JP2005331290A - Operation control system of bwr plant - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide an operation control system of a boiling water reactor (BWR) plant, capable of securing thermal margin and safety margin at the transient time, while maintaining state of nuclear validity, and maintaining soundness of a reactor recirculation pump. <P>SOLUTION: A low deceleration spectrum BWR plant with a reactor core 11 made dense is equipped with a main steam system 23, a reactor water supply system 25, a feed water heating device 26, a reactor recirculation system 17 for circulating reactor water in the reactor 13, a measuring means 37 for measuring process quantities, such as heat output from the reactor 13, a reactor pressure, a main steam flow rate, a water supply flow rate, a water supply temperature, a core flow rate or a recirculation pump flow rate, a core nuclear hydrothermal force performance operation device 40 for calculating an average void fraction and the minimum output ratio in the core by using the process quantities measured by the measuring means 37, and a core nuclear hydrothermal force control device 41 for controlling operation of the feed water heating device 26 and the reactor recirculation system, based on the average void fraction and the minimum output ratio in the core acquired by operation processing by the core nuclear hydrothermal force performance operating device 40. <P>COPYRIGHT: (C)2006,JPO&NCIPI

Description

  The present invention relates to an operation control system for a BWR plant having a dense core, and in particular, a reduction in volume ratio of water as a moderator, reduction of fast neutrons generated by fission, and reduction of effective utilization of plutonium. The present invention relates to an operation control system for a fast spectrum BWR plant.

  A boiling water reactor (BWR) has a reduced speed spectrum BWR in which a fuel assembly mounted on a core is densified and a volume ratio of water (coolant) as a moderator is reduced. This reduced-speed spectrum BWR is intended to reduce the speed of fast neutrons generated by fission and to make effective use of plutonium.

  In the low-deceleration spectrum BWR, in order to reduce fast neutrons generated by fission, the core is densified compared to the conventional BWR, and the volume ratio of water (coolant) as a moderator is reduced. In order to increase the speed reduction effect of fast neutrons, reactor cores have been developed in which the void ratio of the core is increased or the moderator water (coolant) is eliminated.

  As this type of nuclear reactor core, there are those employing a void tube as described in JP-A-6-258472 and those employing a streaming channel as described in JP-A-11-23765. is there.

  The BWR adopts a void tube. The gas-filled tube filled with helium gas in the reactor core enters and exits as a void tube, and voids are created in the core to improve the core void ratio, reactor power control, burnup compensation, and reactor It is designed to stop.

  In addition, the BWR adopting the streaming channel forms a streaming path by the streaming channel above the axial direction of the reactor core, promotes neutron core leakage in the core direction when the coolant is voided, and reduces the void reactivity. The negative value is used to effectively use the reactor fuel.

Furthermore, a control system for a BWR plant that improves fuel soundness during a transient change caused by a failure of a nuclear reactor device or the like has also been proposed (Japanese Patent Laid-Open No. 9-304587).
JP-A-6-258472 Japanese Patent Laid-Open No. 11-23765 Japanese Patent Laid-Open No. 9-305487

  Unlike the general BWR core, the core of the reduced-speed spectrum BWR in which the core is densified and the volume fraction of water as the moderator is reduced is different from the core of a general BWR. The core is designed to reduce the amount of water that is a moderator and coolant.

  In this reduced speed spectrum BWR, the core flow rate of the coolant is smaller than that of a general conventional BWR, and the feed water temperature into the reactor is increased to increase the void ratio in the reactor core.

  For this reason, in the reduced-speed spectrum BWR, even if the core is established nuclearly, from the viewpoint of heat, such as the removal of heat generated in the reactor core and the safety margin at the time of transient, It is a severe condition. Furthermore, in a nuclear reactor core in which the nuclear reactor fuel is densified, the coolant flow passage area in the core is small and the void ratio is high, so the pressure loss in the core increases, and the recirculation flow rate in the reactor is increased. In order to ensure it, the design margin for the reactor recirculation system is small, which is a severe condition.

  In the low-deceleration spectrum BWR, the core average void ratio must be about 60% or more in order to achieve the reduction rate nuclear establishment (nucleation establishment condition) that reduces the fast neutrons. In a conventional BWR core, the core average void fraction is about 40%.

  In order to secure the core establishment condition of the low deceleration spectrum BWR, that is, the condition that the core average void ratio is about 60% or more, it is considered to decrease the core flow rate or raise the feed water temperature. According to the preliminary evaluation, in order to secure the nuclear feasibility (nuclear establishment condition) of the reduction speed, the core flow rate is set to about 50% of the conventional BWR, and the feed water temperature is changed from the feed water temperature 490K to 530K of the conventional BWR. It is necessary to raise it to the extent.

  On the other hand, when the core average void ratio is as high as 60% in order to secure the nuclear feasibility of the reduction speed, the thermal margin and the safety margin at the time of transition become small.

  FIGS. 8A and 8B show the nuclear establishment condition and the transient safety establishment condition of the reduced speed spectrum BWR, respectively. It can be seen that in the low deceleration spectrum BWR, the nuclear establishment condition and the transient safety establishment condition depend on the core average void ratio, the core flow rate of the critical power ratio, the feed water temperature, and the required dependency. Under the nuclear establishment condition of the low deceleration spectrum BWR, the nuclear establishment of {core average void fraction> 60%} is secured, and under the transient safety establishment condition, {the limit output ratio during operation> 1.07 + ΔMCPR (transient It is important to ensure the thermal and transient safety feasibility of the minimum limit output ratio at the time)}, and it is understood that there are limitations on the ranges of the core flow rate and the feed water temperature.

  FIG. 8A shows the core flow rate and the feed water temperature dependence of the core average void rate. The core average void rate increases as the core flow rate decreases and the feed water temperature increases. In order to ensure nuclear feasibility, if the core average void ratio is 60% or more, it is necessary to reduce the core flow rate and raise the feed water temperature.

  FIG. 8B shows the dependence of the critical power ratio during operation on the core flow rate and the feed water temperature, and increases as the core flow rate increases and the feed water temperature decreases. In order to ensure the thermal / transient safety feasibility, it is necessary to increase the core flow rate and lower the feed water temperature when the limit output ratio during operation is set to {1.07 + ΔMCPR} or more.

  Further, when the feed water temperature to the reactor is raised, the subcooling degree at the core inlet is lowered, and there is a possibility that cavitation occurs in the reactor recirculation pump.

  The present invention has been made in consideration of the above-mentioned circumstances, and while maintaining the nuclear feasibility, the thermal margin and the safety margin at the time of transient are ensured, and the soundness of the reactor recirculation pump is maintained. An object of the present invention is to provide an operation control system for a BWR plant.

  In order to solve the above-described problem, the BWR plant operation control system according to the present invention is generated in the reactor in the reduced-speed spectrum BWR plant in which the reactor core is densified as described in claim 1. A main steam system for directing steam, a reactor water supply system for guiding feed water into the reactor, a feed water heating device for heating the feed water provided in the reactor feed water system, and circulating the reactor water in the reactor Reactor recirculation system, measurement means for measuring the process amount of the reactor such as heat output, pressure in the reactor, main steam flow rate, feed water flow rate, feed water temperature, core flow rate, recirculation pump flow rate, etc. The core nuclear hydro-hydraulic performance calculator that calculates the average void fraction and minimum power ratio in the core using the process quantity measured in step 1, and the average in the core that is processed by the nuclear thermo-hydraulic performance calculator Void rate and minimum output The basis, in which a core nuclear thermal hydraulic control device for controlling the operation of the water supply heating and the reactor recirculation system.

  The operation control system of the BWR plant according to the present invention can maintain thermal integrity and safety margin during transition, and maintain the soundness of the recirculation pump while maintaining nuclear feasibility.

  An embodiment of an operation control system for a BWR plant according to the present invention will be described with reference to the accompanying drawings.

  FIG. 1 is a diagram schematically showing an operation control system of a BWR plant according to the present invention, and reference numeral 10 in the figure indicates a reduced speed spectrum BWR in which a nuclear reactor core 11 is densified. In the BWR 10, a reactor pressure vessel 13 is stored in a reactor containment vessel 12, and the reactor pressure vessel 13 is erected on a support pedestal 14 thereof.

  The reactor pressure vessel 13 contains a reactor core 11 and is covered with a core shroud 15. An annular or sleeve-shaped downcomer portion 16 is formed between the reactor pressure vessel 13 and the core shroud 15, and this downcomer portion 16 is an inter-reactor as a reactor recirculation pump that constitutes an in-reactor recirculation system 17. A null pump 18 is provided. A plurality of, for example, ten internal pumps 18 are provided along the circumferential direction of the downcomer portion 16.

  A core lower plenum 19 is formed below the reactor core 11, and the reactor water (coolant) descending the downcomer portion 16 by driving the internal pump 18 is guided to the core lower plenum 19. The reactor water led to the lower core plenum 19 is inverted and guided to the nuclear reactor core 11, absorbs heat generated by the fission reaction while passing through the core 11, and becomes a gas-liquid two-phase flow. 20 leads.

  The coolant heated in the nuclear reactor core 11 becomes a gas-liquid two-phase flow and is led to the core upper plenum 20, and this gas-liquid two-phase flow is subsequently led to the steam separator 21 for gas-liquid separation. The The liquid component separated from the gas and liquid is dropped in the reactor pressure vessel 13 and guided to the downcomer 16.

  On the other hand, the steam component separated by the steam separator 21 is dried through the steam dryer 22 to become dry superheated steam. The superheated steam is led to a steam turbine (not shown) through the main steam system 23, and the steam turbine is rotated to drive a generator (not shown) to generate power. Although there are two main steam systems 23, one system is omitted in FIG.

  The steam that has worked in the steam turbine is guided to a condenser (not shown), where it is cooled and condensed to condensate. This condensate then passes through the reactor condensate / water supply system, is returned from the reactor water supply system 25 into the reactor pressure vessel 13, and is mixed with the recirculation flow rate in the reactor pressure vessel 13. The reactor water supply system 25 is provided with a multi-stage feed water heating device 26 that heats the feed water, and the feed water heating device 26 heats the feed water to the reactor pressure vessel 13 in multiple stages.

  On the other hand, in the reduced-speed spectrum BWR 10, a reactor core 11 in which the reactor fuel is densified is provided in the reactor pressure vessel 13, and a uranium / plutonium mixed oxide fuel (hereinafter referred to as MOX fuel) 27 is loaded in the core 11. Therefore, effective utilization of plutonium is achieved.

  The low-deceleration spectrum BWR10 includes a pressure gauge 30 for detecting the reactor pressure P in the reactor pressure vessel 13, the reactor recirculation system 17 includes a flowmeter 31 for measuring the recirculation flow rate W, and the reactor. In the pressure vessel 13, a core flow meter 32 for measuring the core flow rate Wcore, a measuring device 33 such as an LPRM detector for detecting the neutron flux and measuring the thermal output Q in the reactor core 11, and the main steam The system 23 has a flow meter and a thermometer 34 for measuring the main steam flow rate Wg and the main steam temperature Tg. Further, the reactor feed water system 25 has a flow meter and a thermometer 35 for measuring the feed water flow rate Wfeed and the feed water temperature Tfeed. Are provided, and the measuring means 37 for measuring the process amount of the reduced speed spectrum BWR10 is constituted by each measuring instrument such as a pressure gauge, a flow meter, and a thermometer.

  The measuring means 37 constitutes a measuring instrument group for measuring the process amount of the BWR plant, and the thermal output (furnace output) Q, the furnace pressure P, the main steam flow rate Wg, the main steam temperature Tg, the feed water flow rate of the reduced speed spectrum BWR10. Process quantities (physical quantities) such as Wfeed, feed water temperature Tfeed, core flow rate Wcore, and recirculation flow rate W are measured.

  The measuring means 37 detects the process amount of the reduced speed spectrum BWR10, while the data of the detected process amount is input to the core nuclear thermal output performance calculation device 40 which is a process computer, and this core nuclear thermal hydraulic performance calculation In the apparatus 40, calculation processing is performed by three-dimensional nuclear thermal hydraulic calculation, and the average void ratio and the minimum critical power ratio in the reactor core 11 are calculated.

  A calculation result signal based on the average void ratio in the core and the minimum limit output ratio calculated by the core nuclear thermal hydraulic performance calculation device 40 is output to the core nuclear thermal hydraulic control device 41. The core nuclear thermal hydraulic control device 41 receives calculation result signals of the average void fraction in the core and the minimum limit output ratio, and is compared with preset nuclear / thermal set values. The core nuclear thermal hydraulic control device 41 is configured so that the internal water pump 26 and the internal recirculation system 17 of the reactor recirculation system 17 have an average core void fraction and a minimum critical power ratio that satisfy the nuclear and thermal set values. The operation of 18 is controlled.

  By controlling the operation of the reactor recirculation system 17 and the feed water heater 26, it is possible to provide an operation control system for a reduced-speed spectrum BWR plant that can secure thermal margin and safety margin during transition while maintaining nuclear feasibility. .

  Further, the core nuclear thermal hydraulic performance calculation device 40 is connected to the core nuclear thermal hydraulic performance display device 43, and the average void ratio in the core calculated by the core nuclear thermal hydraulic performance calculation device 40 and the minimum limit power ratio. Is displayed on the core nuclear thermal hydraulic performance display device 43. For this reason, since the average void ratio and the minimum limit power ratio in the core can be monitored by the core nuclear hydraulic performance display device 43, each operating state is considered from the viewpoint of nuclear feasibility, thermal / transient feasibility. The safety margin at can be evaluated.

  In FIG. 1, reference numeral 45 denotes a control rod drive mechanism, and the operation of this control rod drive mechanism 45 allows the control rod 46 to be taken in and out of the reactor core 11 so that the output operation of the reduced speed spectrum BWR 10 can be controlled. It has become. Reference numeral 48 denotes a fuel pool.

  The reactor core 11 of the low deceleration spectrum BWR 10 has a flat cross-sectional structure shown in FIG. The reactor core 11 is loaded with, for example, 132 normal fuel channels 50 and 76 streaming channels 51. Each fuel channel 50 (51) is as large as about four times the fuel channel 52 of the existing BWR. The diameter of the fuel rod 53 accommodated in each fuel channel 50 (51) is smaller than that of the fuel rod 54 accommodated in the fuel channel 52 of the existing BWR, and the water removal effect can be increased accordingly. The inner diameter of the fuel rod 53 is, for example, about 10 mmφ.

  Between the fuel channels 50 of the low-deceleration spectrum BWR10, control rods 46 having a cross-shaped cross section are provided so as to be freely inserted and removed from below, while the fuel rods 53 have a triangular cross-sectional structure in the fuel channel 50 (51). Arranged and densified. The flat cross section triangular arrangement structure is more dense than the flat cross section quadrangular arrangement structure, like the fuel rods 54 of the existing fuel channel 52, and the degree of this densification makes the diameter of the fuel rod 53 smaller than the existing fuel rod 54. By doing so, it is further densified. For the purpose of densification, a flat hexagonal array structure with seven fuel rods may be used.

  By making each fuel rod 53 accommodated in the fuel channel 50 into a dense arrangement structure, the amount of coolant (water) that is a moderator present in the reactor core 11 can be reduced. In the reactor core 11, in addition to the normal fuel channel 50, a streaming channel 51, which forms a streaming path, is disposed on the upper part, so that the amount of coolant is reduced and the coolant (water) is voided. The leakage of neutrons in the axial direction of the core can be promoted. The reactor core 11 loaded with the streaming channel 51 can reduce the void reactivity and can have a negative value. Thereby, the soundness of each fuel rod can be improved.

  Incidentally, the core nuclear thermal hydraulic performance calculation device 40 provided in the operation control system of the reduced-speed spectrum BWR plant outputs the output for each dense fuel assembly 50, 51 based on the measurement data of the process amount by the measuring means 37. The flow rate is calculated, and the core average void fraction is accurately evaluated by three-dimensional nuclear thermal hydraulic calculation.

  The core average void ratio is a ratio of the volume of steam in the coolant present in the reactor core 11 and the volume of coolant (liquid), and is generally a function of quality.

  The calculation of the core average void ratio is performed as shown in FIG. 3, and the measurement means 37 uses the process amount of each measuring instrument group as measurement data (process data). Based on the measured process data of the measuring instrument groups 30 to 35, the power and flow rate are calculated for each dense fuel assembly 50 and 51, and the core average void ratio can be obtained with high accuracy.

  On the other hand, in the operation control system of the reduced-speed spectrum BWR plant, as shown in FIG. 4, the core average void ratio is roughly determined from the heat balance of the reactor core 11 and the quality void correlation equation.

  The heat balance of the reactor core 11 is calculated as in the following equation (1). The heat output of the BWR plant is Q, the main steam flow rate (= feed water flow rate) is Wg, the main steam enthalpy obtained from the main steam temperature Tg is hg, the feed water enthalpy obtained from the feed water temperature Tfeed is hfeed, the total core flow rate Wtotal, the core inlet enthalpy Is the heat balance of the BWR plant, where h is the core exit enthalpy and hf is the saturated water enthalpy.

[Equation 1]
Q = Wg * (hg-hfeed) = Wtotal * (hout-hin)
Wtotal * hin = (Wtotal−Wg) * hf + Wg * hfeed
...... (1)
It is represented by

  From this equation (1), the core average quality Xave can be calculated.

[Equation 2]
Xave = (hout + hin) / 2 * (hg−hf) (2)
From the equation (2) representing the core average quality Xave, the core average void ratio αave can be obtained using a known quality void correlation equation (for example, refer to FIG. 1 of Japanese Patent Publication No. 61-28117).

  That is, the core nuclear thermal hydraulic performance calculation device 40 can easily improve the calculation accuracy of the core average void ratio by calculating the heat balance of the BWR plant, and can maintain the nuclear feasibility. .

  Next, the operation limit control ratio MCPR1 during operation and the limit output ratio MCPR2 during transition are calculated in the steps shown in FIG. 5 by the operation control system of the BWR plant.

  First, the core nuclear hydro-hydraulic performance calculation device 40 is installed in the reactor core 11 based on process amount data (process data) measured by the measuring instrument groups 30 to 35 of the measuring means 37. The outputs, flow rates, and limit outputs of the dense fuel assemblies 50 and 51 are sequentially measured, and the limit output ratio MCPR1 during operation of the BWR plant is calculated.

  The limit output ratio MCPR1 during operation in each operation state of the BWR plant is calculated, and abnormal transient changes in each operation state are analyzed and evaluated, and the minimum limit output ratio change (ΔMCPR) during transients is calculated. This ΔMCPR is compared with the safety limit value for operating the BWR plant, and the minimum limit output ratio MCPR2 at the time of transition is obtained by calculation.

  The core nuclear thermal hydraulic performance calculation device 40 can improve the calculation accuracy of the limit output ratio in each operation state, and can maintain a safety margin at the time of transition.

  The minimum limit output ratios MCPR1 and MCPR2 are one of parameters for evaluating the soundness of the BWR plant. The minimum limit power ratio is obtained by dividing the limit at which the coolant in contact with the fuel rod (fuel cladding) in the reactor core 11 causes film boiling by the current reactor power. It is said that no film boiling occurs in the fuel rod.

  The limit value at which 99.9% of the fuel rods do not cause film boiling is called the safety limit minimum limit output ratio SLMCPR. If this safety limit minimum limit output ratio SLMCPR is greater than 1.07, 99.9% of the fuel rods will not be damaged even if abnormal transient changes occur during operation.

  Therefore, even in the No. 6 and No. 7 reactors of the Kashiwazaki-Kariwa Nuclear Power Station, the criterion for judging abnormal transient changes during operation is set such that the minimum limit output ratio MCPR is 1.07 or more.

  Further, the core nuclear thermal hydraulic performance calculation device 40 calculates the process amount measurement data from each measuring instrument group of the measuring means 37, calculates the average void ratio and the minimum limit output ratio in the core 11, and The calculation result signal is sent to the core nuclear thermal hydraulic control device 41. In the core nuclear thermal hydraulic control device 41, the core average void ratio and the minimum limit power ratio are compared with preset nuclear / thermal set values.

  Specifically, as shown in FIG. 6, the core nuclear thermal hydraulic control device 41 determines whether or not the core average void ratio exceeds 60%. If the core average void ratio> 60%, Then, the process proceeds to the determination of the limit output ratio MCPR1 during operation.

  Next, it is determined whether or not the limit output ratio MCPR1 during operation exceeds (1.07 + ΔMCPR). If this limit output ratio MCPR1> (1.07 + ΔMCPR), the rotation of the recirculation pump (internal pump 18). Operation control of the in-reactor recirculation system 17 and the feed water heater 26 is performed so that the number of the feed water heaters 26 can be maintained. ΔMCPR indicates a change in the minimum limit output ratio at the time of transition.

  On the other hand, when the core average void ratio is ≦ 60%, the core nuclear thermal hydraulic control device 41 is configured so that the rotation speed of the recirculation pump (internal pump 18) decreases and the feed water temperature of the feed water heating device 26 rises. The operation of the in-furnace recirculation system 17 and the feed water heater 26 is controlled.

  Further, when the critical output ratio MCPR1 ≦ (1.07 + ΔMCPR) during operation, the reactor core thermal hydraulic control device 41 increases the number of revolutions of the recirculation pump and decreases the feed water temperature. 17 and the feed water heater 26 are controlled.

  The core nuclear thermal hydraulic control device 41 compares the average void fraction in the core and the minimum limit output ratio output from the core nuclear thermal hydraulic performance calculation device 40 with a predetermined nuclear / thermal set value. By controlling the operation of the feed water heating device 26 and the recirculation pump (internal pump 18) so as to satisfy the nuclear and thermal set values, the thermal feasibility is maintained while maintaining the nuclear feasibility. It is possible to provide an operation control system for a reduced-speed spectrum BWR plant that ensures a safety margin.

  On the other hand, in the operation control system of the BWR plant, the cavitation characteristics of the recirculation pump are calculated in the steps shown in FIG. Evaluate. By this pump characteristic evaluation, the operation margin of the recirculation pump, that is, the operation margin to the cavitation condition can be calculated, and the soundness of the recirculation pump can be maintained.

  When the condition for the cavitation of the internal pump 18 is reached, the core nuclear thermal hydraulic performance calculation device 40 outputs a feed water heating stop signal from the core nuclear thermal hydraulic control device 41 to the feed water heating device 26, and stops feed water heating. Thus, cavitation is prevented from occurring in the internal pump 18. Thereby, the soundness of the internal pump 18 can be maintained.

  That is, as shown in FIG. 8, the core nuclear thermal hydraulic performance calculation device 40 calculates the cavitation characteristics of the recirculation pump (internal pump 18), and sends the calculation result to the core nuclear thermal hydraulic control device 41. The controller 41 determines whether or not the effective suction head (effective NPSH) of the recirculation pump is larger than the required suction head (required NPSH). If effective NPSH> required NPSH, the recirculation pump is sound. Can be maintained.

  Further, if the core nuclear thermal hydraulic control device 41 determines that effective NPSH ≦ necessary NPSH, it is determined that cavitation may occur in the recirculation pump, and the core nuclear thermal hydraulic control device 41 The operation of the device 26 is stopped to stop heating the feed water.

  By the way, in order to prevent the occurrence of cavitation in the internal pump 18 which is a recirculation pump, it is necessary to always make the effective NPSH of the pump larger than the necessary NPSH of the internal pump 18.

  As the water temperature increases, the saturated steam pressure corresponding to the water temperature increases and the effective NPSH of the pump decreases.

  In order to prevent the occurrence of cavitation at the pump impeller inlet of the internal pump 18, even if there is a certain pressure drop inside the internal pump 18, there is a margin that does not reach the steam pressure corresponding to the water temperature at that time. There is a need. The required NPSH of the pump represents this extra pressure in terms of the head.

  On the other hand, the effective NPSH is NPSH determined at the suction port of the pump from the suction water surface and the position of the internal pump 18 as shown by the equation (3).

The pressure acting on the suction liquid surface is ps [Pa], the liquid density is ρ [kg / m ³ ], the acceleration of gravity is g (= 9.80 m / s ² ), and the suction height is hs [m]. If the loss head of the suction piping system is his [Pa] and the saturated vapor pressure is pv [Pa], the difference between the total liquid pressure and the saturated vapor pressure at the pump inlet is the effective NPSH, and this effective NPSH is the cavitation. It is the margin of the pump for generation.

[Equation 3]
Effective NPSH = {ps / ρg + hs−his} −pv / ρg [m] (3)
On the other hand, if NPSH is reduced while keeping the rotation speed and flow rate of a certain pump constant, cavitation starts to occur below a certain value, and if it is further lowered, the cavitation generation range is expanded. The limit value of the occurrence of cavitation is expressed by required NPSH (Required NPSH).

  In one embodiment of the present invention, the reduced speed spectrum BWR is provided with an internal pump as a recirculation pump of the reactor recirculation system, and an example of the in-reactor recirculation system is shown. It can also be applied to an extra-reactor recirculation system provided outside the reactor pressure vessel.

Schematic which shows one Embodiment of the operation control system of the BWR plant which concerns on this invention. (A) is a plan sectional view showing an example of a reactor core having a reduced speed spectrum BWR, and (B) is a densified reactor core and a fuel assembly (fuel channel) loaded in an existing BWR reactor core. FIG. The figure which shows the core average void ratio calculation example of the core nuclear thermal-hydraulic performance calculating apparatus in the operation control system of the BWR plant which concerns on this invention. The figure which shows the example which performs the core average void ratio calculation of the core nuclear thermal-hydraulic performance calculating apparatus in the operation control system of the BWR plant which concerns on this invention using heat balance. The figure which shows the example of calculation of the limit output ratio at the time of operation | movement of the core nuclear thermal-hydraulic performance calculating apparatus in the operation control system of the BWR plant which concerns on this invention, and a transient. The figure which shows the example of operation control by the core nuclear thermal hydraulic control apparatus in the operation control system of the BWR plant which concerns on this invention. The figure which shows the example which performs the soundness evaluation of a recirculation pump with the core nuclear thermal-hydraulic performance calculating apparatus in the operation control system of the BWR plant which concerns on this invention. The figure which shows the operation control example of a feed water heating apparatus in order to maintain the soundness of a recirculation pump in the core nuclear thermal-hydraulic performance calculating apparatus in the operation control system of the BWR plant which concerns on this invention. (A) And (B) is a figure which respectively shows the nuclear heat establishment conditions and transient safety establishment conditions of the reduction speed spectrum BWR.

Explanation of symbols

10 Low deceleration spectrum BWR
11 Reactor Core 12 Reactor Containment Vessel 13 Reactor Pressure Vessel 14 Reactor Pressure Vessel Support Pedestal 15 Core Shroud 16 Downcomer 17 Reactor Recirculation System 18 Internal Pump 19 Core Lower Plenum 20 Core Upper Plenum 21 Air / Water Separation Equipment 22 Steam dryer 23 Main steam system 25 Reactor feed water system 26 Feed water heating device 27 MOX fuel 29 Reactor recirculation system 30 Pressure gauge 31 Flow meter 32 Core flow meter 33 Measuring instruments 34, 35 Flow meter and transient meter 37 Measuring means 40 Core nuclear thermal hydraulic performance calculation device 41 Core nuclear thermal hydraulic control device 43 Core nuclear thermal hydraulic performance display device 45 Control rod drive mechanism 46 Control rod 48 Fuel pool 50, 52 Fuel channel (fuel assembly)
51 Streaming channel 53, 54 Fuel rod

Claims (8)

In a reduced speed spectrum BWR plant with a densified reactor core,
A main steam system that directs steam generated in the reactor;
A reactor water supply system for guiding the water supply into the reactor,
A feed water heating device provided in the reactor water supply system for heating the feed water;
A reactor recirculation system for circulating reactor water in the reactor;
Measuring means for measuring the process amount of the nuclear reactor, such as heat output, reactor pressure, main steam flow rate, feed water flow rate, feed water temperature, core flow rate, recirculation pump flow rate,
A core nuclear hydro-hydraulic performance calculation device that calculates an average void fraction and a minimum power ratio in the core using the process amount measured by the measuring means;
A core nuclear thermal hydraulic control device that controls the operation of the feed water heating device and the reactor recirculation system based on the average void ratio in the core and the minimum output ratio that are processed by the nuclear thermal hydraulic performance calculation device. An operation control system for a BWR plant, comprising: The core nuclear thermal hydraulic performance calculation device simulates the fuel assembly loaded in the reactor core by inputting the measurement data of the process amount measured by the measuring means, and uses the three-dimensional nuclear heat calculation formula to 2. The operation control system for a BWR plant according to claim 1, wherein an average void ratio is calculated. 2. The operation control system for a BWR plant according to claim 1, wherein the core nuclear thermal hydraulic performance calculation device obtains a core average void ratio by calculation based on a heat balance of the BWR plant. 3. The core nuclear thermal hydraulic performance calculation device calculates the operating limit output in each operating state based on the measurement data of the process amount measured by the measuring means, while calculating the transient change during operation in each operating state. 2. The operation control system for a BWR plant according to claim 1, wherein a minimum limit output change at the time of transient is calculated and set so as to calculate a transient minimum limit output ratio by comparison with an operational safety limit value. . The core nuclear thermal hydraulic control device compares the average void fraction and the minimum limit power ratio in the core from the core nuclear thermal hydraulic performance calculation device with a preset nuclear thermal set value, and 2. The operation control system for a BWR plant according to claim 1, wherein the operation control of the feed water heating device and the reactor recirculation system is performed so as to satisfy a thermal setting value. The core nuclear thermal hydraulic performance calculation device is set to perform a characteristic evaluation of a recirculation pump constituting a nuclear reactor recirculation system and to calculate a cavitation characteristic of the recirculation pump by calculation. The operation control system for a BWR plant according to claim 1. The core nuclear thermal hydraulic performance calculation device determines whether or not the recirculation pump of the reactor recirculation system has reached a cavitation condition, and when the cavitation generation condition is reached, 7. The operation control system for a BWR plant according to claim 6, wherein a stop signal is outputted to stop feed water heating by the feed water heating device to prevent cavitation from occurring in the recirculation pump. The core nuclear thermal hydraulic performance calculation device outputs signals of the average void ratio and the minimum critical power ratio in the core to the core thermal hydraulic performance display device, while the core thermal hydraulic performance display device The operation control system for a BWR plant according to claim 1, wherein the void ratio and the minimum limit output ratio are displayed and can be monitored.

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