阅读说明：本技术 一种传热管破裂事故下核电站一回路降压控制方法 (Nuclear power station primary circuit pressure reduction control method under heat transfer pipe fracture accident ) 是由 钱虹 白秀春 苏晓燕 张栋良 于 2019-09-18 设计创作，主要内容包括：本发明涉及一种传热管破裂事故下核电站一回路降压控制方法,包括以下步骤：S1、获取核电站故障信息,以计算得到使能信号；S2、判断冷却剂平均温度是否达到过冷度要求,得到降温判断结果；S3、获取蒸汽发生器隔离信息,结合使能信号和降温判断结果,计算得到更新信号；S4、根据蒸汽发生器二次侧压力和压控系统压力设定值,基于更新信号,得到压控系统压力参考值；S5、获取稳压器压力,结合压控系统压力参考值,以控制喷淋阀的开度值,从而对一回路进行降压。与现有技术相比,本发明通过对核电站的监测数据进行计算处理,能够自动得到压控系统压力参考值,避免了人工操作不准确以及判断时间过长的问题,提高了应急处理的准确度及速度。(The invention relates to a nuclear power station primary circuit pressure reduction control method under a heat transfer pipe rupture accident, which comprises the following steps: s1, acquiring fault information of the nuclear power station to obtain an enabling signal through calculation; s2, judging whether the average temperature of the coolant meets the supercooling degree requirement or not to obtain a cooling judgment result; s3, obtaining isolation information of the steam generator, and calculating to obtain an updating signal by combining the enabling signal and the cooling judgment result; s4, obtaining a pressure control system pressure reference value based on the update signal according to the secondary pressure of the steam generator and the pressure control system pressure set value; and S5, acquiring the pressure of the pressure stabilizer, and combining the pressure reference value of the pressure control system to control the opening value of the spray valve, so as to reduce the pressure of the loop. Compared with the prior art, the method and the device have the advantages that the pressure reference value of the pressure control system can be automatically obtained by calculating and processing the monitoring data of the nuclear power station, the problems of inaccurate manual operation and overlong judgment time are solved, and the accuracy and the speed of emergency treatment are improved.)

1. A nuclear power station primary circuit pressure reduction control method under a heat transfer pipe rupture accident is characterized by comprising the following steps:

s1, acquiring fault information of the nuclear power station to obtain an enabling signal through calculation;

s2, acquiring the secondary pressure of the steam generator and the average temperature of the coolant, and judging whether the average temperature of the coolant meets the supercooling degree requirement to obtain a cooling judgment result;

s3, obtaining isolation information of the steam generator, and calculating to obtain an updating signal by combining the enabling signal and the cooling judgment result;

s4, obtaining a pressure control system pressure reference value based on the update signal according to the secondary pressure of the steam generator and the pressure control system pressure set value;

and S5, acquiring the pressure of the pressure stabilizer, and combining the pressure reference value of the pressure control system to control the opening value of the spray valve, so as to reduce the pressure of the loop.

2. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture accident according to claim 1, wherein the nuclear power plant fault information in step S1 includes a heat transfer pipe rupture signal and a shutdown signal, the enable signal is obtained by performing logical and calculation on the heat transfer pipe rupture signal and the shutdown signal, when both the shutdown signal and the heat transfer pipe rupture signal are "1", the enable signal is "1", and if not, the enable signal is "0".

3. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture as recited in claim 1, wherein said step S2 specifically comprises the steps of:

s21, acquiring secondary pressure of the steam generator, and calculating to obtain saturation temperature of the steam generator according to a relation function of the pressure and the saturation temperature of the steam generator;

s22, obtaining the average temperature of the coolant, and calculating the deviation of the cooling degree by combining the saturation temperature of the steam generator;

s23, judging whether the cooling degree deviation meets the requirement:

-δ＜TP＜+δ

wherein TP represents supercooling degree deviation, and delta represents supercooling degree fluctuation margin;

and S24, if the supercooling degree deviation TP meets the requirement in the step S23, the average temperature of the coolant meets the supercooling degree requirement, the obtained cooling judgment result is '1', otherwise, the average temperature of the cooling degree does not meet the supercooling degree requirement, and the obtained cooling judgment result is '0'.

4. A primary circuit pressure reducing control method for a nuclear power plant in the event of a heat transfer pipe rupture as set forth in claim 3, wherein the calculation formula of the cooling deviation TP in step S22 is:

TP＝Tav-(T2-Ts)

T2＝F(P2)

wherein Tav represents the average coolant temperature, T2 represents the saturation temperature of the failed steam generator, Ts represents the subcooling threshold value, P2 represents the secondary steam generator pressure, F (x) is the steam generator pressure as a function of the saturation temperature, x is the steam generator pressure, and F (P2) is the secondary steam generator pressure as a function of the saturation temperature.

5. The method for controlling the primary circuit pressure drop of the nuclear power plant in the event of a heat transfer tube rupture accident as recited in claim 1, wherein the update signal in step S3 is obtained by logically and-calculating the steam generator isolation information, the enable signal and the cooling determination result, and the update signal is "1" when the steam generator isolation information, the enable signal and the cooling determination result are all "1", and otherwise the update signal is "0".

6. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture accident according to claim 1, wherein the steam generator isolation information in step S3 includes a main steam isolation valve closing signal, a steam-driven auxiliary water-feeding pump steam supply valve closing signal, and a blow-off valve closing signal, wherein the signal values of the steam generator isolation information are obtained by performing logical and calculation on the main steam isolation valve closing signal, the steam-driven auxiliary water-feeding pump steam supply valve closing signal, and the blow-off valve closing signal, and when the main steam isolation valve closing signal, the steam-driven auxiliary water-feeding pump steam supply valve closing signal, and the blow-off valve closing signal are all "1", the steam generator isolation information is "1", otherwise, the steam generator isolation information is "0".

7. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture as recited in claim 1, wherein said step S4 specifically comprises the steps of:

s41, taking the update signal as the switching condition, executing step S42 when the update signal is '1', otherwise executing step S43;

s42, setting the pressure value of the secondary side of the steam generator as a pressure control system pressure reference value;

and S43, setting the pressure set value of the pressure control system as the pressure reference value of the pressure control system, namely maintaining the pressure set value of the current pressure control system unchanged.

8. The primary circuit pressure reduction control method for the nuclear power plant in the event of a heat transfer pipe rupture as recited in claim 1, wherein said step S5 specifically comprises the steps of:

s51, obtaining the pressure of the pressure stabilizer, and then performing difference calculation on the pressure of the pressure stabilizer and a pressure reference value of a pressure control system to obtain pressure deviation;

and S52, controlling the opening value of the spray valve of the voltage stabilizer through the PID control according to the pressure deviation to spray the spray valve, and reducing the pressure of the primary loop.

Technical Field

The invention relates to the technical field of nuclear power station steam generator heat transfer pipe rupture accident treatment, in particular to a nuclear power station primary circuit pressure reduction control method under a heat transfer pipe rupture accident.

Background

The rupture of the heat transfer pipe of the steam generator of the nuclear power plant refers to the rupture of one or more heat transfer pipes in the steam generator, so that the pressure boundary of a primary loop loses integrity, the primary loop is communicated with a secondary loop, and a nuclear leakage accident is caused. For this reason, the existing emergency treatment method is usually that after a heat transfer pipe rupture accident occurs, an operator enters a standard operation rule according to the requirements of a corresponding alarm or technical specification program to perform accident control, and the main control strategy is as follows:

1. identifying and isolating a faulty steam generator;

2. under the condition of ensuring the supercooling degree, controlling a loop to reduce the pressure as soon as possible, and reducing and eliminating leakage;

3. when the pressure of the first loop and the pressure of the second loop tend to be balanced, the reactor is withdrawn to a safe state by adopting a mode of synchronously reducing the pressure of the first loop and the second loop.

The above process requires that the operator continuously judges the master control state and then controls the accident by adopting corresponding means, but in the actual manual operation process, the judgment operation time is long due to the complex steps, so that the manual intervention measures are not timely and accurate enough, and especially when the manual pressure reduction of a loop is carried out, the operation difficulty is further increased due to the excessive conditions needing to be judged, the pressure of the loop cannot be timely, accurately and quickly reduced, and more serious nuclear leakage accidents are easily caused.

Disclosure of Invention

The invention aims to overcome the defects in the prior art and provide a primary circuit pressure reduction control method for a nuclear power station in the event of a heat transfer pipe rupture accident.

The purpose of the invention can be realized by the following technical scheme: a nuclear power station primary circuit pressure reduction control method under a heat transfer pipe rupture accident comprises the following steps:

s1, acquiring fault information of the nuclear power station to obtain an enabling signal through calculation;

s2, acquiring the secondary pressure of the steam generator and the average temperature of the coolant, and judging whether the average temperature of the coolant meets the supercooling degree requirement to obtain a cooling judgment result;

s3, obtaining isolation information of the steam generator, and calculating to obtain an updating signal by combining the enabling signal and the cooling judgment result;

s4, obtaining a pressure control system pressure reference value based on the update signal according to the secondary pressure of the steam generator and the pressure control system pressure set value;

and S5, acquiring the pressure of the pressure stabilizer, and combining the pressure reference value of the pressure control system to control the opening value of the spray valve, so as to reduce the pressure of the loop.

Further, the nuclear power plant fault information in step S1 includes a heat transfer pipe rupture signal and a shutdown signal, the enable signal is obtained by performing logical and calculation on the heat transfer pipe rupture signal and the shutdown signal, when both the shutdown signal and the heat transfer pipe rupture signal are "1", the enable signal is "1", and if not, the enable signal is "0".

Further, the step S2 specifically includes the following steps:

s21, acquiring secondary pressure of the steam generator, and calculating to obtain saturation temperature of the steam generator according to a relation function of the pressure and the saturation temperature of the steam generator;

s22, obtaining the average temperature of the coolant, and calculating the deviation of the cooling degree by combining the saturation temperature of the steam generator;

s23, judging whether the cooling degree deviation meets the requirement:

-δ＜TP＜+δ

wherein TP represents supercooling degree deviation, and delta represents supercooling degree fluctuation margin;

and S24, if the supercooling degree deviation TP meets the requirement in the step S23, the average temperature of the coolant meets the supercooling degree requirement, the obtained cooling judgment result is '1', otherwise, the average temperature of the cooling degree does not meet the supercooling degree requirement, and the obtained cooling judgment result is '0'.

Further, the calculation formula of the cooling degree deviation TP in the step S22 is:

TP＝Tav-(T2-Ts)

T2＝F(P2)

wherein Tav represents the average coolant temperature, T2 represents the saturation temperature of the failed steam generator, Ts represents the subcooling threshold value, P2 represents the secondary steam generator pressure, F (x) is the steam generator pressure as a function of the saturation temperature, x is the steam generator pressure, and F (P2) is the secondary steam generator pressure as a function of the saturation temperature.

Further, the update signal in step S3 is obtained by performing logic and calculation on the steam generator isolation information, the enable signal, and the cooling determination result, where the update signal is "1" when the steam generator isolation information, the enable signal, and the cooling determination result are all "1", and the update signal is "0" otherwise.

Further, in step S3, the steam generator isolation information includes a main steam isolation valve closing signal, a steam-driven auxiliary water-feeding pump steam-feeding valve closing signal, and a blowoff valve closing signal, where a signal value of the steam generator isolation information is obtained by performing logical and calculation on the main steam isolation valve closing signal, the steam-driven auxiliary water-feeding pump steam-feeding valve closing signal, and the blowoff valve closing signal, and when the main steam isolation valve closing signal, the steam-driven auxiliary water-feeding pump steam-feeding valve closing signal, and the blowoff valve closing signal are all "1", the steam generator isolation information is "1", otherwise, the steam generator isolation information is "0".

Further, the step S4 specifically includes the following steps:

s41, taking the update signal as the switching condition, executing step S42 when the update signal is '1', otherwise executing step S43;

s42, setting the pressure value of the secondary side of the steam generator as a pressure control system pressure reference value;

and S43, setting the pressure set value of the pressure control system as the pressure reference value of the pressure control system, namely maintaining the pressure set value of the current pressure control system unchanged.

Further, the step S5 specifically includes the following steps:

s51, obtaining the pressure of the pressure stabilizer, and then performing difference calculation on the pressure of the pressure stabilizer and a pressure reference value of a pressure control system to obtain pressure deviation;

and S52, controlling the opening value of the spray valve of the voltage stabilizer through the PID control according to the pressure deviation to spray the spray valve, and reducing the pressure of the primary loop.

Compared with the prior art, the invention has the following advantages:

the invention uses the logic and result of the heat transfer pipe rupture signal and the shutdown signal as the enabling signal, combines the evaporator isolation information and the cooling judgment result, and uses the logic and result as the updating signal for switching and setting the pressure reference value of the pressure control system, thereby being capable of automatically reducing the pressure after ensuring effective cooling, rapidly and timely reducing the pressure and reliably meeting the requirement of stopping nuclear leakage.

The invention can quickly and accurately set the pressure reference value of the pressure control system through automatic switching, avoids the problems of inaccurate manual operation and overlong judgment time, is beneficial to the follow-up timely control of the opening value of the spray valve by the pressure control system, does not influence the normal operation of the existing pressure control system, and has strong practical operability.

Drawings

FIG. 1 is a flow chart of a method of the present invention;

FIG. 2 is a schematic diagram of a system for implementing a voltage reduction control method in an embodiment;

fig. 3 is a logic structure diagram of the voltage reduction control method in the embodiment.

The notation in the figure is:

Detailed Description

The invention is described in detail below with reference to the figures and specific embodiments.

As shown in fig. 1, a method for controlling the primary circuit depressurization of a nuclear power plant in the event of a heat transfer pipe rupture comprises the following steps:

s1, acquiring fault information of the nuclear power station to obtain an enabling signal through calculation;

s2, acquiring the secondary pressure of the steam generator and the average temperature of the coolant, and judging whether the average temperature of the coolant meets the supercooling degree requirement to obtain a cooling judgment result;

s3, obtaining isolation information of the steam generator, and calculating to obtain an updating signal by combining the enabling signal and the cooling judgment result;

s4, obtaining a pressure control system pressure reference value based on the update signal according to the secondary pressure of the steam generator and the pressure control system pressure set value;

and S5, acquiring the pressure of the pressure stabilizer, and combining the pressure reference value of the pressure control system to control the opening value of the spray valve, so as to reduce the pressure of the loop.

As shown in fig. 2, in this embodiment, monitoring data is read by a nuclear power plant instrumentation and control system, a program of a nuclear power plant primary circuit automatic voltage reduction control logic is compiled after a heat transfer pipe rupture accident is performed on a NETCONTROL system platform, mutual communication between the NETCONTROL system program and the nuclear power plant instrumentation and control system is realized by using OPC communication, specifically, an OPC is used as a communication tool, a variable database corresponding to an acquired monitoring point is established in the NETCONTROL system, data acquired from the nuclear power plant instrumentation and control system is set as an input variable group, data output to the nuclear power plant instrumentation and control system is set as an output variable group, and a variable calculated inside the NETCONTROL system is set as an intermediate variable group. On the basis of setting the finished variable group, editing a script program in a NETCONTROL system according to the logical relation of the automatic voltage reduction control method, and finally realizing automatic voltage reduction control in emergency response after a steam generator heat transfer pipe breakage accident.