CN116168854A - Reactor pressure vessel protection method and system - Google Patents
Reactor pressure vessel protection method and system Download PDFInfo
- Publication number
- CN116168854A CN116168854A CN202310155996.2A CN202310155996A CN116168854A CN 116168854 A CN116168854 A CN 116168854A CN 202310155996 A CN202310155996 A CN 202310155996A CN 116168854 A CN116168854 A CN 116168854A
- Authority
- CN
- China
- Prior art keywords
- cold
- pressure vessel
- overpressure protection
- reactor pressure
- state
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Pending
Links
Images
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C9/00—Emergency protection arrangements structurally associated with the reactor, e.g. safety valves provided with pressure equalisation devices
- G21C9/004—Pressure suppression
- G21C9/012—Pressure suppression by thermal accumulation or by steam condensation, e.g. ice condensers
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C13/00—Pressure vessels; Containment vessels; Containment in general
-
- 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
Landscapes
- Physics & Mathematics (AREA)
- Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
Abstract
The invention relates to a reactor pressure vessel protection method and system, wherein the method comprises the following steps: when the unit is in a downlink state, judging whether the waste heat discharging system is in an effective state, and if so, generating a cold overpressure protection signal for activating a cold overpressure protection means; if the judgment result is negative, further judging whether the average temperature of a loop is smaller than the first conversion temperature, and if the judgment result is positive, generating a cold overpressure protection signal; when the unit is in an uplink state, judging whether the average temperature of a loop is greater than a second switching temperature, and if so, generating a thermal state overpressure protection signal for activating a thermal state overpressure protection means; the invention can be switched between the cold overpressure protection state and the hot overpressure protection state based on the actual condition according to the average temperature control unit of the loop, thereby pertinently executing the corresponding overpressure protection means and effectively preventing the reactor pressure vessel from being broken.
Description
Technical Field
The invention relates to the technical field of nuclear power plant equipment protection, in particular to a method and a system for protecting a reactor pressure vessel.
Background
The primary loop of the nuclear power plant mainly comprises a reactor pressure vessel, a voltage stabilizer, a main pump, a steam generator, a main pipeline and other devices, wherein the reactor pressure vessel contains nuclear fuel for generating heat, and other devices contain reactor coolant for conducting heat. In any event, the nuclear power plant must monitor and ensure the integrity of the primary circuit to prevent potential radioactive emissions from the loss of reactor coolant.
Among all the breaks of the primary circuit devices, the consequences of the reactor pressure vessel break are often the most serious, since the nuclear fuel contained therein is extremely radioactive, and the nuclear power plant does not have any means of isolating the break after the break, which would lead to a continuous leakage of coolant and radioactivity. Therefore, the nuclear power plant has the most stringent requirements on the manufacturing process of the reactor pressure vessel to ensure its reliability. At the same time, the primary circuit pressure must be monitored strictly during the operation of the nuclear plant, ensuring at all times that it does not exceed the allowable pressure of the reactor pressure vessel (the highest pressure that the plant is allowed to withstand during use), which is temperature-dependent. Based on the material characteristics of the reactor pressure vessel, the allowable pressure of the reactor pressure vessel decreases with the decrease of temperature, and particularly when the temperature reaches the ductile-brittle transition temperature, the allowable pressure of the reactor pressure vessel decreases sharply, so that the reactor pressure vessel becomes easier to break.
At present, although the nuclear power plant designs corresponding protection schemes aiming at the thermal state and the cold state of the unit, the technical means of the protection schemes are large in difference, the association relationship between the thermal state and the cold state is weak, and the design of protection fixed values of some equipment (such as a safety valve) is also greatly different, so that the thermal state and the cold state of the unit are blocked in switching, and the safety of the nuclear power plant is reduced.
Disclosure of Invention
The invention aims to provide a reactor pressure vessel protection method and a reactor pressure vessel protection system.
The technical scheme adopted for solving the technical problems is as follows: a reactor pressure vessel protection method is constructed, comprising the steps of:
when the unit is in a downlink state, judging whether the waste heat discharging system is in an effective state, and if so, generating a cold overpressure protection signal for activating a cold overpressure protection means; if the judging result is negative, further judging whether the average temperature of a loop is smaller than the first conversion temperature, and if the judging result is positive, generating the cold overpressure protection signal;
when the unit is in an uplink state, judging whether the average temperature of the first loop is greater than a second switching temperature, and if so, generating a thermal overpressure protection signal for activating a thermal overpressure protection means.
Preferably, the cold overpressure protection means comprises:
the safety valve opening threshold of the one-loop voltage stabilizer is set to at least one of a cold state safety threshold, a predetermined unnecessary device is closed, and a safety valve of the waste heat discharging system is started.
Preferably, the thermal overpressure protection means comprises:
setting the relief valve opening threshold to at least one of a power relief threshold, opening the predetermined unnecessary device, and disabling the waste heat removal system relief valve.
Preferably, the reactor pressure vessel protection method further comprises:
and performing cold overpressure analysis to calculate the cold safety threshold.
Preferably, the cold overpressure analysis comprises:
the unit is in a cold state, only a voltage stabilizer safety valve in the overvoltage protection equipment is started, the pressure of a loop is increased, the maximum bearing pressure of the voltage stabilizer safety valve under various transient working conditions is collected, corresponding protection thresholds are calculated by combining the overvoltage protection criteria of the various transient working conditions, and then the corresponding protection thresholds are selected as the cold state safety thresholds according to the type of the transient working conditions where the loop is located; the transient working conditions comprise two types of transient working conditions, three types of transient working conditions and four types of transient working conditions.
Preferably, the overpressure protection criterion for the second-class transient working condition is: the transient pressure of the voltage stabilizer does not exceed 105% of the design pressure of the primary circuit;
the overpressure protection criteria of the three transient working conditions are as follows: the transient pressure of the voltage stabilizer does not exceed 120% of the design pressure;
the overpressure protection criteria of the four transient working conditions are as follows: the transient pressure of the voltage stabilizer does not exceed 130% of the design pressure.
Preferably, in the cold overpressure analysis, the means for increasing the pressure of the first circuit comprises:
injecting the lowest temperature liquid into the first loop at maximum flow rate by controlling a safety injection system in the unit; or alternatively
Injecting mass energy by controlling chemical and volumetric control systems in the unit; or alternatively
By controlling the reactor coolant pumps in the unit, operation is continued.
Preferably, the first conversion temperature is calculated as:
wherein T is 1 For the first transition temperature, T NDT For the ductile-brittle transition temperature of the reactor pressure vessel ΔRT NDT For the ductile-brittle transition temperature increase of the reactor pressure vessel with the use time,% Cu is the copper content of the reactor pressure vessel,% P is the phosphorus content of the reactor pressure vessel, and f is the maximum neutron fluence of the current inner surface of the reactor pressure vessel.
Preferably, the second switching temperature is equal to the sum of the first switching temperature and a set return difference value.
Preferably, the reactor pressure vessel protection method further comprises: the cold safety threshold is limited to be less than the maximum pressure value that the reactor pressure vessel can withstand at the ductile-brittle transition temperature.
The invention also constructs a reactor pressure vessel protection system comprising:
the first judging unit is used for judging whether the waste heat discharging system is in an effective state when the unit is in a downlink state, and generating a cold overpressure protection signal for activating a cold overpressure protection means when the waste heat discharging system is in an effective state;
the second judging unit is used for judging whether the average temperature of a loop is smaller than the first conversion temperature when the waste heat discharging system is not in an effective state, and generating the cold overpressure protection signal when the average temperature of the loop is smaller than the first conversion temperature; and
and the third judging unit is used for judging whether the average temperature of the first loop is greater than the second switching temperature when the unit is in an uplink state, and generating a thermal state overpressure protection signal for activating a thermal state overpressure protection means when the average temperature of the first loop is greater than the second switching temperature.
Preferably, the reactor pressure vessel protection system further comprises:
the first execution unit is used for executing the cold overpressure protection means when the cold overpressure protection signal is received; the cold overpressure protection means comprises at least one of setting a safety valve opening threshold of a loop voltage stabilizer as a cold safety threshold, closing a preset unnecessary device and starting a safety valve of a waste heat discharging system; and/or
The second execution unit is used for executing the thermal overvoltage protection means when the thermal overvoltage protection signal is received; the thermal overpressure protection signal includes at least one of setting the relief valve opening threshold to a power relief threshold, opening the predetermined unnecessary device, and disabling the waste heat removal system relief valve.
The implementation of the invention has the following beneficial effects: providing a reactor pressure vessel protection method; when the unit is in a downlink state, if the waste heat discharging system is in an effective state, generating a cold overpressure protection signal to activate a cold overpressure protection means, so that the unit is in the cold overpressure protection state, and the reactor pressure vessel of the unit is prevented from being embrittled along with the temperature reduction under the cold state, and then the reactor pressure vessel is broken; when the unit is in an uplink state, if the average temperature of a loop is greater than the second switching temperature, a thermal overpressure protection signal is generated to activate a thermal overpressure protection means, so that the unit is in a thermal overpressure protection state, and the overpressure rupture of a reactor pressure vessel of the unit in a cold state due to overhigh temperature is prevented; the invention generates a cold overpressure protection signal or a hot overpressure protection signal according to the average temperature of a loop so as to control a unit to switch between the cold overpressure protection state and the hot overpressure protection state based on actual conditions, and pertinently execute corresponding overpressure protection means based on the toughness characteristic of the reactor pressure vessel changing along with the temperature change, thereby effectively preventing the reactor pressure vessel from being broken and playing a positive role in improving the safety of a nuclear power plant.
Drawings
The invention will be further described with reference to the accompanying drawings and examples, in which:
FIG. 1 is a flow chart of a method of protecting a reactor pressure vessel in some embodiments of the invention;
fig. 2 is a block diagram of a reactor pressure vessel protection system in accordance with some embodiments of the invention.
Detailed Description
For a clearer understanding of technical features, objects and effects of the present invention, a detailed description of embodiments of the present invention will be made with reference to the accompanying drawings.
It should be noted that the flow diagrams depicted in the figures are merely exemplary and do not necessarily include all of the elements and operations/steps, nor are they necessarily performed in the order described. For example, some operations/steps may be decomposed, and some operations/steps may be combined or partially combined, so that the order of actual execution may be changed according to actual situations.
The block diagrams depicted in the figures are merely functional entities and do not necessarily correspond to physically separate entities. That is, the functional entities may be implemented in software, or in one or more hardware modules or integrated circuits, or in different networks and/or processor devices and/or microcontroller devices.
Referring to fig. 1, a flow chart of a method of protecting a reactor pressure vessel in accordance with some embodiments of the invention is shown. The protection method is used for preventing the reactor pressure vessel from being broken due to overpressure, and comprises the steps S1 and S2.
Step S1: when the unit is in a downlink state, judging whether the waste heat discharging system is in an effective state, and if so, generating a cold overpressure protection signal for activating a cold overpressure protection means; if the judgment result is negative, further judging whether the average temperature of a loop is smaller than the first conversion temperature, and if the judgment result is positive, generating a cold overpressure protection signal;
step S2: when the unit is in an uplink state, judging whether the average temperature of a loop is greater than a second switching temperature, and if so, generating a thermal state overpressure protection signal for activating a thermal state overpressure protection means.
The descending state refers to the process of the unit from the power running state to the shutdown and refueling state.
The ascending state refers to the process of the unit from the shutdown refueling state to the power running state after the unit starts to heat and boost.
Overpressure refers to a phenomenon in a system where the system fluid pressure increases and reaches or exceeds the design pressure due to thermodynamic imbalance, excessive pump flow, or other similar phenomenon.
The cold state refers to that during the shutdown and refueling of the unit, a loop is communicated with the atmosphere, and the reactor coolant is in a normal temperature and pressure state. In a nuclear power plant, it is determined that the reactor coolant is cold when the coolant temperature is lower than 180 ℃ and the coolant pressure is lower than 3.2 mpa.abs.
Thermal state refers to the reactor coolant being in a high temperature and pressure state during power generation in a nuclear power plant so as to transfer heat generated by nuclear fuel to the secondary loop.
The waste heat discharging system is used for carrying out waste heat of the reactor core in a cold state and ensuring that the temperature of a loop is within an allowable range.
In this embodiment, when the unit is in a downlink state, if the residual heat removal system is in an effective state, it is indicated that the reactor is shut down, and the residual heat of the reactor core is being discharged, that is, the reactor pressure vessel is being rapidly cooled (compared with natural cooling), in order to avoid the reactor pressure vessel from cracking due to the sudden drop of temperature to the ductile-brittle transition temperature, a cold overpressure protection signal needs to be generated, so as to activate the cold overpressure protection means, to make the unit in a cold overpressure protection state, prevent the reactor pressure vessel from embrittling due to the temperature drop of the unit in a cold state, and then cause the reactor pressure vessel to crack, and further, if the residual heat removal system is not in effect, when the average temperature of a loop is reduced below the first transition temperature, a cold overpressure protection signal is generated, so as to prevent the reactor pressure vessel from cracking;
when the unit is in an uplink state, if the average temperature of a loop is greater than the second switching temperature, a thermal overpressure protection signal is generated to activate a thermal overpressure protection means, so that the unit is in a thermal overpressure protection state, and the overpressure rupture of a reactor pressure vessel of the unit due to overhigh temperature in a thermal state is prevented;
according to the embodiment, a cold overpressure protection signal or a hot overpressure protection signal is generated according to the average temperature of a loop, so that a control unit is switched between the cold overpressure protection state and the hot overpressure protection state based on actual conditions, corresponding overpressure protection means are pertinently executed based on the toughness characteristic that a reactor pressure container changes along with temperature change, the reactor pressure container is effectively prevented from being broken, and positive effects are played on improving the safety of a nuclear power plant.
In an alternative embodiment, the cold overpressure protection means in step S1 comprises: the safety valve opening threshold of the one-loop voltage stabilizer is set to at least one of a cold state safety threshold, a predetermined unnecessary device is closed, and a safety valve of the waste heat discharging system is started.
In this embodiment, the safety valve opening threshold refers to an air pressure threshold at which a safety valve in a loop regulator is opened, for preventing the regulator from being over-pressurized; when the unit is in a hot state, the safety valve opening threshold is generally set to be a power safety threshold (which may be 17 mpa.abs); the cold overpressure protection means is implemented, which means that the unit is in a cold state, and the reactor pressure vessel is embrittled accordingly, so that the cold safety threshold is smaller than the power safety threshold, and the reactor pressure vessel can be prevented from being broken due to overpressure after cooling embrittlement.
The predetermined unnecessary device is a device which can inject mass energy into a loop and can normally operate without using the device after the unit enters a downlink state, such as a device which can raise the temperature or pressure of the loop, for example, a pressurizing pipeline of a safety injection system, so that unexpected increase of the temperature or pressure of the loop in the downlink state can be avoided as much as possible.
The safety valve of the waste heat discharging system refers to a safety valve connected with a loop (the connection relation of the safety valve is irrelevant to whether the waste heat discharging system is effective or not) in the waste heat discharging system; the safety valve of the waste heat discharging system is an important protection means for preventing the overtemperature and the overpressure of the loop, is used for preventing the overpressure of the reactor pressure vessel from cracking, can be matched with the safety valve in the loop pressure stabilizer for use, and improves the protection effect on the reactor pressure vessel. It should be noted that, the waste heat discharging system can only take effect after the unit is in a cold state, so that the opening threshold value of the safety valve of the waste heat discharging system is smaller than the power safety threshold value.
In an alternative embodiment, the reactor pressure vessel protection method further comprises: and (5) performing cold overpressure analysis to calculate a cold safety threshold.
In this embodiment, the function of the cold overpressure analysis is to calculate a suitable cold safety threshold by analyzing the pressure change of the reactor pressure vessel when the unit is cold, and combining the relevant protection criteria of the nuclear power plant.
Further, in one embodiment, the cold overpressure analysis comprises:
the method comprises the steps of enabling a unit to be in a cold state, only starting a voltage stabilizer safety valve in an overvoltage protection device, increasing the pressure of a loop, collecting the maximum bearing pressure of the voltage stabilizer safety valve under various transient working conditions, calculating corresponding protection thresholds by combining the overvoltage protection criteria of various transient working conditions, and selecting the corresponding protection thresholds as the cold state safety thresholds according to the type of the transient working conditions where the loop is located; the transient working conditions comprise two types of transient working conditions, three types of transient working conditions and four types of transient working conditions.
The overpressure protection equipment refers to related equipment for preventing the overpressure of a primary circuit and mainly comprises various safety valves, such as a pressure stabilizer safety valve, a waste heat discharging system safety valve and the like.
The second-class transient working condition refers to a working condition that a loop has disturbance under the normal working condition, and the overpressure protection criterion of the working condition is as follows: the transient pressure of the voltage stabilizer does not exceed 105% of the design pressure of the primary circuit.
Three transient working conditions refer to an emergency working condition of a loop, and the overpressure protection criterion of the working conditions is as follows: the transient pressure of the voltage stabilizer does not exceed 120% of the design pressure of the primary circuit.
Four transient conditions, namely that a loop is in an accident condition or multiple accident sequences, and the overpressure protection criterion of the condition is as follows: the transient pressure of the voltage stabilizer does not exceed 130% of the design pressure of the primary circuit.
In order to prevent the damage to the equipment caused by the overpressure, the unit must be subjected to overpressure protection, such as setting overpressure protection criteria for different working conditions and designing overpressure protection devices (such as safety valves are common). In the embodiment, the cold overpressure analysis firstly makes the unit in a cold state, only enables the safety valve of the voltage stabilizer, namely other equipment in the overpressure protection equipment fails, and prepares for the subsequent targeted analysis of the pressure bearing capacity of the safety valve of the voltage stabilizer; then, by gradually increasing the pressure of a loop, the voltage regulator safety valve is simulated to work under various transient working conditions so as to acquire the maximum bearing pressure of the voltage regulator safety valve under various transient working conditions, so that the corresponding protection threshold under various transient working conditions is calculated by combining the overpressure protection criteria of various transient working conditions, and then the cold state safety threshold is selected according to the current transient working condition type, thereby ensuring the protection effect of the cold state overpressure protection means.
In some embodiments, the pressure of the one circuit may be increased by: injecting a minimum temperature liquid (water or coolant) into a loop at a maximum flow rate by controlling a safety injection system in the unit; or injecting the mass energy through a chemical and volume control system in the control unit; or by controlling the reactor coolant pumps in the unit to continue to operate to increase coolant flow and thus the pressure of the primary circuit.
In an alternative embodiment, the thermal overpressure protection means in step S1 comprises: the safety valve opening threshold is set to at least one of a power safety threshold, opening a predetermined unnecessary device, and disabling a waste heat removal system safety valve.
In an alternative embodiment, the first transition temperature is calculated as:
wherein T is 1 For the first transition temperature, T NDT For the ductile-brittle transition temperature of the reactor pressure vessel ΔRT NDT For the ductile-brittle transition temperature increase of the reactor pressure vessel over time,% Cu is the copper content of the reactor pressure vessel,% P is the phosphorus content of the reactor pressure vessel, and f is the maximum neutron fluence of the current inner surface of the reactor pressure vessel.
Further, in one embodiment, the reactor pressure vessel protection method further comprises: the cold safety threshold is limited to a value less than the maximum pressure that the reactor pressure vessel can withstand at the ductile-brittle transition temperature. Alternatively, the maximum pressure value may be calculated in combination with a material analysis first transition temperature of the reactor pressure vessel. Alternatively, the cold safety threshold is 8Mpa.
In an alternative embodiment, the second transition temperature is equal to the sum of the first transition temperature and the set return difference value. In addition, the set return difference value can be determined according to the measuring range and the accuracy of the instrument for monitoring the first conversion temperature, and 1% of the measuring range of the instrument can be taken. Alternatively, the return difference is set to 3 ℃.
Referring to fig. 2, the present invention also provides a reactor pressure vessel protection system comprising: a first judgment unit 1, a second judgment unit 2, and a third judgment unit 3.
The first judging unit 1 is configured to judge whether the waste heat discharging system is in an effective state when the unit is in a downlink state, and generate a cold overpressure protection signal for activating the cold overpressure protection means when the waste heat discharging system is in an effective state.
The second judging unit 2 is configured to judge whether an average loop temperature is less than the first conversion temperature when the waste heat removal system is not in an effective state, and generate a cold overpressure protection signal when the average loop temperature is less than the first conversion temperature.
The third judging unit 3 is configured to judge whether an average loop temperature is greater than a second switching temperature when the unit is in an uplink state, and generate a thermal overpressure protection signal for activating the thermal overpressure protection means when the average loop temperature is greater than the second switching temperature.
As shown in fig. 2, the reactor pressure vessel protection system further includes: a first execution unit 4. The first execution unit 4 is used for executing a cold overpressure protection means when receiving the cold overpressure protection signal; the cold overpressure protection means includes at least one of setting a relief valve opening threshold of the primary circuit voltage regulator to a cold relief threshold, closing a predetermined unnecessary device, and enabling a waste heat removal system relief valve.
As shown in fig. 2, the reactor pressure vessel protection system further includes: a second execution unit 5. The second execution unit 5 is used for executing a thermal overpressure protection means when receiving the thermal overpressure protection signal; the thermal overpressure protection signal includes at least one of setting a relief valve opening threshold to a power relief threshold, opening a predetermined unnecessary device, and disabling a waste heat removal system relief valve.
In the present specification, each embodiment is described in a progressive manner, and each embodiment is mainly described in a different point from other embodiments, and identical and similar parts between the embodiments are all enough to refer to each other. For the device disclosed in the embodiment, since it corresponds to the method disclosed in the embodiment, the description is relatively simple, and the relevant points refer to the description of the method section.
Those of skill would further appreciate that the various illustrative elements and algorithm steps described in connection with the embodiments disclosed herein may be implemented as electronic hardware, computer software, or combinations of both, and that the various illustrative elements and steps are described above generally in terms of functionality in order to clearly illustrate the interchangeability of hardware and software. Whether such functionality is implemented as hardware or software depends upon the particular application and design constraints imposed on the solution. Skilled artisans may implement the described functionality in varying ways for each particular application, but such implementation decisions should not be interpreted as causing a departure from the scope of the present invention.
The steps of a method or algorithm described in connection with the embodiments disclosed herein may be embodied directly in hardware, in a software module executed by a processor, or in a combination of the two. The software modules may be disposed in Random Access Memory (RAM), memory, read Only Memory (ROM), electrically programmable ROM, electrically erasable programmable ROM, registers, hard disk, a removable disk, a CD-ROM, or any other form of storage medium known in the art.
It is to be understood that the above examples only represent preferred embodiments of the present invention, which are described in more detail and are not to be construed as limiting the scope of the invention; it should be noted that, for a person skilled in the art, the above technical features can be freely combined, and several variations and modifications can be made without departing from the scope of the invention; therefore, all changes and modifications that come within the meaning and range of equivalency of the claims are to be embraced within their scope.
Claims (12)
1. A method of protecting a reactor pressure vessel, comprising the steps of:
when the unit is in a downlink state, judging whether the waste heat discharging system is in an effective state, and if so, generating a cold overpressure protection signal for activating a cold overpressure protection means; if the judging result is negative, further judging whether the average temperature of a loop is smaller than the first conversion temperature, and if the judging result is positive, generating the cold overpressure protection signal;
when the unit is in an uplink state, judging whether the average temperature of the first loop is greater than a second switching temperature, and if so, generating a thermal overpressure protection signal for activating a thermal overpressure protection means.
2. The reactor pressure vessel protection method of claim 1, wherein the cold overpressure protection means comprises:
the safety valve opening threshold of the one-loop voltage stabilizer is set to at least one of a cold state safety threshold, a predetermined unnecessary device is closed, and a safety valve of the waste heat discharging system is started.
3. The reactor pressure vessel protection method of claim 2, wherein the hot overpressure protection means comprises:
setting the relief valve opening threshold to at least one of a power relief threshold, opening the predetermined unnecessary device, and disabling the waste heat removal system relief valve.
4. The reactor pressure vessel protection method of claim 2, further comprising:
and performing cold overpressure analysis to calculate the cold safety threshold.
5. The reactor pressure vessel protection method of claim 4, wherein the cold overpressure analysis comprises:
the unit is in a cold state, only a voltage stabilizer safety valve in the overvoltage protection equipment is started, the pressure of a loop is increased, the maximum bearing pressure of the voltage stabilizer safety valve under various transient working conditions is collected, corresponding protection thresholds are calculated by combining the overvoltage protection criteria of the various transient working conditions, and then the corresponding protection thresholds are selected as the cold state safety thresholds according to the type of the transient working conditions where the loop is located; the transient working conditions comprise two types of transient working conditions, three types of transient working conditions and four types of transient working conditions.
6. The method of claim 5, wherein the overpressure protection criteria for the second transient operating condition is: the transient pressure of the voltage stabilizer does not exceed 105% of the design pressure of the primary circuit;
the overpressure protection criteria of the three transient working conditions are as follows: the transient pressure of the voltage stabilizer does not exceed 120% of the design pressure;
the overpressure protection criteria of the four transient working conditions are as follows: the transient pressure of the voltage stabilizer does not exceed 130% of the design pressure.
7. The method of claim 5, wherein in the cold overpressure analysis, the means for increasing the pressure of the primary circuit comprises:
injecting the lowest temperature liquid into the first loop at maximum flow rate by controlling a safety injection system in the unit; or alternatively
Injecting mass energy by controlling chemical and volumetric control systems in the unit; or alternatively
By controlling the reactor coolant pumps in the unit, operation is continued.
8. The reactor pressure vessel protection method of claim 1, wherein the first transition temperature is calculated as:
wherein T is 1 For the first transition temperature, T NDT For the ductile-brittle transition temperature of the reactor pressure vessel ΔRT NDT For the ductile-brittle transition temperature increase of the reactor pressure vessel with the use time,% Cu is the copper content of the reactor pressure vessel,% P is the phosphorus content of the reactor pressure vessel, and f is the maximum neutron fluence of the current inner surface of the reactor pressure vessel.
9. The reactor pressure vessel protection method of claim 8, wherein the second transition temperature is equal to a sum of the first transition temperature and a set return difference.
10. The reactor pressure vessel protection method of claim 8, further comprising: the cold safety threshold is limited to be less than the maximum pressure value that the reactor pressure vessel can withstand at the ductile-brittle transition temperature.
11. A reactor pressure vessel protection system, comprising:
the first judging unit is used for judging whether the waste heat discharging system is in an effective state when the unit is in a downlink state, and generating a cold overpressure protection signal for activating a cold overpressure protection means when the waste heat discharging system is in an effective state;
the second judging unit is used for judging whether the average temperature of a loop is smaller than the first conversion temperature when the waste heat discharging system is not in an effective state, and generating the cold overpressure protection signal when the average temperature of the loop is smaller than the first conversion temperature; and
and the third judging unit is used for judging whether the average temperature of the first loop is greater than the second switching temperature when the unit is in an uplink state, and generating a thermal state overpressure protection signal for activating a thermal state overpressure protection means when the average temperature of the first loop is greater than the second switching temperature.
12. The reactor pressure vessel protection system of claim 11, further comprising:
the first execution unit is used for executing the cold overpressure protection means when the cold overpressure protection signal is received; the cold overpressure protection means comprises at least one of setting a safety valve opening threshold of a loop voltage stabilizer as a cold safety threshold, closing a preset unnecessary device and starting a safety valve of a waste heat discharging system; and/or
The second execution unit is used for executing the thermal overvoltage protection means when the thermal overvoltage protection signal is received; the thermal overpressure protection signal includes at least one of setting the relief valve opening threshold to a power relief threshold, opening the predetermined unnecessary device, and disabling the waste heat removal system relief valve.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202310155996.2A CN116168854A (en) | 2023-02-13 | 2023-02-13 | Reactor pressure vessel protection method and system |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| CN202310155996.2A CN116168854A (en) | 2023-02-13 | 2023-02-13 | Reactor pressure vessel protection method and system |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| CN116168854A true CN116168854A (en) | 2023-05-26 |
Family
ID=86416145
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| CN202310155996.2A Pending CN116168854A (en) | 2023-02-13 | 2023-02-13 | Reactor pressure vessel protection method and system |
Country Status (1)
| Country | Link |
|---|---|
| CN (1) | CN116168854A (en) |
-
2023
- 2023-02-13 CN CN202310155996.2A patent/CN116168854A/en active Pending
Similar Documents
| Publication | Publication Date | Title |
|---|---|---|
| KR101463440B1 (en) | Passive safety system and nuclear power plant having the same | |
| CN111540483B (en) | Method for coping with small-sized crevasse loss of coolant accident in pressurized water reactor nuclear power plant | |
| Wright | Simulated AP1000 response to design basis small-break LOCA events in APEX-1000 test facility | |
| EP3667678A1 (en) | Depressurisation valve | |
| TW200419586A (en) | Structured approach for risk-informing deterministic safety analyses | |
| CN116168854A (en) | Reactor pressure vessel protection method and system | |
| Zhao et al. | A study of BWR mark I station blackout accident with GOTHIC modeling | |
| Liang et al. | Low temperature overpressurization analysis for CPR1000 | |
| CN112906201B (en) | Design method for loop system overpressure protection in pressurized water reactor operation state | |
| Harwood et al. | Assessment of thermalhydraulic phenomena for external water makeup | |
| Tusheva et al. | Investigation on primary side oriented accident management measures in a hypothetical station blackout scenario for a VVER-1000 pressurized water reactor | |
| Choa et al. | Review on emergency operating procedures for APR1400: feed-and-bleed and RCS rapid cooldown in SLOCA. | |
| Gonzalez-Cadelo et al. | Effects of break location and time uncertainties in small-break and medium-break LOCA sequences with unavailability of HPSI | |
| Li et al. | THE RESEARCH OF OPERATION STRATEGY TO LOSS HEAT SINK OF SMALL MODULAR REACTORS | |
| Wu et al. | Research about the influence of passive residual heat removal system on SGTR accident | |
| Naitoh et al. | Function of Isolation Condenser of Fukushima Unit-1 Nuclear Power Plant | |
| JP2020204498A (en) | Residual heat removal equipment, method of operating the same, and residual heat removal method | |
| Yu et al. | Accident and Safety Analysis | |
| Min et al. | Development of Automated Cycle Counting Algorithms for APR1400 Fatigue Monitoring System | |
| Chen et al. | Effect of power uprate on core damage frequency in a MBLOCA with multiple operator actions | |
| Zhu et al. | Application of Level 2 PSA in the Design of Cavity Injection System for Nuclear Power Plant | |
| Ke et al. | The emergency operating strategy analysis of medium LOCA with MHSI unavailable | |
| Kim et al. | The Assessment of SG U-tubes Integrity During a Postulated Loss of All Heat Sinks in Wolsong NPPs by the CATHENA Code | |
| Leea et al. | Sensitivity Analysis on Pressurizer Safety Valve Capacity for Overpressure Protection of an Integral Reactor | |
| Lin et al. | The ultimate response guideline simulation and analysis by using TRACE for Lungmen ABWR nuclear power plant |
Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| PB01 | Publication | ||
| PB01 | Publication | ||
| SE01 | Entry into force of request for substantive examination | ||
| SE01 | Entry into force of request for substantive examination |