thermal power measurement value.
Subsequent, general examination of the associated valves contained in the hot side revealed no significant valve leakages. Thus, attention is paid to thermal power measurement. Thermal power measurements that deviate from true values are often caused by the feedwater flow measurement subsystem. Therefore, trend comparison is carried out on the measured value of the feed water flow of the thermal power measuring system, the measured value of the main steam flow, the measured value of the venturi feed water flow and the measured value of the main feed water pump flow, however, the instruments of the main steam flow and the venturi flow are over-adjusted since the machine set is operated, so that the relative change of the flow trend is not enough to prove the stability of the feed water flow measurement. The consistency of the thermal power measurement system remains to be further investigated.
After completion of the hot end analysis, cold end analysis was started. The correlation trend of the condensation water temperature at the outlet of the condenser hot well and the seawater temperature is plotted, as shown in fig. 10, wherein before H299 minor repair, the time segment from commercial operation to minor repair of the H2 unit in the first fuel cycle is indicated, and after H299 minor repair, the time segment from minor repair to shutdown and refueling in the first fuel cycle is indicated. It can be seen that the corresponding relation between the temperature of the condensed water and the temperature of the seawater is basically consistent since the H2 machine set is operated by a commercial company, and no obvious abnormality exists in the cold end.
Finally, trend analysis is carried out on the performance indexes and key parameters of the steam turbine and the regenerative/reheat system (namely, the middle part), and no obvious abnormality is seen.
According to the output capacity fault analysis flow shown in fig. 2, the reason analysis of the output capacity reduction is performed on the H2 unit, and the conclusion is that the current output capacity of the H2 unit is reduced compared with the initial production period, and the most possible reason is caused by the fact that the thermal power measurement system is high in fault.
The invention relates to a nuclear power pressurized water reactor unit output capacity fault analysis method, which is characterized in that key parameter correlation analysis is carried out according to daily unit output tracking data and by using a steam turbine principle correlation theory, thermal fault range definition is realized according to three-section type division, and then a conventional method for analyzing the performance of thermal equipment is used for achieving rapid positioning of output capacity loss. The method avoids a large amount of repeated calculation, and is quicker and more intuitive and higher in analysis efficiency compared with the traditional method.
The above description is only for the preferred embodiment of the present invention and is not intended to limit the scope of the present invention, and all modifications of equivalent structures and equivalent processes that can be directly or indirectly applied to other related technical fields using the contents of the present specification and the accompanying drawings are included in the scope of the present invention.
Claims (7)
1. A nuclear power pressurized water reactor unit output capacity fault analysis method is characterized by comprising the following steps: the thermal circulation of the nuclear power pressurized water reactor unit is divided into a hot end, a cold end and a middle part, a parameter correlation analysis method is adopted, fault analysis and judgment are firstly carried out on the hot end, then fault analysis and judgment are carried out on the cold end, and then fault analysis and judgment are carried out on the middle part,
the step of analyzing and judging the fault of the hot end comprises the following steps:
s11, monitoring a deviation value of the output according to the daily output tracking table, and executing S12 if the deviation value is more than 2 MW;
s12, comparing the accumulated value of output force within a certain time with the generator watt-hour meter to determine whether the indicated value of the electric power meter is accurate, if so, executing S13, otherwise, executing S13 after calibrating the electric power meter;
s13, collating output tracking data in a preset recent time period to update the following relation curve;
s14, updating a relation curve of the thermal power of the reactor, the front pressure of the high-pressure cylinder, the first-stage steam extraction pressure, the temperature of main feed water and the equivalent flow of the steam turbine;
s15, updating a relation curve of the temperature of the main water supply and the first-stage steam extraction pressure;
s16, judging whether the positions of the mark points in the recent time period are abnormal or not according to the updated relation curve, if so, judging that the hot end is in fault, otherwise, executing S17;
s17, updating a relation curve of the temperature of the condensate water and the temperature of the seawater;
s18, judging whether the position of the mark point in the recent time period is abnormal or not, if so, performing fault analysis and judgment on the cold end, otherwise, performing fault analysis and judgment on the middle part;
the fault analysis and judgment of the middle part comprises the steps of firstly analyzing the whole middle part and then analyzing the split parts of the middle part;
the hot end comprises a system and equipment for flowing new steam, wherein the system and equipment for flowing new steam comprise one or more of a steam generator and a thermal power measurement thereof, a steam extraction pipeline of a steam-driven auxiliary water feeding pump, a main steam pipeline and a drain pipe, a main steam bypass discharge system, a new steam extraction pipeline of a steam-water separation reheating system, a high-pressure steam inlet valve of a steam turbine and a steam guide pipe;
the cold end comprises a condenser, a circulating water system, a vacuumizing system and a vacuum boundary;
the middle part comprises a steam turbine body, a steam-water separation reheating system and a backheating heating system.
2. The method for analyzing the output capacity fault of the nuclear power pressurized water reactor unit according to claim 1, wherein after the hot end fault is determined in step S16, the following measures are further taken:
s19, updating the relation curves of the thermal power of the reactor and the main steam flow, the main water supply flow and the water supply pump flow so as to judge the stability of the thermal power measuring system of the reactor;
and S110, carrying out general survey on system valves related to the main steam, and/or carrying out reliability analysis on an evaporator blowdown flow instrument, and/or checking a reactor thermal power measuring system.
3. The nuclear power pressurized water reactor unit capacity fault analysis method according to claim 1, wherein the step of performing fault analysis and judgment on the cold end in step S18 further comprises:
s21, looking up the efficiency test data of the condenser at all times;
and S22, updating the temperature of the condensed water, the oxygen of the condensed water, the temperature of the seawater, the temperature rise of the circulating water and the change trend of the seawater tide level along with the time, and analyzing the change trend.
4. The method for analyzing the output capacity fault of the nuclear power pressurized water reactor unit according to claim 3, wherein if the variation trend is abnormal as a result of the analysis in the step S22, the following measures are implemented:
and detecting whether the air leakage of the vacuum boundary is increased or not, and/or detecting whether the vacuum pump is cavitated or not, and/or detecting whether the cleanliness of the titanium pipe is reduced or not, and/or detecting whether the circulating water flow is reduced or not, and/or detecting whether the area of the condenser is reduced or not.
5. The method for analyzing the output capacity fault of the nuclear power pressurized water reactor unit according to claim 1, wherein the step S18 of analyzing and judging the fault of the middle part further comprises the following steps: the steam turbine and reheat/reheat system are analyzed.
6. The method for analyzing the output capacity fault of the nuclear power pressurized water reactor unit according to claim 5, wherein the analyzing the steam turbine and the reheating/reheating system comprises the following steps:
s31, calculating electric power corrected to the rated main steam flow and the rated back pressure, and comparing the electric power with a reference value to determine whether the performance of the middle part is changed;
s32, looking up the operation log and the alarm record, and judging whether the important valve is abnormal;
s33, carrying out abnormity detection on the pressure ratio of each stage of the high-pressure cylinder and the low-pressure cylinder of the steam turbine body;
s34, carrying out abnormity detection on the heating steam flow, the circulating steam temperature rise, the steam sweeping flow, the drainage flow and the reheating pressure drop of the reheating system;
s35, carrying out abnormity detection on the temperature rise of water supply, the opening degree of a drainage regulating valve, the drainage flow, the upper/lower end difference and the steam extraction pressure loss of the regenerative system;
s36, carrying out abnormity detection on the opening of the adjusting valve of the shaft seal system and the temperature rise of the water side of the shaft seal heater;
and S37, generally checking valves of the thermodynamic system.
7. The nuclear power pressurized water reactor unit capacity-output fault analysis method according to claim 5, wherein in steps S16 and S18, if a deviation value of the updated relation curve from the empirical reference curve exceeds a preset threshold value, it is determined that the position of the identification point in the recent time period is abnormal.
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