JP2002181984A - Monitoring controller for boiling water reactor - Google Patents

Abstract

PROBLEM TO BE SOLVED: To obtain a monitoring controller improved in the transient characteristics of core and safety in a boiling water reactor for power generation. SOLUTION: When an event of loss of supply water heating occurs and core inlet subcooling degree increases, the monitoring controller increases the margin of thermal integrity in a rated power operation state and simultaneously the margin for generation of neutron flux fluctuation in a low flow/high power operation state. More concretely, when the difference between a ratio of a pseudo fuel surface heat flux calculated by considering a delay corresponding to heat transfer time constant at fuel rods in a core average neutron flux measurement value (APRM) to a rated value and a ratio of a main steam flow rate measurement value to a rated value in a boiling water reactor becomes larger than a predetermined value, a reactor scram command or a selective control rod insertion command is generated only in a condition that an operation point specified with the pseudo fuel surface heat flux and core inlet flow rate has come in a specific region in a two-dimensional coordinates.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a monitoring and control system for improving transient characteristics and stability of a core of a boiling water reactor for power generation, and more particularly to a monitoring system for controlling a reactor core inlet subcool when a feedwater heating loss event occurs. When the power level increases, it is possible to increase the margin for thermal integrity in the rated output operation state, and at the same time to increase the margin for the occurrence of neutron flux vibration in the low flow rate / high output operation state. To a monitoring and control device.

[0002]

2. Description of the Related Art A boiling water nuclear power plant (hereinafter, referred to as a nuclear power plant)
In “BWR”), a reactor pressure vessel 2 is provided in a reactor containment vessel 1 as shown in FIG. 11, and a plurality of fuel assemblies 3 and control A core 5 composed of a rod 4 and the like and a coolant 6 (for example, water) are accommodated. The coolant 6 is forcibly circulated by the recirculation system 7, and uranium 235 (hereinafter referred to as “U23”)
5)), the saturated water and the saturated steam are mixed by receiving the heat generated by the nuclear fission, and move to the upper part of the core 5. Then, it is dried by a steam-water separator and a steam dryer (not shown) and sent to a turbine 9 via a main steam piping system 8 connected to the reactor pressure vessel 2 to drive the turbine 9. The driving of the turbine 9 rotates the generator 10 to generate power. The steam that has worked in the turbine 9 is introduced into the condenser 11 to be condensed, and is sent to the feed water heater 13 by the condensing pump 12 and is supplied to the feed water heater 13.
After the temperature is raised in the reactor, the water is supplied again into the reactor pressure vessel 2 by the water supply pump 14.

FIG. 12 shows a typical operation characteristic diagram of the BWR having the above-described configuration. Normal operation is performed on the rated power curve, the design flow control curve, the stability limit curve, the minimum pump speed curve, the cavitation limit curve, the maximum pump speed curve, the area surrounded by them and on the natural circulation curve. . In the example of FIG. 12, the rated output is achieved when the core flow rate is 85% (point A) to 105% (point B). Reactor operation is typically 8 early in the cycle.
The flow rate is near 5%, and moves to a higher flow side as the cycle burnup advances to utilize the reactivity gain due to the increase in the coolant flow rate. At the end of the cycle, the flow rate is near 105%. In addition, the core power is normally designed so that the core axial power distribution becomes lower strain at the beginning of the cycle, and the void fraction at the core outlet is increased. Thereby, the uranium 238 (hereinafter referred to as “U
238 "), the production of plutonium 239 (hereinafter referred to as" Pu239 "), which is a fissile material, is promoted. From the middle stage to the end of the cycle, the operation of burning Pu239 in addition to U235 is performed, so that the power distribution in the axial direction of the core tends to be upwardly distorted. In a core with improved economy, optimization is performed by combining the above-mentioned core flow rate adjustment and a nuclear design that enables axial power distribution adjustment.

[0004] More specifically, a core of a boiling water reactor is configured as shown in FIG. A plurality of fuel assemblies 30 are inserted into the core, and each of the fuel assemblies 30 is covered by a channel box 31.
In the core, a control rod 32 and a plurality of local power region monitors 33 (hereinafter, “LPRM”) for detecting neutron flux are provided.
33 "). Further, each channel box 31 is provided with a core support plate 34 as shown in FIG.
And the cylindrical shroud 3 supported by the upper lattice plate 35
It is surrounded by 6. The coolant 6 flows into the channel box 31 from below through the orifice of the fuel support fitting 37 and the lower tie plate, is heated by the fuel assembly 30, generates steam (void) by boiling, and generates gas-liquid It becomes a two-phase flow. The fuel assembly effective length of a currently operating commercial BWR is about 3.7 m.

Incidentally, the boiling state in the core is more specific.
Specifically, it is as follows. That is, the two-phase flow is thermally
Subcooled boiling in non-equilibrium state and equilibrium with void
Classified as a certain saturated boiling. Known Document 1 "" THE THERMA
L-HYDRAULICS OF A BOILINGWATER REACTOR -Second Edi
tion- by R.T.Lahey, Jr. & F.J.Moody ”ANS
According to Fig.5-16 of "Rev.
Voids in the state occur locally along the heat transfer surface
However, the enthalpy of the liquid phase at the center of the flow reaches the saturation point
I haven't. On the other hand, voids in saturated boiling state
Means that the liquid phase is saturated enthalpy_fHas reached the gas phase
(Void) enthalpy h_gIs the latent heat of evaporation h_fgso
It is constant. From liquid phase on heat transfer surface in subcooled boiling state
Two representative correlation equations are proposed for the heat flux to the central liquid phase.
And is used in thermal design of nuclear reactors. One
Is a Profile-Fit Model by Zuber-Staub et al.
The other is the Mechanistic Model by Lahey et al. this
According to these models, from the bubble generation section to the central subcool section
Heat flux q_h″ Is the heat flux q on the wall_W″ Besides the center
Subcool part enthalpy h₁, Mixed enthalpy h ',
It is given by a function such as the Ority X. And q_li″
Is h_lOr a phenomenon greatly affected by h '
ing. On the other hand, even in saturated boiling, BWR mainly
In the nucleate boiling region used, vapor bubbles separate from the heat transfer surface.
And the liquid phase enthalpy is constant h _fKept in
The heat flux from the wall q_wIs stable
You.

[0006] As a fuel assembly in the boiling water reactor described above, since the commercial power generation in Japan, 7
Row 7 column type, 8 row 8 column type, improved 8 row 8 column type, high burnup 8
A row 8 column type and a high burnup 9 row 9 column type have been adopted. Due to these improvements, the capacity of fissile material per aggregate is increased, and high burnup and long-term operation cycle are realized by optimizing the distribution of enrichment in the aggregate and optimizing the arrangement of burnable poisons. Due to this and the effective use of the core flow rate adjustment width, the economic efficiency of the core is improved. High burnup From 8x8 fuel to 9x9 fuel, high burnup /
Deterioration of transient characteristics and stability due to nuclear factors due to an increase in the absolute value of the void reactivity coefficient due to prolonged operation cycle is 2
This is prevented by the use of a large water rod. In addition, the deterioration of stability based on thermo-hydraulic factors due to the increase in two-phase pressure loss due to the increase in the aggregate lattice is due to the reduction of the spacer pressure loss coefficient, the eight partial length fuel rods, and the high pressure loss type lower tie plate. It is prevented by adoption.

Here, the transient characteristic means, in addition to a time change of a process parameter such as an output and a pressure when an abnormal transient change at the time of operation such as a loss of heating of feed water or a load cut off of a generator occurs in a plant, It means the thermal integrity of the fuel assembly.
It is desirable that the fuel assembly be able to perform good heat removal during the transient, and it is engineered to design so that 0.1% or more of all fuel rods do not undergo transition boiling during the transient. . There is a minimum critical power ratio (hereinafter, referred to as MCPR) as a parameter relating to the heat removal performance of the fuel, and the MCPR at the time of a transient change is a safety limit MCPR (Safety Limit MCPR; hereinafter, SLMCPR).
MCPR during operation is restricted so that it does not fall below. Operation Limit MCPR (Operation Limit
MCPR; hereinafter referred to as “OLMCPR”) analyzes various transient changes expected to occur during the life of the plant, and a change in MCPR (hereinafter referred to as ΔMCPR) is obtained. Change (hereinafter, “ΔMCPRma
x ”) to the SLMCPR.

OLMCPR = ΔMCPRmax + SL
MCPRΔMCPRmax is determined by a turbine trip bypass valve inoperative in a plant at the early stage of BWR construction having a relatively low scram speed (hereinafter referred to as “conventional scrum plant”), and a recent scram speed of a BWR is high.
In an R plant (hereinafter referred to as a "high-speed scram plant"), it is often determined by the loss of feedwater heating. FIG.
FIG. 5 shows a transient change analysis result when a feedwater heating loss occurs in a high-speed scram plant loaded with 9 × 9 fuel in a rated output / rated flow rate operating state. It is assumed that the core inlet flow rate is in a constant operation (recirculation system manual) state. The neutron flux and the fuel surface heat flux gradually increase with the increase in the moderator density because the core inlet subcool degree increases as the feedwater temperature decreases. As a result, the main steam flow rate also tends to increase.However, since the increase in the core subcool degree causes a decrease in the outlet quality, the rate of increase in the main steam flow rate is
The neutron flux is smaller than the heat flux on the fuel surface. The pseudo-surface heat flux signal obtained by considering the time constant of 6 seconds for the APRM signal which is the measurement result of the neutron flux is about
Reactor scram is occurring when the scram set point reaches 115% in 100 seconds. Because the scram reduces the amount of voids in the core and reduces the two-phase flow pressure loss,
The core inlet flow rate is set to a value larger than the initial value. MCPR decreases due to an increase in heat flux on the surface of the fuel, and ΔMCPR
Reached 0.15, it was mitigated by the reactor scram.

[0009] Next, the stability means that when the operating point is in a low flow rate / high output state at the time of starting or stopping the plant, or when a transient change such as a single recirculation pump trip occurs in the plant, the operation is stopped. It means the neutron flux vibration damping characteristics when the point shifts to low flow / high output. The core is desirably stable in the entire operation region, and is confirmed by analyzing that the reduction ratio, which is a parameter for determining stability, is less than 1.0. Conversely, the operating area with a small margin for the reduction ratio of 1.0 is selected rods (Selected Rods Inserter
on; hereinafter, referred to as “SRI”) and stability limit curves. The types of stability include channel stability focusing on the thermo-hydraulic stability of the highest power channel, core stability, which is neutron flux vibration with the entire core phase aligned,
There is a region stability that is a neutron flux oscillation having a symmetry axis in the core circumferential direction and 180 degrees out of phase. The sensitivity of each stability to the axial power distribution is in a severe direction as the core stability is generally flat, and the channel stability and the regional stability are in a severe direction as the distribution has a lower peak. In core stability, the sensitivity to the axial power distribution differs from other stability.The core stability has a large effect of nuclear feedback, which has a large effect when the power peak is high at a high void fraction. It is because it appears.

FIG. 16 shows the results of a transient change analysis when feedwater heating loss occurs in a high-speed scram plant loaded with 9 × 9 fuel in the minimum pump speed maximum output operation state. The neutron flux and the fuel surface heat flux gradually increase with the increase in the moderator density because the core inlet subcool degree increases as the feedwater temperature decreases. As a result, the main steam flow rate also tends to increase.However, since the increase in the core subcool degree causes a decrease in the outlet quality, the rate of increase in the main steam flow rate is
The neutron flux is smaller than the heat flux on the fuel surface. An increase in the reactor power and an increase in the core inlet subcooling cause deterioration of the stability, so that the channel, the core, and the zone stability reduction ratio tend to increase. In particular, the core stability reduction ratio exceeds 0.8 after about 130 seconds.

Thus, a BWR using 9 × 9 fuel
In the reactor core, transient and stability deterioration has been prevented by various design improvements, but further measures are needed to increase the core sub-cooling degree.

[0012]

The monitoring and control device for a boiling water reactor proposed so far has a method for coping with an abnormal increase in the subcool degree at the core inlet.
By monitoring the simulated fuel surface heat flux, the core inlet flow rate, and the core thermal output, etc., and when the monitored parameters exceed the set values, the transient control is performed by operating the selective control rod insertion mechanism or scram. Deterioration of stability is prevented. However, in order to perform reliable monitoring of cores with even higher burnup and long operation cycles,
It is essential to realize a monitoring and control device that detects an abnormal increase in the subcool degree at the core entrance, which is a cause of changing the above-mentioned monitoring parameters, at the earliest possible stage and with high accuracy, and appropriately copes with it. To do so, the factors that increase the core inlet subcooling, or the core nuclear thermal hydraulic power when the core inlet subcooling degree increases, and the effects on the process volume, take into account the change in individual parameters. Rather, it is essential to focus on the increasing tendency of the deviation between related parameters, and to develop a monitoring and control device provided with appropriate abnormality detection set values.

In view of the above, the present invention provides a selection control rod insertion mechanism or a selection control rod insertion mechanism in a case where an increasing tendency of a deviation between parameters closely related to an increase in a core inlet subcool exceeds a predetermined set value. It is an object of the present invention to obtain a monitoring and control device for a boiling water reactor that effectively increases a margin for transient and stability by operating a scram.

[0014]

According to a first aspect of the present invention, when a deviation between monitoring parameters in a boiling water reactor becomes larger than a predetermined value, a pseudo fuel surface heat flux and a core inlet flow rate are determined. The reactor scram command or the selection control rod insertion command is issued only when the operating point designated by the formula (1) exists within a predetermined area on the two-dimensional coordinates for a certain period of time.

According to a second aspect of the present invention, in the first aspect of the present invention, the deviation between the monitoring parameters in the boiling water reactor is such that the core average neutron flux measurement value (APRM) has a heat transfer time constant at the fuel rod. The ratio of the simulated fuel surface heat flux to the rated value calculated by taking into account the considerable delay,
It is characterized by a deviation from the ratio of the measured value of the main steam flow rate to the rated value.

According to a third aspect of the present invention, in the invention according to the first aspect, the deviation between the monitoring parameters in the boiling water reactor is determined by the core of the boiling water reactor by using a core operation monitoring process computer. At the core inlet subcool degree obtained by performing heat balance calculation including steam generation and work in the turbine system, and heat loss, and the atomic power specified by the reactor heat output and core inlet flow before starting power generation of the plant. It is characterized by the deviation of the core inlet subcool degree obtained using the plant design conditions for each operating point of the furnace.

According to a fourth aspect of the present invention, in the first aspect of the present invention, the deviation between the monitoring parameters in the boiling water reactor includes a measured value of a feed water temperature which is input to a core operation monitoring process computer, Before the start, it is characterized by a deviation of the feedwater temperature obtained by using plant design conditions for each operating point of the reactor specified by the reactor heat output and the core inlet flow rate.

According to a fifth aspect of the present invention, in the first aspect of the present invention, the deviation between the monitoring parameters in the boiling water reactor is such that the maximum peaking coefficient in the radial direction of the core is determined by using a core operation monitoring process computer. The plant design for each reactor operating point specified by the subcooled boiling length calculated for the channel with the largest or the product of the radial peaking coefficient and the axial peaking coefficient and the reactor heat output and core inlet flow rate It is characterized by the deviation of the subcooled boiling length obtained using the conditions.

According to a sixth aspect of the present invention, in the first aspect of the present invention, the deviation between the monitoring parameters in the boiling water reactor is obtained by incorporating a core thermal hydraulic calculation code with a given heat balance as a boundary condition. And a deviation between the core outlet quality calculated by the computer and a predetermined value.

According to a seventh aspect of the present invention, in the first aspect of the present invention, the deviation between the monitoring parameters in the boiling water reactor is a deviation between the pseudo fuel surface heat flux and the turbine steam flow rate or the pseudo fuel surface. It is characterized by a deviation between a heat flux and a turbine control valve opening degree, a deviation between a pseudo fuel surface heat flux and a generator output, or a deviation between a pseudo fuel surface heat flux and a feedwater flow rate.

According to the present invention, the peak position of the core average axial power distribution, which is calculated by the core operation monitoring process computer, moves toward the core inlet despite the fact that the control rod pattern is not changed. When the moving distance exceeds a predetermined distance, the operating point of the reactor specified by the pseudo surface heat flux and the core inlet flow rate has been present in a predetermined area on a two-dimensional coordinate system for a certain period of time. A monitoring and control device for a boiling water reactor characterized by issuing a reactor scram command or a selection control rod insertion command only therewith.

According to a ninth aspect of the present invention, in the invention according to any one of the first to eighth aspects, a deviation between a pair of monitoring parameters represented by a deviation between a pseudo fuel surface heat flux and a main steam flow rate, or , Deviation of the monitored parameter from the design value,
Alternatively, when any of the movement distances exceeds the abnormality detection set point, the measured value of the core inlet flow rate can be regarded as constant within a certain predetermined width, and specified by the pseudo surface heat flux and the core inlet flow rate The reactor scram command or the selection control rod insertion command is issued only when the operating point of the reactor to be performed exists within a predetermined area on the two-dimensional coordinates for a certain period of time.

The invention according to claim 10 is the invention according to claims 1 to 9
In the invention according to any one of the above, the abnormality detection set value for operating the scram or the selection control rod insertion mechanism is dependent on the operating point specified by the pseudo fuel surface heat flux and the core inlet flow rate measurement value. I do.

The invention according to claim 11 is the invention according to claims 1 to 1
0, the deviation between the paired monitoring parameters, or the deviation from the design value of the monitoring parameter, or the moving distance, with respect to the deviation or the moving distance calculated a predetermined time before the present time. Calculate the expansion width from the average value for a certain time, and when the expansion width exceeds the abnormality detection set point for each monitoring parameter, issue a reactor scram command or a selection control rod insertion command It is characterized by.

The twelfth aspect of the present invention is the first aspect of the present invention.
In the invention according to any one of the first to third aspects, a time change rate of the deviation is calculated for each measurement cycle, and when the time change rate exceeds a predetermined set value continuously for a predetermined number of times, or
When the time change rate of the deviation exceeds a predetermined set value, or when the deviation exceeds a predetermined set value, a reactor scram command or a selection control rod insertion command is issued. Features.

The invention according to claim 13 is the invention according to claims 1 to 1
2. In the invention described in any one of (2) and (3), as a method of regarding the core inlet flow rate to be constant within a predetermined range, the above-described monitoring parameter including the deviation between the pseudo fuel surface heat flux and the main steam flow rate may be set to the abnormality detection set value. The difference between the measured value of the core inlet flow rate at the time exceeding, or the measured value of the recirculation pump speed, or the measured value of the recirculation pump driving flow rate a fixed time ago, or the measured value a fixed number of times before the measurement cycle Is confirmed to be within a variation range set for each measured value.

[0027] The invention according to claim 14 is the invention according to claims 1 to 1
2. In the invention described in any one of (2) and (3), as a method for assuming that the core inlet flow rate is constant within a predetermined range, the above-described monitoring parameter including the deviation between the pseudo fuel surface heat flux and the main steam flow rate is set to the abnormality detection set value. When viewed from the exceeded time, the core inlet flow rate change rate, or the recirculation pump speed change rate, or the recirculation pump drive flow rate change at all past predetermined times in each measurement cycle, including the nearest measurement time It is characterized in that the rate is confirmed to be within the set change rate.

[0028] The invention according to claim 15 is the invention according to claims 1 to 1.
2. The method according to claim 1, wherein the core inlet flow rate is regarded as being constant within a predetermined range.
4. The method according to claim 4, wherein both of the methods of confirming the rate of change for each measurement cycle at least twice are used.

The invention according to claim 16 is the invention according to claims 1 to 1
5. In the invention according to any one of the aspects 5, when the core flow rate can be considered to be constant, when the above-described monitoring parameter including the deviation between the pseudo fuel surface heat flux and the main steam flow rate exceeds the abnormality detection set value, the core It is characterized in that an abnormal increase in the entrance subcool degree is output as a diagnosis result.

The present invention is characterized in that it depends on the operating point specified by the measured values of the surface heat flux and the core inlet flow rate.

[0031]

Embodiments of the present invention will be described below. FIG. 1 is a block diagram showing a first embodiment of the present invention. In FIG. 1, reference numeral 50 denotes a time constant circuit. At that time, an APRM signal is input to a constant circuit 50, and a pseudo fuel surface heat flux is calculated in the APRM in consideration of a delay corresponding to a heat transfer time constant at a fuel rod. Then, the pseudo fuel surface heat flux is input to the adding circuit 51. The main steam flow rate is also input to the addition circuit 51, where the ratio of the pseudo fuel surface heat flux to the rated value and the ratio of the measured main steam flow value to the rated value are compared. 60 is input. Core monitoring device 6
0 is the pseudo surface heat flux together with the deviation signal, and the core inlet flow rate are input as input signals,
Determination unit 60a, information storage unit 60b, and display unit 60c
It is composed of The information storage unit 60b stores the operation area and the monitoring enhancement area (area 1 and area 2) of the monitoring target plant as shown in FIG. 2, an allowable deviation value (5% in area 1 and 10% in area 2), And an allowable time in a state of deviating from the allowable deviation in the monitoring enhancement area (10 in the area 1).
Seconds and 20 seconds in the area 2) are input in advance.

At the end of the operating cycle of the core, when the feedwater heating loss event occurs at a flow rate of 105% of rated output (point B in FIG. 12), the ratio (%) of the pseudo fuel surface heat flux to the rated value ) And the ratio of the main steam flow rate to the rated value (%) exceeds 5%, and the operating point is as shown in FIG.
When there is for 10 seconds in the area 1 divided by the dotted line,
A reactor scram command or a selection control rod insertion command is issued. That is, the deviation of the ratio between the pseudo fuel surface heat flux and the main steam flow rate with respect to each rated value is increased by a certain time (Δt1 second) from the current time (0 second), for example, a certain time (Δt2) of the deviation calculated two seconds before. Second = (Δt2 + Δt1) −Δt
1) That is, it is detected with respect to the average value from -12 seconds to -2 seconds, and abnormal enlargement of the deviation is determined when the average value becomes 5% or more (FIG. 3). In addition, the time change rate of the deviation between the parameters is calculated every 0.2 seconds of the measurement cycle. (0.2% / 0.2 second) or more (FIG. 4). Alternatively, a reactor scram command or the like may be issued when the time change rate of the deviation exceeds a predetermined set value or when the deviation exceeds a predetermined set value.

The operating point becomes the minimum pump speed and the maximum output point (point D in FIG. 12) by reducing the recirculation pump speed to the minimum pump speed in the rated output and the rated flow rate operation state in the middle of the cycle. If the feedwater loss of heat event occurs at that time, the ratio of the pseudo fuel surface heat flux to the rated value (%) and the ratio of the main steam flow to the rated value (%)
Is greater than 10%, and the operating point is in the region 2 indicated by the dotted line in FIG. 20 for 20 seconds, the reactor scram is operated.

The main steam flow rate is directly measured, and the pseudo surface heat flux is a calculated value based on the APRM, so that both are highly reliable. For this reason, the pseudo surface heat flux and the main steam flow rate are monitored as standard parameters in the current commercial BWR, and are used as input signals for various safety personnel and control systems. By the way, when the core inlet subcooling degree increases, the pseudo surface heat flux gradually increases due to an increase in moderator density → an increase in neutron flux → an increase in the core thermal output. The main steam flow rate is smaller than the pseudo surface heat flux because the increase in the core heat output promotes steam generation, but the increase in the core sub-cooling degree decreases the core exit quality. Increases gradually.
For this reason, when normalization is performed at the initial point, a deviation in the expanding direction occurs between the pseudo surface heat flux and the main steam flow rate. Therefore, by monitoring the deviation as described above and generating a reactor scram signal or the like by expanding the deviation,
It is possible to detect a change in the core more quickly than by simply monitoring a certain parameter as in the related art, thereby improving safety.

FIG. 5 is a diagram showing a time change of a main variable when the reactor scram is operated at the end of the core operation cycle.
As the RM increases, the MCPR decreases. However, in the present invention, before the pseudo fuel surface heat flux reaches the scrum set point (115%), a large allowable deviation of the pseudo fuel surface heat flux-main steam flow rate is detected in 54 seconds, and the reactor is detected in 64 seconds. Since the scrum is activated, the amount of decrease in MCPR (Δ
MCPR) remains at 0.13.

FIG. 6 is a diagram showing a time change of a main variable when the reactor scram is operated in the middle stage of the core cycle.
Although the damping characteristics of the PRM deteriorate, the pseudo fuel surface heat flux-main steam flow allowable deviation is detected in 60 seconds before the APRM enters the oscillation state, and the reactor scram operates in 80 seconds. Therefore, the core stability reduction ratio, the zone stability reduction ratio, and the channel stability reduction ratio are less than 1.0.

The current commercial BWR is operated in a state where the turbine inlet pressure is controlled to be constant even when the output is controlled to follow the load. This can be achieved by controlling the opening degree of the steam inlet / outlet valve of the turbine by the turbine pressure control system and controlling the number of revolutions of the recirculation pump by the recirculation flow rate control system. And the operating state of the base load operation.

When the degree of subcooling at the core inlet increases during the automatic operation of the main controller, the opening of the regulator valve is controlled to increase so that the turbine inlet pressure becomes constant when the turbine steam flow rate increases, and at the same time, the rotation of the recirculation pump is controlled. The number is controlled in a decreasing direction. For this reason, the flow rate at the core inlet decreases, and the increase in the core heat output, that is, the pseudo surface heat flux is also suppressed.

On the other hand, if the core inlet subcool degree increases during the main controller manual operation, the core inlet flow rate is kept constant, so that the pseudo surface heat flux increases and the minimum critical power ratio increases. Changes (ΔMCPR) and stability (reduction ratio) are severe events.

When the core inlet flow rate is increased by increasing the recirculation pump speed from the low core flow rate / low core heat output state, the nuclear reaction is promoted due to the decrease in the void ratio in the core, and the core heat output increases. As a result, the main steam flow rate also increases, but there is a delay due to heat transfer from the fuel, so that the deviation between the core heat output and the main steam flow rate tends to temporarily increase. Since this is outside the scope of the monitoring control of the apparatus of the present invention, the operation of the scrum or the selection control rod is unnecessary.

In view of the above, the main control operation mode is correctly monitored, and as a preparatory means, the constant flow rate at the core inlet is accurately monitored by other means.
Regardless of the operation mode of the main controller, when the monitoring parameter increases due to a factor other than an increase in the core inlet subcool degree, the core inlet flow rate is increased for the purpose of avoiding the scram operation of the control device of the present invention. It is also possible to operate the monitoring and control device only in a state that can be regarded as constant.

That is, for example, the measured value at the core inlet flow rate at the time (0 second) at which the deviation between the pseudo fuel surface heat flux and the main steam flow rate exceeds the abnormality detection set value (5%) in the monitoring enhancement region 1 And that the deviation from the measured value at the core inlet flow rate before the fixed time (10 seconds) is within the set variation range (3%), or the deviation between the pseudo fuel surface heat flux and the main steam flow rate When viewed from the time when the above-mentioned monitoring parameter including the abnormal detection setting value exceeds the abnormality detection set value, the core inlet is detected at all predetermined times (10 times) in the past every measurement cycle (0.2 seconds) including the nearest measurement time. The change rate at which the flow rate change rate is set (0.2% / 0.2
By confirming that the flow rate is within 1% / sec), the scram operation can be performed only in a state where the core inlet flow rate can be considered to be constant. The latter method can be easily achieved by adopting a digital measurement system for measuring the process amount. Also, in the case of an analog measurement system, the same monitoring can be performed by fixing the data sampling period.

Another event in which the deviation between the core heat output and the main steam flow rate transiently increases is the closing of the main steam isolation valve. In this case, a scram operation is required. There is no inconsistency with the basic operation of the monitoring control device. Further, when the forced circulation BWR having the recirculation pump is continuous in the natural circulation state when all the recirculation pumps are tripped, or the natural circulation BW without the recirculation pump is provided.
During the steady operation of R, the core inlet flow rate is substantially constant. Therefore, in this case, the original function of the monitoring and control device of the present invention can be expected without monitoring the core inlet flow rate.

FIG. 7 is a diagram showing a second embodiment of the present invention. The process comprises an LPRM signal, a reactor pressure P, a core inlet flow rate W, a feedwater flow rate _WFW , and a feedwater temperature _TFW. Signal is input and radial output peaking Ψ (r)
And core power distribution Φ (z),
A process calculator 52 for calculating the reactor heat output Q and the core inlet subcool degree 1 is provided. Core monitoring device 60
In the information storage unit, the core inlet subcool degree 2 obtained using the plant design conditions for each operating point of the reactor specified by the reactor heat output and the core inlet flow rate is recorded in advance. Accordingly, the process computer 52 uses the core of the boiling water reactor to generate steam and work in the turbine system,
And the core inlet subcool degree 1 and the core inlet subcool degree 2 obtained by performing the heat balance calculation including the heat loss are input to the adder 53, and the deviation is obtained.
The deviation signal of both becomes larger than a predetermined value,
Moreover, the core monitoring is performed only when the operating point of the reactor specified by the pseudo surface heat flux and the measured value of the core inlet flow rate exceeds the predetermined operating range on the two-dimensional coordinates as in the first embodiment. A reactor scram command or a selection control rod insertion command is issued from the device 60.

The core subcool degree is one parameter for determining the plant heat balance and is calculated as an operation result using a process computer, and has high reliability. Thus, this embodiment has the same operation and effect as the first embodiment.

FIG. 8 is a view showing a third embodiment of the present invention. In the information storage unit of the core monitoring device 60, the operating point of the reactor specified by the reactor heat output and the core inlet flow rate is provided. The feedwater temperature 2 obtained by using the plant design conditions for each time is recorded in advance. Therefore, the actual feed water temperature 1 and the feed water temperature 2 whose sign is inverted are added by the adder 54,
The deviation signal is input to the core monitoring device 60, the deviation signal becomes larger than a predetermined value, and the operating point of the reactor specified by the pseudo surface heat flux and the measured core inlet flow rate is the As in the case of the first embodiment, a reactor scram command or a selection control rod insertion command is issued from the core monitoring device 60 only when a predetermined operation range on two-dimensional coordinates is exceeded. The feedwater temperature is measured by a thermometer in the operating plant, and its reliability is high. In this case, the same operation and effect as those of the first embodiment can be obtained.

FIG. 9 shows a fourth embodiment of the present invention. The information storage unit of the core monitoring device 60 stores, for each operating point of the reactor specified by the reactor heat output and the core inlet flow rate. The subcool boiling length 2 obtained using the plant design conditions is recorded in advance. On the other hand, a subcool boiling length 1 is calculated by a computer incorporating a core thermal hydraulic calculation code (program) using the given heat balance as a boundary condition, and the subcool boiling length 1 is inverted from the subcool boiling length 1. The boiling length 2 is added by the adder 55, and the deviation signal is input to the core monitoring device 60,
The deviation signal becomes larger than a predetermined value, and the operating point of the reactor specified by the pseudo surface heat flux and the measured value of the flow rate at the core inlet is defined on the two-dimensional coordinates as in the first embodiment. The reactor core monitoring device 60 issues a reactor scram command or a selective control rod insertion command only when the predetermined operating range is exceeded.

In this case, the monitoring performance including the error of the model is slightly inferior, but the calculation time of the subcool boiling length 1 is short, and almost the same effect as in the first embodiment is obtained.

In the fourth embodiment, the reactor scram command or the selection control rod insertion command is issued according to the deviation of the subcooled boiling length. However, the reactor scram command is determined according to the deviation of the core outlet quality. A command or a selection control rod insertion command may be issued.

FIG. 10 shows a fifth embodiment of the present invention. In the information storage unit of the core monitoring device 60, each operating point of the reactor specified by the reactor heat output and the core inlet flow rate is provided. The reference distance for movement of the core average axial power distribution obtained using the plant design conditions is recorded in advance.

On the other hand, the core average axial power distribution peak position moving distance is calculated by the process computer on the basis of the LPRM measurement value, and the core average axial power distribution peak position moving distance is inverted from the sign. The reference distance of the movement of the axial power distribution is added by the adder 56, the deviation signal is input to the core monitoring device 60, the deviation signal becomes larger than a predetermined value, and the pseudo surface heat flux and Only when the operating point of the reactor specified by the measured value at the core inlet flow rate exceeds the predetermined operating range on the two-dimensional coordinates as in the first embodiment, the reactor scram command Alternatively, a selection control rod insertion command is issued.

However, in this case, the peak position may not occur smoothly, so that the overall monitoring performance is slightly inferior to the previous embodiment, but is substantially the same as in each of the above embodiments. It works.

By the way, as a method for assuming that the core inlet flow rate is constant within a predetermined range, the core at the time when the monitoring parameter including the deviation between the pseudo fuel surface heat flux and the main steam flow rate exceeds the abnormality detection set value is used. A recirculation pump speed measurement value or a recirculation pump drive flow rate can be used in addition to the deviation of the inlet flow measurement value from the measurement value before a certain time. Alternatively, it may be confirmed that the recirculation pump speed change rate or the recirculation pump drive flow rate change rate is within the set change rate.

Further, as a method for assuming that the core inlet flow rate is constant within a predetermined range, there is a method in which the above-mentioned monitoring parameter including the deviation between the pseudo fuel surface heat flux and the main steam flow rate exceeds a set value for abnormality detection. Core inlet flow measurement, or
The deviation from the measured value of the recirculation pump speed or the measured value of the recirculation pump drive flow rate a fixed time ago, or the deviation from the measured value a fixed number of times before the measurement cycle is set for each measured value. Once it is confirmed that the variation is within the range, the monitoring parameter including the deviation between the pseudo fuel surface heat flux and the main steam flow exceeds the set value for abnormality detection. A method is used in which the rate of change of the core inlet flow rate, the rate of change of the recirculation pump speed, or the rate of change of the recirculation pump drive flow rate within the set change rate is checked at least twice, including the measurement time You can also.

[0055]

As described above, the present invention monitors the increasing tendency of the deviation between the parameters deeply related to the core inlet subcooling degree, and detects the tendency for the predetermined abnormality detection setting in a predetermined operation region. When the value is exceeded, the selected control rod mechanism or scram is operated, so the transient characteristics in the rated output operation state can be improved in the initial state, and the low flow rate / high output operation state can be improved. The stability of the core can be improved. In addition, natural circulation type B
In the case of a furnace having no forced recirculation pump, such as a WR, the present monitoring device can prevent both MCPR reduction and neutron flux oscillation when feedwater heating loss occurs during rated output operation.

[Brief description of the drawings]

FIG. 1 is a configuration diagram showing a first embodiment of a supervisory control device of the present invention.

FIG. 2 is a diagram showing an operation area to which the monitoring device of the present invention is applied.

FIG. 3 is an explanatory diagram for judging a deviation between monitoring parameters.

FIG. 4 is an explanatory diagram for determining a deviation between monitoring parameters.

5A and 5B are transient change diagrams when the monitoring device of the present invention is applied, wherein FIG. 5A shows a change in a ratio to a rated value, and FIG.
The figure which shows the change of (DELTA) MCPR.

6A and 6B are transient change diagrams when the monitoring device of the present invention is applied, wherein FIG. 6A shows a change in the ratio to a rated value, and FIG.
FIG. 4 is a diagram showing a change in a reduction ratio.

FIG. 7 is a configuration diagram showing a second embodiment of the monitoring control apparatus of the present invention.

FIG. 8 is a configuration diagram showing a third embodiment of the monitoring control apparatus of the present invention.

FIG. 9 is a configuration diagram showing a fourth embodiment of the monitoring control apparatus of the present invention.

FIG. 10 is a configuration diagram showing a fifth embodiment of the monitoring control apparatus of the present invention.

FIG. 11 is a configuration diagram of a boiling water reactor.

FIG. 12 is an operation characteristic diagram of a boiling water reactor.

FIG. 13 is a plan view of a core of a boiling water reactor.

FIG. 14 is a sectional view of a core of a boiling water reactor.

FIG. 15 is a transient change diagram in a conventional example, in which (a)
FIG. 7B is a diagram illustrating a change in the ratio with respect to the rated value, and FIG.

FIG. 16 is a transient change diagram in a conventional example, in which (a)
FIG. 7B is a diagram showing a change in a ratio with respect to a rated value, and FIG.

[Explanation of symbols]

 DESCRIPTION OF SYMBOLS 1 Reactor containment vessel 3 Fuel assembly 4 Control rod 6 Coolant 7 Recirculation system 50 Time constant circuit 51, 53, 54, 55, 56 Addition circuit 52 Process computer 60 Core monitoring device

Continued on the front page (51) Int.Cl. ⁷ Identification FI FI Theme Court II (Reference) G21D 3/04 G21C 17/10 GR (72) Inventor Takanori Fukahori 2-3-1 Kawakawa, Yokosuka City, Kanagawa Prefecture Inside Niuclear Fuyuel Co., Ltd. (72) Inventor Tohru Kanazawa 2-3-1 Kawaguchi, Yokosuka City, Kanagawa Prefecture Inside Niuclear Fuyuel Co., Ltd. (72) Inventor Akira Motoya 2-chome River, Yokosuka City, Kanagawa Prefecture No. 3-1 Inside Niuclear Fuyuel Co., Ltd. (72) Inventor Masatoshi Sugawara 2-3-1 Kawakawa, Yokosuka City, Kanagawa Prefecture F-term inside Niuclear Huyell Co., Ltd. 2G075 BA13 BA14 CA08 CA20 CA27 CA40 DA05 DA20 FA06 FB09 FB10 FC03 GA15 GA29

Claims (16)

[Claims] When the deviation between the monitoring parameters in a boiling water reactor becomes larger than a predetermined value, the operating point specified by the pseudo fuel surface heat flux and the core inlet flow rate is defined by two-dimensional coordinates. A monitoring and control device for a boiling water reactor, which issues a reactor scram command or a selection control rod insertion command only when it has been in a predetermined area for a predetermined time. 2. The deviation between monitored parameters in a boiling water reactor is calculated by considering a core average neutron flux measurement (APRM) in consideration of a delay corresponding to a heat transfer time constant in a fuel rod. The ratio of the heat flux to the rated value,
2. The monitoring and control device for a boiling water reactor according to claim 1, wherein the deviation is a deviation from a ratio of a measured value of a main steam flow rate to a rated value. 3. The deviation between the monitoring parameters in the boiling water reactor is calculated by using a process computer for monitoring the operation of the core to determine the steam generation by the core of the boiling water reactor, the work in the turbine system, and the heat loss. Using the plant design conditions for each reactor operating point specified by the reactor heat output and core inlet flow rate before starting power generation, 2. The monitoring and control apparatus for a boiling water reactor according to claim 1, wherein the difference is a deviation of a subcool degree at a core inlet. 4. A deviation between monitoring parameters in a boiling water reactor is specified by a measured value of a feed water temperature which is input to a core operation monitoring process computer, and a reactor heat output and a core inlet flow before starting power generation of a plant. The boiling water reactor monitoring and control apparatus according to claim 1, wherein the deviation is a feedwater temperature deviation obtained by using a plant design condition for each operating point of the reactor. 5. The deviation between monitoring parameters in a boiling water reactor is calculated by using a process operation monitoring computer for a core operation, wherein a channel having a maximum radial peaking coefficient of the core or a radial peaking coefficient and an axial peaking coefficient are used. Is the subcooled boiling length calculated for the channel with the largest product and the deviation of the subcooled boiling length obtained using the plant design conditions for each operating point of the reactor specified by the reactor heat output and the core inlet flow rate. Characterized in that there is
The monitoring and control device for a boiling water reactor according to claim 1. 6. The core outlet quality calculated by a computer incorporating a core thermal-hydraulic calculation code, using a given heat balance as a boundary condition, and a deviation between monitored parameters in the boiling water reactor are determined in advance. 2. The monitoring and control device for a boiling water reactor according to claim 1, wherein the deviation is a deviation from the calculated value. 7. A deviation between monitored parameters in a boiling water reactor is a deviation between a pseudo fuel surface heat flux and a turbine steam flow rate, or a deviation between a pseudo fuel surface heat flux and a turbine control valve opening or a pseudo fuel surface heat flux. 2. The monitoring and control device for a boiling water reactor according to claim 1, wherein the deviation is between a fuel surface heat flux and a generator output or a difference between a pseudo fuel surface heat flux and a feedwater flow rate. 8. The peak position of the core average axial power distribution calculated by the core operation monitoring process computer moves in the direction of the core inlet despite the fact that the control rod pattern is not changed. When the operating point of the reactor specified by the pseudo surface heat flux and the core inlet flow rate has been within the specified area on the two-dimensional coordinates for a certain period of time when the specified distance is exceeded, the reactor scram A monitoring and control apparatus for a boiling water reactor, which issues a command or a selection control rod insertion command. 9. A deviation between a pair of monitored parameters represented by a deviation between a pseudo fuel surface heat flux and a main steam flow rate, or
When either the deviation of the monitored parameter from the design value or the moving distance exceeds the abnormality detection set point, the measured value of the core inlet flow rate can be regarded as constant within a certain predetermined width, and A reactor scram command or a selection control rod insertion command is issued only when the operating point of the reactor specified by the surface heat flux and the core inlet flow rate exists within a predetermined area on the two-dimensional coordinates for a certain period of time. Characterized by the fact that
A monitoring and control device for a boiling water reactor according to any one of claims 1 to 8. 10. An abnormality detection set value for operating a scram or a selection control rod insertion mechanism is dependent on an operating point specified by a pseudo fuel surface heat flux and a core inlet flow rate measurement value. 10. The monitoring and control device for a boiling water reactor according to any one of 1 to 9. 11. A deviation between a pair of monitoring parameters, a deviation from a design value of a monitoring parameter, or a moving distance, the deviation or the moving distance calculated a predetermined time before the present time, for a certain time. Calculating an expansion width from the average value, and when the expansion width exceeds a set point for abnormality detection for each monitoring parameter, issues a reactor scram command or a selection control rod insertion command, A monitoring and control device for a boiling water reactor according to any one of claims 1 to 10. 12. The time change rate of the deviation is calculated for each measurement cycle, and when the time change rate exceeds a predetermined value continuously for a predetermined number of times, or when the time change rate of the deviation is 2. A reactor scram command or a selective control rod insertion command is issued when a predetermined set value is exceeded or when a deviation exceeds a predetermined set value. 12. The monitoring and control apparatus for a boiling water reactor according to any one of claims 11 to 11. 13. A method for assuming that the core inlet flow rate is constant within a predetermined range is a method in which the above-mentioned monitoring parameter including a deviation between the pseudo fuel surface heat flux and the main steam flow rate exceeds a set value for abnormality detection. The deviation from the measured value at the core inlet flow rate, the measured value of the recirculation pump speed, or the measured value of the recirculation pump drive flow rate a certain time before, or the difference between the measured value a fixed number of times before the measurement cycle, respectively The monitoring and control device for a boiling water reactor according to any one of claims 1 to 12, wherein it is confirmed that the measured value is within a variation range set for the measured value. 14. A method for assuming that the core inlet flow rate is constant within a predetermined range, assuming that the above-mentioned monitoring parameter including the deviation between the pseudo fuel surface heat flux and the main steam flow rate exceeds a set value for abnormality detection. At the time, including the closest measurement time, the core inlet flow rate change rate, or the recirculation pump speed change rate, or the recirculation pump drive flow rate change rate was set at all of the past predetermined times in each measurement cycle. The monitoring and control device for a boiling water reactor according to any one of claims 1 to 12, wherein it is confirmed that the change rate is within a change rate. 15. A method for assuming that the core inlet flow rate is constant within a predetermined range according to claim 13. The monitoring and control device for a boiling water reactor according to any one of claims 1 to 12, wherein both methods for confirming the rate twice or more are used. 16. When the core flow rate can be regarded as constant and the above-mentioned monitoring parameter including the deviation between the pseudo fuel surface heat flux and the main steam flow rate exceeds a set value for abnormality detection, the core inlet subcool degree becomes abnormal. The monitoring and control device for a boiling water reactor according to any one of claims 1 to 15, wherein the increase is output as a diagnosis result.