CN113793707B - Irregular boron load tracking operation and control method for pressurized water reactor nuclear power plant - Google Patents

Irregular boron load tracking operation and control method for pressurized water reactor nuclear power plant Download PDF

Info

Publication number
CN113793707B
CN113793707B CN202110961988.8A CN202110961988A CN113793707B CN 113793707 B CN113793707 B CN 113793707B CN 202110961988 A CN202110961988 A CN 202110961988A CN 113793707 B CN113793707 B CN 113793707B
Authority
CN
China
Prior art keywords
control
power
temperature
loop
rod group
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.)
Active
Application number
CN202110961988.8A
Other languages
Chinese (zh)
Other versions
CN113793707A (en
Inventor
吴世发
王鹏飞
万甲双
Current Assignee (The listed assignees may be inaccurate. Google has not performed a legal analysis and makes no representation or warranty as to the accuracy of the list.)
Xian Jiaotong University
Original Assignee
Xian Jiaotong University
Priority date (The priority date 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 date listed.)
Filing date
Publication date
Application filed by Xian Jiaotong University filed Critical Xian Jiaotong University
Priority to CN202110961988.8A priority Critical patent/CN113793707B/en
Publication of CN113793707A publication Critical patent/CN113793707A/en
Application granted granted Critical
Publication of CN113793707B publication Critical patent/CN113793707B/en
Active legal-status Critical Current
Anticipated expiration legal-status Critical

Links

Images

Classifications

    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C17/00—Monitoring; Testing ; Maintaining
    • G21C17/02—Devices or arrangements for monitoring coolant or moderator
    • G21C17/022—Devices or arrangements for monitoring coolant or moderator for monitoring liquid coolants or moderators
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C17/00—Monitoring; Testing ; Maintaining
    • G21C17/02—Devices or arrangements for monitoring coolant or moderator
    • G21C17/035—Moderator- or coolant-level detecting devices
    • 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
    • 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

A method for tracking operation and control without boron load of a pressurized water reactor nuclear power plant is characterized in that a control rod group is controlled in a closed-loop fine adjustment mode in a small-range power change process; for the working condition that the reactor core power is more than 70% FP and the steam turbine instantly throws load shedding in a large range more than 50% FP, the method adopts a mode of controlling rod group open-loop rough adjustment and controlling rod group closed-loop fine adjustment to control, and coordinates with a primary loop control system and a secondary loop control system to control, thereby realizing large-range load shedding or shutdown non-stop operation and realizing safe operation; the method greatly reduces boron regulation operation in the operation process, greatly reduces radioactive waste liquid, and improves the automation degree, the economy and the operation safety of the pressurized water reactor nuclear power station.

Description

Irregular boron load tracking operation and control method for pressurized water reactor nuclear power plant
Technical Field
The invention belongs to the technical field of nuclear power science and engineering, and particularly relates to a method for tracking operation and control of a pressurized water reactor nuclear power plant without regulating boron load.
Background
At present, with the increasingly prominent energy problems and environmental problems, nuclear power is used as clean energy which can be applied in a large scale, plays more and more important roles in ensuring energy safety, optimizing energy structures, reducing carbon emission and building beautiful China, and has an important strategic position in energy supply in China. In order to meet the requirement of a power grid on realizing load tracking operation of a nuclear power station, for an active nuclear power station, on the premise of not changing the overall layout of one loop and two loops, research is carried out on the core fuel arrangement and the core control system is improved through redesigning to obtain good load tracking operation capacity, and the method becomes an important means for solving the problem.
The CPR1000 of the second-generation pressurized water reactor nuclear power station is a mainstream reactor type for operating the nuclear power station in China, is operated in a G mode, has poor load tracking capability, needs an operator to manually adjust the boron concentration in the load tracking process, has low automation degree and poor economy, can generate a large amount of radioactive waste liquid, particularly has low boron concentration of a primary coolant at the end of the service life, and is difficult to adjust the reactivity by changing the boron concentration, so the CPR1000 of the G mode is only effective in the first 80 percent of the cycle life, and loses the load tracking capability at 20 percent of the end of the cycle life.
The Mechanical compensation (MSHIM) operation control strategy is an advanced pressurized water reactor core operation and control strategy, and is characterized in that the reactor core power and the axial power distribution of the reactor can be well controlled only by the automatic action of control rod groups (M rod groups and AO rod groups), the frequent boron regulation of operators is not needed, the automation level is high, the radioactive waste liquid generated in the load tracking operation process is greatly reduced, and the non-regulated boron load tracking can be realized in at least 95 percent of the core life.
The advanced MSHIM control strategy is referred, the traditional G operation mode of the CPR1000 nuclear power plant is improved into an advanced irregular boron-adjusting mode so as to realize irregular boron load tracking operation, and the method has important scientific research significance and application value for improving the automation degree, the economy and the operation safety of the second-generation pressurized water reactor nuclear power station in China.
Disclosure of Invention
In order to solve the problems in the background art, the invention aims to provide an irregular boron load tracking operation and control method for a pressurized water reactor nuclear power plant, which has the characteristics of high automation degree, good economy and high safety.
In order to achieve the purpose, the invention provides the following technical scheme:
a method for tracking operation and control without boron regulation of a pressurized water reactor nuclear power plant comprises the following steps:
s1, for a small-range power change process that the reactor core power is not more than 70 percent FP and the steam turbine is instantly thrown to be not more than 50 percent FP, the small-range power change process is controlled in a control rod group closed-loop fine adjustment mode and is coordinated and controlled with a primary loop control system and a secondary loop control system to realize safe operation;
s2, for the large-range load shedding working condition that the reactor core power is more than 70 percent FP and the steam turbine instantly throws load over 50 percent FP, the method adopts the mode of control rod group open-loop rough adjustment and control rod group closed-loop fine adjustment to control, and coordinates with the control systems of the primary loop and the secondary loop to control, so that the large-range load shedding or shutdown non-stop operation is realized, and the safe operation is realized.
The open loop coarse adjustment mode of the control rod group comprises the following steps: the core power is more than 70 percent FP, the steam turbine instantly throws more than 50 percent FP, firstly, the matched control rod group is selected according to the amplitude of the required power reduction, and the control rod group falls into the core under the action of gravity, so that the core power is reduced.
The control rod group closed-loop fine adjustment mode adopts an NSGA-II algorithm to carry out parameter optimization under the transient working condition, and the reactor power P and the average coolant temperature T of the M rod group control systems in the control rod group are subjected to parameter optimization avg Two targets are used for parameter optimization, specifically:
step (1), establishing the following objective function by adopting ITAE evaluation standard:
Figure BDA0003222398990000031
in the formula:
Figure BDA0003222398990000032
in the formula: τ -simulation time/s; p, P ref -normalizing the actual power to the reference power; t, T avg -actual temperature of coolant and reference temperature/° c;
and (2) according to the objective function, adopting an NSGA-II multi-objective optimization algorithm to control the M rods in the control rod group in the system function: and performing multi-objective optimization on a lead time constant of the lead-lag unit, a time constant of the deviation differential unit and upper and lower limits of a temperature dead zone, wherein an optimized control parameter vector x is as follows:
x＝[τ 2 ,τ 3 ,τ 5 ,L d ,L u ]
in the formula: tau is 2 -the lead time constant of the lead-lag unit; tau is 3 -lag time constant of the lead-lag unit; tau is 5 -the time constant of the deviation differentiation unit; l is d -a lower temperature dead band limit; l is u -an upper limit of the temperature dead band,the upper and lower limits of the optimized variable vector x are respectively:
Figure BDA0003222398990000033
and (3) updating corresponding parameters in the M rod control systems in the control rod group according to the optimized parameters obtained in the step (2).
The coordination control with a loop control system specifically comprises:
because the control rod group comprises three control rod groups of an M rod group, an AO rod group and an S shutdown rod group, the M rod and the AO rod are adopted to respectively control the power and the axial power deviation of the reactor, the M rod group controls the nuclear power, the turbine power and the average temperature of the coolant as input signals, and is a three-channel control system, and the protection signals comprise: high neutron flux within the power range, neutron flux rate of change, overtemperature delta T, overpower delta T, high voltage regulator water level, high voltage regulator pressure, too low voltage regulator pressure, low-low steam generator water level, and high-high steam generator water level; the pressure stabilizer comprises pressure control and water level control of the pressure stabilizer, and the water level of the steam generator is controlled by a water supply control system; the over-temperature delta T protection signal prevents the fuel cladding from being burnt out, the over-power delta T protection signal prevents the fuel pellet from being melted, the difference between the over-temperature delta T protection signal and the over-power delta T protection signal and the setting value thereof is required to be more than 0, and the setting value thereof can be respectively calculated by an equation (1) and an equation (2):
Figure BDA0003222398990000041
Figure BDA0003222398990000042
in the formula: k 1 ～K 8 -margin factor under nominal standard conditions; delta T 0 Temperature difference/DEG C between the inlet and the outlet of the reactor core under the rated power standard working condition; t is 0 -average temperature of coolant/° c; p is a radical of 0 -potentiostat reference pressure/MPa; n is 0 -main pump speed/rpm; t is a unit of av -real time coolant inlet and outlet mean temperature/° c; p-actual pressure of stabilizer/MPa; n-main pump speed/rpm; f. of 1 (Delta I) and f 2 (Δ I) -overtemperature Δ T and overpower Δ T protect a piecewise function with respect to Δ I, τ 1 ～τ 5 -a time constant; s represents a complex frequency in the laplace transform.
The coordination control with the two-loop control system specifically comprises:
the two-loop control system includes a steam discharge control system by which excess steam is discharged to ensure safety of the entire nuclear power plant system when the turbine experiences a transient greater than 10% fp step load or greater than 5% fp/min linear load drop, the steam discharge control system being turned on in time when the turbine experiences a large load shedding transient; the steam discharge control system comprises a proportional opening steam discharge valve and a rapid opening steam discharge valve, when the temperature difference between the reference temperature generated by the load level of the steam turbine and the actual average temperature of the core coolant is within a certain minimum limit value, all the valves are closed, when the temperature deviation exceeds the limit value, the steam discharge valve is opened proportionally, when the temperature deviation Max1 is not less than delta T and less than Max2, the first group of rapid opening valves is opened, and when the temperature deviation Max2 is not less than delta T and less than Max3, the second group of rapid opening valves is opened, and the like.
Compared with the prior art, the invention has the beneficial effects that:
according to the invention, the reactivity slow change process of the fuel consumption is compensated by adjusting boron, and the reactivity is quickly adjusted by the control rod, so that the boron adjusting operation in the operation process is greatly reduced, the radioactive waste liquid is greatly reduced, and the automation degree, the economy and the operation safety of the second generation pressurized water reactor nuclear power station in China are improved; and parameters of a three-channel control system are optimized through an NSGA-II algorithm, and the control effect of the reactor core is improved.
Drawings
FIG. 1 is a schematic diagram of the arrangement of the core control rod assembly of the present invention.
FIG. 2 shows the M-rod stacking procedure of the present invention.
FIG. 3 is a schematic diagram of the M-pack, AO-pack control system included in the control stick set of the present invention.
FIG. 4 is a block diagram of the AO rod set control system of the present invention.
FIG. 5 is a block diagram of an M-bar cluster control system according to the present invention.
Fig. 6 is a block diagram of a CPR1000 protection system Simulink of the present invention.
FIG. 7 is a schematic diagram of an over-temperature Δ T and over-power Δ T protection system according to the present invention.
FIG. 8 is a schematic block diagram of a vapor emission control system of the present invention.
Detailed Description
The technical solutions in the embodiments of the present invention will be clearly and completely described below with reference to the accompanying drawings, 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.
A method for tracking operation and control without boron regulation of a pressurized water reactor nuclear power plant comprises the following steps:
s1, for a small-range power change process that the reactor core power is not more than 70% FP and the steam turbine instantly throws power not more than 50% FP, the small-range power change process is controlled by adopting a control rod group closed-loop fine adjustment mode and is coordinated and controlled with a primary loop control system and a secondary loop control system to realize safe operation;
and S2, for the large-range load shedding working condition that the reactor core power is more than 70 percent FP, and the steam turbine instantly throws load over 50 percent FP, the control rod group open-loop rough adjustment and the control rod group closed-loop fine adjustment are adopted for control, and the control rod group open-loop rough adjustment and the control rod group closed-loop fine adjustment are coordinated and controlled with the primary loop control system and the secondary loop control system, so that the large-range load shedding or shutdown and non-shutdown operation is realized, and the safe operation is realized.
The open-loop coarse adjustment mode of the control rod group is as follows: the core power is more than 70 percent FP, the steam turbine instantly throws more than 50 percent FP, firstly, the matched control rod group is selected according to the amplitude of the required power reduction, and the control rod group falls into the core under the action of gravity, so that the core power is reduced.
The control rod group is precisely adjusted in a closed loop mannerIn the method, the NSGA-II algorithm is adopted to carry out parameter optimization under the transient working condition, and the reactor power P and the average coolant temperature T of the M rod group control systems in the control rod group are subjected to parameter optimization avg Two targets are optimized, specifically:
(1) The following objective functions were established using the ITAE evaluation criteria:
Figure BDA0003222398990000061
in the formula:
Figure BDA0003222398990000062
in the formula: τ -simulation time/s; p, P ref -normalizing the actual power to the reference power; t, T avg Actual coolant temperature vs. reference temperature/° c.
(2) According to the objective function, adopting an NSGA-II multi-objective optimization algorithm to perform the following functions on the M rods in the control rod group shown in the figure 5: and performing multi-objective optimization on the lead time constant of the lead-lag unit, the time constant of the deviation differential unit and the upper limit and the lower limit of the temperature dead zone, wherein the optimization control parameters, namely the vector x, are as follows:
x＝[τ 2 ,τ 3 ,τ 5 ,L d ,L u ]
in the formula: tau is 2 -the lead time constant of the lead-lag unit; tau. 3 -lag time constant of lead-lag unit; tau is 5 -the time constant of the deviation differentiation unit; l is d -a lower temperature dead band limit; l is u -upper limit of temperature dead zone, upper and lower limits of optimized variable vector x are respectively:
Figure BDA0003222398990000071
(3) And (3) updating corresponding parameters in the M rod control systems in the control rod group according to the optimized parameters obtained in the step (2).
The coordination control of the loop specifically comprises:
the control rod groups comprise M rod groups 20, AO rod groups 9 and S shutdown rod groups 32, and the three types of control rod groups are 61 groups in total and are arranged in a symmetrical mode, and the detailed grouping and arrangement mode of the control rod groups are shown in FIG. 1. Wherein, the M rod groups are MA, MB, MC, MD and M1 rod groups, the first four rod groups are gray rods and are mainly responsible for power control, and the M1 rod group is black rods and is mainly responsible for shutdown and standby. The AO rod group and the 5S rod groups are black rod groups, and can be completely inserted into the reactor core to realize shutdown and standby in emergency. Referring to fig. 2, a cascade of M rod sets is shown to obtain a more linear reactivity profile. The MA rod, MB rod, MC rod, MD rod and M1 rod groups are all 225 steps, and are inserted into the reactor core in sequence when the power is reduced. Wherein about 1/3 of the top of the MA rods (steps 142-225) overlaps about 1/3 of the bottom of the MB rods, about 1/3 of the top of the MB rods overlaps about 1/3 of the bottom of the MC rods, and so on, and vice versa. Therefore, the M control rod groups can introduce linear reactivity change for the reactor core in the process of lifting power. The M and AO rods are controlled with a double closed loop, as shown in FIG. 3, and control the reactor power and axial power excursion, respectively. The AO closed-loop control system takes the axial power deviation delta I, the axial power deviation target zone delta I and the reference power as input signals, and controls the rod speed of an AO rod according to a control law shown in figure 4, so that the control of the reactor core AO is realized. Wherein the axial power offset AO is:
Figure BDA0003222398990000072
the relationship between AO and axial power deviation Δ I is: Δ I = P T -P B ＝AO×(P T +P B ) (% FP), where: p T Is core upper power/% FP, P B Lower core power/% FP. The M-rod control system controls the M-rod speed according to the control law shown in fig. 5 with the reactor power, the turbine power, and the coolant average temperature as input signals, thereby realizing the control of the reactor power. In the practical process, the power (M rod) and the power distribution control (AO rod) may affect each other, signals of the AO rod and the M rod may occur simultaneously, and when the two control signals conflict, the M rod control system takes precedence. AO control system exceptIn addition to receiving the target AO and actual AO value signals, a lock-off signal from a coolant average temperature control system is also received. When the AO rod and the M rod receive the action signal at the same time, if the movement directions of the AO rod and the M rod are the same, the AO rod is preferentially moved, and at the moment, the AO rod controls the power and the power distribution of the reactor core at the same time; when the AO bar and the M bar move in opposite directions, the locking logic of the M bar priority is still adopted. Because the closed-loop control system has feedback, the closed-loop control system can be insensitive to external disturbance of the system and change of internal parameters, an accurate control system can be constructed by using elements which are not accurate and have low cost, accurate control of object parameters is realized, the closed-loop control is more suitable for accurate adjustment of a slow process, for large-range load shedding transient working conditions from load shedding under a rated load operating condition to service transient state, shutdown without shutdown and the like, overlong adjusting time (about 3000 seconds) is caused by closed-loop control characteristics, M rods are an average temperature control rod group, control rods are driven by temperature difference signals of reference temperature and reactor core average temperature, after most of loads are thrown off by the system, the difference between the reactor core coolant average temperature and the reference temperature is rapidly reduced along with rapid insertion of the control rods, the M rod speed is also slowed, and the temperature change is a slow process, the control rod value of the adopted M rod is about 1/2 of that of the G rod of the CPR1000, the negative reactivity introduced by the same rod inserting speed is only about half of that of the G rod control, so that the power descending process is slowed down, therefore, if a closed loop control strategy only through temperature feedback is adopted, the process that the reactor core power is finally adjusted to be about 30 percent FP by the M rod group is slower, the flexibility is insufficient, the control strategy can provide accurate and rapid enough control when the load is changed in a small range or slowly, but for a large-range load shedding working condition, the power descending process is slowed down due to the characteristic that the control strategy only has the closed loop control, therefore, the original boron irregular control strategy is required to be improved, the closed loop control scheme of the M rod is adopted for the small-range power change process or slowly changing power, and the closed loop fine adjustment mode of the open loop coarse adjustment and the M rod group is adopted for the large-range load shedding working condition, to quickly and accurately reduce the power to a reference value, when the core power is 70% FP or more, the turbine is instantaneously thrown 50% FP or moreThe load of the reactor triggers a rapid power reduction system, a control rod group selected by a control program falls into a reactor core under the action of gravity to rapidly reduce the power of the reactor core, a power mismatch signal of a first loop and a second loop of a nuclear power plant also triggers the action of an irregular boron control system and a steam discharge system, the power of the reactor core is matched with the load of the second loop under the coordination action of the first loop and the second loop, and finally large-range load shedding or shutdown and shutdown operation is realized.
As shown in fig. 5, the M-bar group control, with the nuclear power, the turbine power and the average coolant temperature as input signals, is a three-way control system, comprising: the device comprises a power mismatch channel, a coolant average temperature setting channel and a coolant average temperature measuring channel. The power mismatch channel is an advanced control channel, an input signal of the channel is a nuclear power and steam turbine power mismatch signal, and the control signal enables a control system to quickly react when the power is mismatched, so that an advanced regulation function is provided for the control system, and a transient peak value is reduced; the system comprises an average temperature fixed value channel, a cooling medium control system and an M rod control system, wherein the average temperature fixed value channel is used for receiving signals from a steam turbine of a two-circuit system, the load of the steam turbine generates a set value of the average temperature of the cooling medium through an average temperature fixed value function, and the set value of the average temperature is provided for the M rod control system after being filtered; the average temperature measuring channel measures the average temperature of the coolant in the loop, and the average temperature measuring channel provides a coolant average temperature measuring signal for the M-rod control system through the filter and the lead-lag unit. After summing or difference solving, the three-channel signals are sent to a rod speed program unit to drive the M rods to act, so that the control on the reactor core power is realized; as shown in FIG. 4, the AO rod control system principle takes the reactor power and the axial power deviation as input signals, the actual power signal generates an axial power deviation reference value through a function generator, the reference value is compared with the actual value, and the deviation signal enters AAnd the O rod speed program unit drives the AO rod to execute corresponding actions so as to realize the control of the AO. The M bar group control comprises protection signals as follows: see FIG. 6, high/low neutron flux (neutron flux n) in power range>109% or n<25%), high neutron flux rate of change (| dn/d (2 s) & gt)>5%), over temperature Δ T: (<0 deg.C, super power DeltaT: (A) ((B))<0 deg.C), high pressure stabilizer water level (L) p >86%), high regulator pressure (P)>16.55 MPa), too low a stabiliser pressure (P)<13.1 MPa), low-low water level (L) of steam generator UTSG <15%) and steam generator high-high water level (L) UTSG >88.8%). When any one or more signals exceed the protection limit value, the protection system is triggered to act, and all control rods fall into the reactor core in a free falling mode, so that shutdown protection is realized; the pressure stabilizer comprises pressure control and water level control of the pressure stabilizer, and the water level of the steam generator is systematically controlled through water supply control; the over-temperature delta T and the over-power delta T are used for realizing the prediction of the pellet and cladding temperatures of the fuel elements of the reactor core which cannot be measured through signals such as the temperature of a coolant inlet and an outlet of the reactor core which can be measured, the pressure of a voltage stabilizer and the like, and preventing the occurrence of meltdown, so that the safety margin of the reactor core under the transient working condition can be reflected, the over-temperature delta T protection signal prevents the fuel cladding from being burnt down, the over-power delta T protection signal prevents the fuel pellets from being melted, the difference between the over-temperature delta T protection signal and the over-power delta T protection signal and the setting value thereof is required to be more than 0, and the setting values can be respectively calculated by an equation (1) and an equation (2), as shown in figure 7:
Figure BDA0003222398990000101
Figure BDA0003222398990000102
in the formula: k 1 ～K 8 -margin factor under nominal standard conditions; delta T 0 Temperature difference/DEG C between the inlet and the outlet of the reactor core under the rated power standard working condition; t is 0 -average temperature of coolant/° c; p is a radical of 0 -potentiostat reference pressure/MPa; n is 0 -main pump speed/rpm; t is a unit of av -real time coolant inlet and outlet mean temperature/° c; p-actual pressure of the pressure stabilizer/MPa; n-main pump speed/rpm; f. of 1 (Delta I) and f 2 (Δ I) -over-temperature Δ T and over-power Δ T protect the piecewise function with respect to Δ I, τ 1 ～τ 5 -a time constant; s represents a complex frequency in the laplace transform.
The two-loop coordination control specifically comprises the following steps:
the two-loop control system comprises a steam turbine rotating speed control system, a reheater outlet temperature control system, a condenser shell side steam pressure control system, a deaerator water level control system and a steam discharge control system; wherein the vapor emission control system is open loop control; the method comprises the steps that turbine rotating speed control, reheater outlet temperature control, condenser shell side steam pressure control and deaerator water level control are PI control systems, through the PI control systems, steam turbine rotating speed control adjusts steam inlet flow through changing the opening degree of a steam inlet valve of a high-pressure cylinder, the rotating speed of a steam turbine is controlled, reheater outlet temperature control obtains required outlet steam temperature through changing the size of pipe side heating steam flow, condenser shell side steam pressure control changes the condensation rate of shell side steam through changing circulating cooling water flow, and finally the purpose of controlling shell side steam pressure is achieved, deaerator water level control changes the flow of feed water pumped to a deaerator through changing the pump speed of a condensation water extraction pump, and finally the deaerator water level is close to a reference water level, through a steam discharge system, when the steam turbine experiences the following transient state, namely the steam turbine steam discharge system is larger than 10 FP step load, or larger than 5 FP/min linear load reduction, redundant steam is discharged to ensure the safety of the whole nuclear power plant system, and when the steam turbine experiences transient load reduction transient state, the steam discharge control system is opened in time, redundant heat is discharged to avoid the situation that a coolant stops due to cause excessive expansion pressure to exceed a safety protection limit value, and a loop is triggered by a large amplitude; as shown in fig. 8, the steam discharge control system includes a proportional opening steam discharge valve and a quick opening steam discharge valve, all of which are closed when a temperature difference between a reference temperature generated by a load level of the turbine and an actual average temperature of the core coolant is within a certain minimum limit, the steam discharge valve is proportionally opened when a temperature deviation exceeds the limit, a first group of the quick opening valves is opened when a temperature deviation Max1 is not less than Δ T < Max2, a second group of the quick opening valves is opened when a temperature deviation Max2 is not less than Δ T < Max3, and so on.
Finally, it should be noted that: although the present invention has been described in detail with reference to the foregoing embodiments, it will be apparent to those skilled in the art that changes may be made in the embodiments and/or equivalents thereof without departing from the spirit and scope of the invention. Any modification, equivalent replacement, or improvement made within the spirit and principle of the present invention should be included in the protection scope of the present invention.

Claims (5)

1. A method for tracking operation and control without regulating boron load of a pressurized water reactor nuclear power plant is characterized by comprising the following steps:
s1, for a small-range power change process that the reactor core power is not more than 70% FP and the steam turbine instantly throws power not more than 50% FP, the small-range power change process is controlled by adopting a control rod group closed-loop fine adjustment mode and is coordinated and controlled with a primary loop control system and a secondary loop control system to realize safe operation;
s2, for the large-range load shedding working condition that the reactor core power is more than 70 percent FP and the steam turbine instantly throws load over 50 percent FP, the method adopts the mode of control rod group open-loop rough adjustment and control rod group closed-loop fine adjustment to control, and coordinates with the control systems of the primary loop and the secondary loop to control, so that the large-range load shedding or shutdown non-stop operation is realized, and the safe operation is realized.
2. The method for the irregular boron load tracking operation and control of the pressurized water reactor nuclear power plant according to claim 1, wherein the open loop rough adjustment mode of the control rod set is as follows: the core power is more than 70 percent FP, the steam turbine instantly throws more than 50 percent FP, firstly, the matched control rod group is selected according to the amplitude of the required power reduction, and the control rod group falls into the core under the action of gravity, so that the core power is reduced.
3. The method as claimed in claim 1, wherein the control rod group is refined in closed loop mode, and parameter optimization is performed by using NSGA-II algorithm under transient condition, so as to optimize reactor power P and coolant average temperature T of M rod group control systems in the control rod group avg Two targets are optimized, specifically:
step (1), establishing the following objective function by adopting an ITAE evaluation standard:
Figure FDA0003222398980000011
in the formula:
Figure FDA0003222398980000021
in the formula: τ -simulation time/s; p, P ref -normalizing the actual power to the reference power; t, T avg -actual temperature of coolant and reference temperature/° c;
and (2) according to the objective function, adopting an NSGA-II multi-objective optimization algorithm to control the M rods in the control rod group in the system function: and performing multi-objective optimization on a lead time constant of the lead-lag unit, a time constant of the deviation differential unit and upper and lower limits of a temperature dead zone, wherein an optimized control parameter vector x is as follows:
x＝[τ 2 ,τ 3 ,τ 5 ,L d ,L u ]
in the formula: tau is 2 -lead time constant of the lead-lag unit; tau is 3 -lag time constant of the lead-lag unit; tau. 5 -the time constant of the deviation differentiation unit; l is d -a lower temperature dead band limit; l is u -upper limit of temperature dead zone, upper and lower limits of optimized variable vector x are respectively:
Figure FDA0003222398980000022
and (3) updating corresponding parameters in the M rod control systems in the control rod group according to the optimized parameters obtained in the step (2).
4. The method for tracking operation and control without boron regulation in a pressurized water reactor nuclear power plant as claimed in claim 1, wherein the method is coordinated with a loop control system, and specifically comprises the following steps:
because the control rod group comprises three control rod groups of an M rod group, an AO rod group and an S shutdown rod group, the M rod and the AO rod are adopted to respectively control the power and the axial power deviation of the reactor, the M rod group controls the nuclear power, the turbine power and the average temperature of the coolant as input signals, and is a three-channel control system, and the protection signals comprise: high neutron flux within the power range, neutron flux rate of change, overtemperature delta T, overpower delta T, high voltage regulator water level, high voltage regulator pressure, too low voltage regulator pressure, low-low steam generator water level, and high-high steam generator water level; the pressure stabilizer comprises pressure control and water level control of the pressure stabilizer, and the water level of the steam generator is controlled by a water supply control system; the over-temperature delta T protection signal prevents the fuel cladding from being burnt out, the over-power delta T protection signal prevents the fuel pellet from being melted, the difference between the over-temperature delta T protection signal and the over-power delta T protection signal and the setting value thereof is required to be more than 0, and the setting value thereof can be respectively calculated by an equation (1) and an equation (2):
Figure FDA0003222398980000031
Figure FDA0003222398980000032
in the formula: k 1 ～K 8 -margin factor under nominal standard conditions; delta T 0 Temperature difference/DEG C between the inlet and the outlet of the reactor core under the rated power standard working condition; t is 0 -average temperature of coolant/° c; p is a radical of 0 Voltage regulator referencepressure/MPa; n is a radical of an alkyl radical 0 -main pump speed/rev/min; t is av -real time coolant inlet and outlet mean temperature/° c; p-actual pressure of the pressure stabilizer/MPa; n-main pump speed/rpm; f. of 1 (Delta I) and f 2 (Δ I) -over-temperature Δ T and over-power Δ T protect the piecewise function with respect to Δ I, τ 1 ～τ 5 -a time constant; s represents a complex frequency in the laplace transform.
5. The method for tracking operation and controlling non-regulated boron load of a pressurized water reactor nuclear power plant according to claim 1, wherein the method is coordinated with a two-loop control system, and specifically comprises the following steps:
the two-loop control system includes a steam discharge control system by which excess steam is discharged to ensure safety of the entire nuclear power plant system when the turbine experiences a transient greater than 10% fp step load or greater than 5% fp/min linear load drop, the steam discharge control system being turned on in time when the turbine experiences a large load shedding transient; the steam discharge control system comprises a proportional opening steam discharge valve and a rapid opening steam discharge valve, when the temperature difference between the reference temperature generated by the load level of the steam turbine and the actual average temperature of the core coolant is within a certain minimum limit value, all the valves are closed, when the temperature deviation exceeds the limit value, the steam discharge valve is opened proportionally, when the temperature deviation Max1 is not less than delta T and less than Max2, the first group of rapid opening valves is opened, and when the temperature deviation Max2 is not less than delta T and less than Max3, the second group of rapid opening valves is opened, and the like.
CN202110961988.8A 2021-08-20 2021-08-20 Irregular boron load tracking operation and control method for pressurized water reactor nuclear power plant Active CN113793707B (en)

Priority Applications (1)

Application Number Priority Date Filing Date Title
CN202110961988.8A CN113793707B (en) 2021-08-20 2021-08-20 Irregular boron load tracking operation and control method for pressurized water reactor nuclear power plant

Applications Claiming Priority (1)

Application Number Priority Date Filing Date Title
CN202110961988.8A CN113793707B (en) 2021-08-20 2021-08-20 Irregular boron load tracking operation and control method for pressurized water reactor nuclear power plant

Publications (2)

Publication Number Publication Date
CN113793707A CN113793707A (en) 2021-12-14
CN113793707B true CN113793707B (en) 2022-12-09

Family

ID=79181926

Family Applications (1)

Application Number Title Priority Date Filing Date
CN202110961988.8A Active CN113793707B (en) 2021-08-20 2021-08-20 Irregular boron load tracking operation and control method for pressurized water reactor nuclear power plant

Country Status (1)

Country Link
CN (1) CN113793707B (en)

Citations (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
KR20130028213A (en) * 2011-09-09 2013-03-19 한국수력원자력 주식회사 Automatic reactor regulating system for controlling reactor coolant temperature and axial power distribution
CN105139908A (en) * 2015-06-30 2015-12-09 中国核动力研究设计院 Control modes for PWR (Pressurized Water Reactor) core of nuclear power plant
CN106531236A (en) * 2016-12-15 2017-03-22 中广核工程有限公司 Rod changing system and method for control rod of nuclear power plant reactor
CN107945889A (en) * 2017-11-28 2018-04-20 中国核动力研究设计院 A kind of implementation method of nuclear reactor Steam Generator in Load Follow
CN111462925A (en) * 2020-04-17 2020-07-28 西安交通大学 Nuclear reactor power adjusting method and system based on operation data

Patent Citations (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
KR20130028213A (en) * 2011-09-09 2013-03-19 한국수력원자력 주식회사 Automatic reactor regulating system for controlling reactor coolant temperature and axial power distribution
CN105139908A (en) * 2015-06-30 2015-12-09 中国核动力研究设计院 Control modes for PWR (Pressurized Water Reactor) core of nuclear power plant
CN106531236A (en) * 2016-12-15 2017-03-22 中广核工程有限公司 Rod changing system and method for control rod of nuclear power plant reactor
CN107945889A (en) * 2017-11-28 2018-04-20 中国核动力研究设计院 A kind of implementation method of nuclear reactor Steam Generator in Load Follow
CN111462925A (en) * 2020-04-17 2020-07-28 西安交通大学 Nuclear reactor power adjusting method and system based on operation data

Non-Patent Citations (1)

* Cited by examiner, † Cited by third party
Title
CNP650压水堆不调硼负荷跟踪可行性研究;刘同先等;《核科学与工程》;20161015(第05期);全文 *

Also Published As

Publication number Publication date
CN113793707A (en) 2021-12-14

Similar Documents

Publication Publication Date Title
CN101788809B (en) Coordinated control system (CCS) of large-size circulating fluidized bed boiler (CFBB) unit
CN107420874B (en) Ultra-supercritical thermal generator set coordination control system
CN102437809B (en) Method for controlling speed regulating system of hydroelectric unit in island mode
CN110360540B (en) Boiler main steam temperature control method for power grid AGC load instruction frequent fluctuation
CN105785859A (en) Optimization control system and method ensuring primary frequency modulation motion under AGC mode of thermal power generating unit
CN105135409B (en) Supercritical unit boiler master controller control method based on primary frequency modulation action amplitude
CN102629131A (en) Coordinative control method for pulse-mode regulation of condensate flow to participate in frequency modulation
CN113452092B (en) Control method for improving AGC comprehensive frequency modulation performance index of thermal power generating unit
CN111045321B (en) Method for coordinately controlling embedded internal model controller under deep peak regulation
CN112648029B (en) Coordinated control optimization method for deep peak regulation working condition of thermal power plant
CN108361683B (en) Full load section reheat temperature intelligent control system
CN111462925A (en) Nuclear reactor power adjusting method and system based on operation data
CN111399372A (en) Steam turbine main control transformer PID control method
CN113793707B (en) Irregular boron load tracking operation and control method for pressurized water reactor nuclear power plant
CN116382200A (en) Full-load self-stabilization system of unit and operation method thereof
CN114415496A (en) Condensed water throttling system and method for thermal power generating unit
CN112072678A (en) Primary frequency modulation control method for wind turbine generator
CN113419459A (en) Thermal power generating unit CCS-TF control method based on energy storage system
CN112268271A (en) Undisturbed switching control method for dry-wet state conversion of ultra-supercritical unit
CN114776397B (en) Automatic control method for main steam pressure of small steam turbine generator unit
CN211261305U (en) High pressure feed-forward compensation device for controlling water level of high pressure heater dynamically
CN216281315U (en) Main steam temperature optimization control device of double-slag-chamber coal-fired unit
CN113027545B (en) Bypass control method for stabilizing steam drum liquid level of gas-steam combined cycle unit
CN115301391B (en) Intelligent control shutdown method for direct-blowing pulverizing system
CN117008463A (en) Energy-saving control method for recirculation valve regulation of steam pump in deep peak regulation mode

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
GR01 Patent grant
GR01 Patent grant