Floating reactor power control device and operation method

Abstract

To stably operate a floating type nuclear reactor.SOLUTION: An output control device (plant control device 17) of a floating type nuclear reactor comprises: a measurement unit (swing inclination sensor 9) which measures a swing inclination amount of the floating type nuclear reactor; and an output control unit 23 which adjusts the output of the floating type nuclear reactor according to the swing inclination amount of the floating type nuclear reactor measured by the measurement unit. The output control unit may comprise: first evaluation means 41a which calculates a static pressure difference between the inner side and the outer side of a core, calculates a change amount in the pressure loss of the two-phase flow and the static pressure difference on the basis of the swing inclination amount of the floating type nuclear reactor measured by the measurement unit, and evaluates a change amount in the pressure loss characteristics of a core recirculation flow path; first adjustment means 42a which adjusts a core recirculation pump rotational speed on the basis of the change amount in the core recirculation flow path pressure loss predicted by the first evaluation means; second evaluation means 41b which evaluates reactivity insertion not depending on a core flow rate change amount according to the swing inclination amount of the floating type nuclear reactor; and second adjustment means 42b which adjusts a reactivity insertion command on the basis of the reactivity insertion.SELECTED DRAWING: Figure 1

Description

[Technical Field]

[0001] The present invention relates to a power control device and an operation method for a floating nuclear reactor. [Background technology]

[0002] In order to operate nuclear power plants stably and obtain a desired generator output, for example, there is a configuration described in Patent Document 1. The nuclear power plant described in Patent Document 1 includes an extraction steam control device that sets the amount of steam extraction based on a first generator output deviation signal that is the difference between the generator output target value and the generator actual output, a corrected generator output obtained by subtracting from the generator actual output the contribution to the generator output due to control of the extraction steam rate by the extraction steam control device, an output control device that outputs a second generator output deviation signal that is the difference between the generator output target value and the corrected generator output, a recirculation flow rate control device that adjusts the recirculation flow rate of the recirculation pump in accordance with the second generator output deviation signal from the output control device, and a control rod control device that changes the position of the control rod in accordance with the second generator output deviation signal.

[0003] Furthermore, nuclear power plants equipped with floating reactors are being developed (see, for example, Patent Document 2). The nuclear power plant described in Patent Document 2 comprises a plurality of underwater floats with a plurality of propellers mounted on the bottom that can rotate independently in all directions, a semi-submersible floating structure with a plurality of columns and a platform, or a barge-type floating structure with a barge with a similar plurality of propellers mounted on the bottom, a nuclear power generation facility placed on the box-type platform or barge and connected to onshore facilities by cables, and an automatic positioning system. [Prior art documents] [Patent documents]

[0004] [Patent Document 1] Patent No. 6608324 [Patent Document 2] Patent No. 4324640 Summary of the Invention [Problem to be solved by the invention]

[0005] The nuclear power plant described in Patent Document 1 is installed on land and does not have a floating reactor. Therefore, the nuclear power plant described in Patent Document 1 does not consider the rocking and tilting that occurs in a floating reactor and the impact of this rocking and tilting on the nuclear power plant.

[0006] For example, the rocking of a floating nuclear reactor generates vibrations due to inertial forces, and the tilt reduces gravity in the axial direction of the core. This phenomenon affects the fluid, and in the case of a boiling water reactor, changes the core steam volume fraction and steam volume fraction distribution, core pressure loss, boiling characteristics, etc., causing changes or fluctuations in core power. Furthermore, rocking and tilting may change the critical power at which boiling transition occurs. The nuclear power plant described in Patent Document 1 does not address changes and vibrations in core power due to rocking and tilting, and does not consider the effects of rocking and tilting on critical power, so there is a possibility that stable operation may not be possible in a floating nuclear reactor.

[0007] On the other hand, the nuclear power plant described in Patent Document 2 is equipped with a floating reactor. Patent Document 2 describes a configuration for mitigating the rocking and tilting of the floating reactor itself, but does not describe a power control device or an operating method that mitigates the effects caused in the reactor by rocking and tilting when the floating reactor rocks and tilts. Therefore, there is a demand for a power control device and an operating method that mitigates the effects caused by rocking and tilting when the floating reactor rocks and tilts, and operates the floating reactor stably.

[0008] In order to achieve the above object, the present invention provides a power control device for a floating nuclear reactor, comprising: a measuring unit that measures the amount of tilt of the floating nuclear reactor; and a power control unit that adjusts the power output of the floating nuclear reactor in accordance with the amount of tilt of the floating nuclear reactor measured by the measuring unit. The power control unit has a first evaluation means for calculating a static pressure difference between the inside and outside of the core, calculating a change in the static pressure difference based on the amount of tilt of the floating reactor measured by the measurement unit, and evaluating a change in the pressure loss characteristic of the core recirculation flow path, and a first adjustment means for adjusting a recirculation pump rotation speed based on the pressure loss characteristic of the core recirculation flow path evaluated by the first evaluation means and a core flow rate adjustment command generated by the power control unit. The configuration will be as follows. Other means will be described later. [Means for solving the problem]

[0009] In order to achieve the above-mentioned object, the present invention provides a power control device for a floating reactor, which is configured to include a measurement unit that measures the amount of tilt of the floating reactor, and a power control unit that adjusts the power output of the floating reactor in accordance with the amount of tilt of the floating reactor measured by the measurement unit. Other means will be described later. [Effects of the Invention]

[0010] According to the present invention, the floating nuclear reactor can be operated stably. [Brief explanation of the drawings]

[0011] [Figure 1] 1 is an overall configuration diagram of a nuclear power plant according to a first embodiment. [Figure 2] FIG. 2 is an explanatory diagram of the operation of the output control device according to the first embodiment. [Figure 3] FIG. 1 is a diagram illustrating a state of a nuclear power plant according to a first embodiment. [Figure 4] FIG. 1 is an overall configuration diagram of a nuclear power plant according to a second embodiment. [Figure 5] FIG. 10 is an explanatory diagram of the operation of the output control device according to the second embodiment. [Figure 6] FIG. 10 is an overall configuration diagram of a nuclear power plant according to a third embodiment. [Figure 7] FIG. 10 is an explanatory diagram of the operation of the output control device according to the third embodiment. [Figure 8] FIG. 10 is an overall configuration diagram of a modified nuclear power plant. DETAILED DESCRIPTION OF THE INVENTION

[0012] Hereinafter, an embodiment of the present invention (hereinafter referred to as "the present embodiment") will be described in detail with reference to the drawings. Each drawing is merely a schematic illustration to allow a sufficient understanding of the present invention. Therefore, the present invention is not limited to the illustrated examples. Furthermore, in each drawing, common or similar components are given the same reference numerals, and redundant explanations thereof will be omitted.

[0013] [First embodiment] <Configuration of a nuclear power plant> The configuration of a nuclear power plant 100 according to the first embodiment will be described below with reference to Fig. 1. Fig. 1 is an overall configuration diagram of the nuclear power plant 100 according to the first embodiment. The nuclear power plant 100 according to the first embodiment includes a nuclear reactor 1 (described later) configured as a floating nuclear reactor, a rocking inclination sensor 9 (described later), and an output control unit 23 (described later).

[0014] As shown in FIG. 1, the nuclear power plant 100 according to the first embodiment includes a floating platform 31 for floating the reactor 1 on water (water surface), a plant control device 17 for controlling the operation of the nuclear power plant 100, and an instruction device 50 for inputting various instructions to the plant control device 17.

[0015] The floating platform 31 is equipped with a reactor 1, a recirculation pump 7, a steam control valve 10, a turbine 11, a generator 12, a condenser 14, a feedwater pump 15, a bleed valve 16, a control room 24, and other equipment. The control room 24 is equipped with a device for manual adjustment.

[0016] The plant control device 17 has a recirculation pump control device 18 , a control rod control device 19 , a pressure control device 20 , a steam extraction control device 21 , a feedwater flow rate control device 22 , and an output control unit 23 .

[0017] The recirculation pump control device 18 outputs a flow rate control command to the recirculation pump 7 to control the recirculation pump 7, thereby controlling the recirculation flow rate of the reactor 1. The recirculation pump control device 18 receives an input of a pump rotation speed signal from a pump rotation speed setter 25 of the power control unit 23.

[0018] The control rod control device 19 outputs insertion or withdrawal commands to the control rods 8, and controls the insertion positions of the control rods 8 arranged inside the reactor 1. The control rod control device 19 receives position signals from the control rod position setter 26 of the power control unit 23 and a power control device (not shown).

[0019] The pressure control device 20 outputs an opening command to the steam control valve 10 to adjust the pressure of the reactor 1 to a set value. The pressure control device 20 receives a steam control valve opening signal from a control valve opening setter 27 of the power control unit 23.

[0020] The extraction steam control device 21 outputs an opening command to the extraction steam valve 16 to adjust the temperature of the feedwater flowing to the reactor 1 to a set value. The extraction steam control device 21 inputs an extraction steam amount signal from the extraction steam amount setter 28 of the power control unit 23.

[0021] The feedwater flow rate control device 22 outputs a flow rate command to the feedwater pump 15 to adjust the water level of the reactor 1 to a set value. The feedwater flow rate control device 22 receives a feedwater flow rate signal from the feedwater flow rate setting device 29 of the power control unit 23.

[0022] The power control unit 23 controls the power output of the reactor core 2. The configuration of the power control unit 23 will be described later.

[0023] The plant control device 17 is connected to the reactor core power monitor 3 , the pressure sensor 4 , the flow rate sensor 5 , the water level sensor 6 , the rocking inclination sensor 9 , and the generator output sensor 13 . The core power monitor 3 detects the rate of the chain reaction in the core 2 . The pressure sensor 4 detects the pressure in the reactor 1 . The flow rate sensor 5 detects the flow rate of the cooling water in the reactor core 2 . The water level sensor 6 detects the level of the cooling water in the reactor 1 . The rocking and tilting sensor 9 detects the rocking and tilting of the reactor 1 . The generator output sensor 13 detects the output of the generator 12 .

[0024] Among these, the sway tilt sensor 9 functions as a measurement unit that measures the amount of sway tilt of the reactor 1 (floating reactor). Here, the "sway tilt amount" will be explained as meaning both the sway of the reactor 1 (the acceleration of the vertical vibration) and the amount of tilt of the reactor 1 (the tilt angle of the vertical axis of the reactor 1 relative to the water surface). The sway tilt sensor 9 has a sway sensor that detects the sway of the reactor 1 (the acceleration of the vertical vibration and the acceleration of the horizontal vibration) and an inclination sensor that detects the amount of tilt of the reactor 1 (the tilt angle of the vertical axis of the reactor 1 relative to the water surface). The sway sensor and the inclination sensor may be configured separately or integrated into one. The sway tilt sensor 9 measures the acceleration due to the sway and the tilt of the reactor 1, linearly converts the acceleration due to the inertia and gravity due to the sway into a coordinate system linked to the reactor 1, and outputs the acceleration in the axial direction of the reactor 1 and the plane direction perpendicular to the axis.

[0025] The output control unit 23 has a pump rotation speed setter 25, a control rod position setter 26, a regulator valve opening setter 27, an extraction amount setter 28, a feedwater flow rate setter 29, a first evaluation means 41a, a first adjustment means 42a, a second evaluation means 41b, a second adjustment means 42b, a third evaluation means 41c, and a third adjustment means 42c.

[0026] The pump rotation speed setting device 25 controls the recirculation flow rate of the reactor 1 by setting the rotation speed of the recirculation pump 7 . The control rod position setting device 26 sets the insertion position of the control rod 8 . The regulator valve opening setter 27 adjusts the pressure in the reactor 1 to a set value by setting the opening of the steam regulator valve 10 . The extraction amount setting device 28 adjusts the temperature of the feedwater flowing to the reactor 1 to a set value by setting the opening of the extraction valve 16 . The feedwater flow rate setting device 29 adjusts the water level of the reactor 1 to a set value by setting the flow rate of the feedwater from the feedwater pump 15 .

[0027] The first evaluation means 41a evaluates the amount of change in the pressure loss characteristics of the core recirculation channel. The first evaluation means 41a calculates the static pressure difference between the inside and outside of the core (core shroud, not shown), and calculates the amount of change in the static pressure difference and the amount of change in the two-phase flow pressure loss based on the rocking tilt of the reactor 1 measured by the rocking tilt sensor 9 (measurement unit), thereby evaluating the amount of change in the pressure loss characteristics of the core recirculation channel.

[0028] In order to mitigate the effect on the recirculation flow rate due to the rocking and tilting of the reactor 1, the first adjustment means 42a evaluates the head command of the recirculation pump 7 based on the change in the pressure loss characteristics of the core recirculation flow path predicted by the first evaluation means and the core flow rate command, and adjusts the rotation speed of the recirculation pump 7.

[0029] The second evaluation means 41b is a means for evaluating the amount of change in reactivity due to fluctuation and tilt in the core, and for evaluating the amount of change in reactivity. The second evaluation means 41b evaluates the amount of change in reactivity due to changes in various parameters including the boiling characteristics, steam volume fraction, and steam volume fraction distribution in the core fuel, based on the amount of fluctuation tilt of the reactor 1 measured by the swing tilt sensor 9 (measurement unit), and predicts the amount of change in reactivity.

[0030] The second adjustment means 42b is a means for adjusting a command (reactivity insertion command) for changing the reactivity in the core. The second adjustment means 42b adjusts the reactivity change amount command based on the amount of change in reactivity evaluated by the second evaluation means 41b.

[0031] The third evaluation means 41c is a means for evaluating the amount of change in the critical power at which boiling transition occurs and predicting the margin for the critical power based on the amount of rocking tilt of the reactor 1 measured by the rocking tilt sensor 9 (measurement unit).

[0032] The third adjustment means 42c is a means for suppressing the output when the margin of the critical output predicted by the third evaluation means 41c becomes smaller than a preset threshold value.

[0033] A core 2 is installed inside the nuclear reactor 1, and further, the core 2 contains fuel (core fuel), control rods 8 that control the fuel chain reaction, and coolant (cooling water) used as a moderator to maintain the chain reaction. The nuclear power plant 100 generates steam by boiling the coolant while maintaining the fuel chain reaction. As steam is generated, the density of the moderator decreases, the chain reaction rate decreases, and the heat generation and evaporation rate also decrease, establishing a balance between the steam volume fraction and the reactivity through a feedback process. In addition, a recirculation pump 7 that recirculates the coolant is installed around the nuclear reactor 1. The nuclear power plant 100 adjusts the recirculation flow rate of the cooling water using the recirculation pump 7 to control the steam volume fraction and the reactivity resulting from the steam volume fraction, and by setting the insertion position of the control rods 8 inside the nuclear reactor 1, adjusts the power output of the core 2 and controls the amount of steam generated (produced) in the nuclear reactor 1.

[0034] Steam generated (created) in the reactor 1 is led to the turbine 11 through a steam pipe and a steam control valve 10 provided on the steam pipe. The steam control valve 10 adjusts the flow rate of steam flowing from the reactor 1 to the turbine 11. A generator 12 is connected to the turbine 11, and the generator 12 generates electricity, converting the energy of the steam into electricity. A generator output sensor 13 is provided near the generator 12, and the generator output sensor 13 detects the generator output.

[0035] After driving the turbine 11, the steam is condensed by heat exchange in a heat exchanger (not shown) in the condenser 14 and returns to cooling water. The cooling water is pressurized by a feedwater pump 15 to a pressure similar to that inside the reactor 1, heated by a feedwater heater (not shown), and then supplied to the reactor 1 together with turbine extraction steam that passes through an extraction valve 16. The feedwater pump 15 (or a valve (not shown)) adjusts the feedwater flow rate. The extraction valve 16 adjusts the feedwater temperature.

[0036] <Nuclear power plant operation> The operation of the nuclear power plant 100 will be described below with reference to Figures 2 and 3. Figure 2 is an explanatory diagram of the operation of the output control unit 23 of the nuclear power plant 100. Figure 3 is an explanatory diagram of the state of the nuclear power plant 100. Since the nuclear power plant 100 is floated on the water surface, in this embodiment, even if the reactor 1 sways and tilts, the rotation speed of the recirculation pump 7 is controlled particularly by the pump rotation speed setter 25 to adjust the flow rate of cooling water into the reactor core, thereby mitigating the effects of the swaying and tilting and allowing the reactor 1 to operate stably.

[0037] As shown in FIG. 2, in the nuclear power plant 100, the output control unit 23 receives the generator output set value I50 of the generator 12 from the indicating device 50, the generator output I13 of the generator 12 from the generator output sensor 13, and the core pressure I4, which is the pressure of the core, from the pressure sensor 4.

[0038] The output control unit 23 starts adjusting the output according to the pressure based on the difference between the generator output set value I50 and the generator output I13 and the core pressure I4 so that the difference approaches zero (step S10).Then, the output control unit 23 performs PID control (proportional-integral-derivative control) (step S15) and generates a core power command for adjusting the core power (step S20).

[0039] In addition, the power control unit 23 receives input of the core power I3 from the core power monitor 3, input of the core flow rate I5, which is the flow rate of the cooling water flowing through the core, from the flow rate sensor 5, input of the water level I6 of the cooling water in the core from the water level sensor 6, and input of the rocking tilt amount I9 of the floating platform 31 from the rocking tilt sensor 9.

[0040] The power control unit 23 generates a reactivity insertion command based on the difference between the core power command generated in step S20 and the core power I3 (step S25). Next, the power control unit 23 starts adjusting the reactivity insertion command according to the swing tilt amount of the floating platform 31 based on the reactivity insertion command generated in step S25 and the swing tilt amount I9 (step S30). Then, the power control unit 23 generates a core flow rate adjustment command to adjust the reactivity (step S35).

[0041] The power control unit 23 also calculates the static pressure difference between the inside and outside of the core (core shroud, not shown) based on the core flow rate I5, water level I6, and core power I3 (step S40), and evaluates the amount of change in the static pressure difference according to the swing ramp I9 (step S42). The power control unit 23 then evaluates the amount of change in the pressure loss characteristics of the core recirculation channel based on the amount of change in the static pressure difference, the core flow rate I5, the swing ramp I9, and the core power I3 (step S44). The static pressure difference between the inside and outside of the core in step S40, the amount of change in the static pressure difference according to the swing ramp I9 in step S42, and the amount of change in the pressure loss characteristics of the core recirculation channel in step S44 have the relationship shown in FIG. 3. In FIG. 3, the horizontal axis represents the core flow rate, and the vertical axis represents pressure amounts representing pressure loss and pump head, showing changes in the state of the nuclear power plant 100. In FIG. 3, the solid line indicates the pressure loss characteristics before the change, and the dashed line indicates the pressure loss characteristics after the change due to the oscillation and tilt.

[0042] In addition, the power control unit 23 evaluates the head of the recirculation pump 7 based on the core flow rate I5, the core flow rate adjustment command generated in step S35, and the change in the pressure loss characteristics of the core recirculation flow path evaluated in step S44 (step S48).

[0043] Next, the power control unit 23 starts adjusting the rotation speed of the recirculation pump 7 based on the head of the recirculation pump 7 evaluated in step S48 and the core flow rate adjustment command generated in step S35 (step S50). Then, the power control unit 23 generates a recirculation pump rotation speed change command using the pump rotation speed setter 25 (see FIG. 1) (step S55). The power control unit 23 outputs the recirculation pump rotation speed change command generated in step S55 to the recirculation pump control device 18 (see FIG. 1) to cause the recirculation pump control device 18 (see FIG. 1) to adjust the rotation speed of the recirculation pump 7. As a result, as shown in FIG. 3, the state of the nuclear power plant 100 changes from a state point where the pressure loss characteristic indicated by the solid line before the change intersects with the pump characteristic to a state command point where the pressure loss characteristic indicated by the dashed line after the change intersects with the pump characteristic after the rotation speed change.

[0044] Here, we will provide additional information about the operation of the power control unit 23 shown in Figure 2. The power control unit 23 receives signals from the core power monitor 3, pressure sensor 4, flow rate sensor 5, water level sensor 6, rocking inclination sensor 9, and generator output sensor 13. The power control unit 23 also receives a generator output setting value from the control room 24.

[0045] The power control unit 23 calculates a generator output change command by subtracting the generator output set value from the generator output input from the generator output sensor 13. The power control unit 23 adjusts the pressure input from the pressure sensor 4, performs PID control (proportional integral differential control), and calculates a core power command. The power control unit 23 calculates a reactivity insertion command from the difference between the core power command and the core power input from the core power monitor 3. The power control unit 23 creates a table of the relationship between the amount of change in the axial direction and the planar direction perpendicular to the axis of the reactor 1 input from the swing tilt sensor 9, and the amount of change in tilt and the amount of change in reactivity under the condition that the core flow rate does not change, adjusts the reactivity insertion command, calculates a command for changing the steam volume fraction from this reactivity insertion command, and calculates a command for changing the core flow rate based on the core flow rate input from the flow sensor 5 and the core power input from the core power monitor 3. The power control unit 23 calculates the amount of change in the static pressure difference between the outside and inside of the core (core shroud, not shown) using the core power input from the core power monitor 3, the core flow rate input from the flow sensor 5, the water level of the reactor 1 input from the water level sensor 6, and the axial acceleration of the core 2 (reactor 1) input from the swing inclination sensor 9. The power control unit 23 calculates the amount of change in the pressure loss characteristics of the recirculation flow path from the amount of change in the static pressure difference, the core flow rate input from the flow sensor 5, the core power input from the core power monitor 3, and the acceleration and inclination in the axial direction and the plane direction perpendicular to the axis of the reactor 1 input from the swing inclination sensor 9. The power control unit 23 stores recirculation pump characteristic data in a table, calculates the recirculation pump head from the core flow rate change command and the amount of change in the pressure loss characteristics of the recirculation flow path, and calculates a recirculation pump rotation speed change command. The power control unit 23 outputs the recirculation pump rotation speed change command.

[0046] The control rod position setter 26 outputs a position signal to the control rod control device 19 and receives signals from the control room 24 and an operation monitoring function not shown in FIG.

[0047] The control valve position setter 27 outputs a steam control valve position signal to the pressure control device 20, and adjusts the pressure of the reactor 1 to a set value by inputting the pressure of the core 2 input from the pressure sensor 4, the generator output input from the generator output sensor 13, the main steam pipe flow rate input from a main steam pipe flow rate sensor (not shown), and the main steam pipe pressure input from a main steam pipe pressure sensor (not shown). At that time, the power control unit 23 calculates the steam control valve flow rate and the steam control valve position.

[0048] The extraction air volume setting device 28 outputs an extraction air volume signal to the extraction air control device 21, inputs the feedwater flow rate of the feedwater pump 15, and adjusts the feedwater temperature to a set value. At this time, the output control unit 23 calculates the extraction air volume.

[0049] The feedwater flow rate setting device 29 outputs a feedwater flow rate signal to the feedwater flow rate control device 22, inputs the water level of the reactor 1 measured by the water level sensor, the main steam flow rate, and the feedwater flow rate, and adjusts the water level of the reactor 1 to the set value.In addition, it predicts the amount of change in the water level of the core 2 due to a mismatch between the feedwater flow rate and the main steam flow rate, and adjusts the feedwater flow rate command.

[0050] Next, the operation monitoring function will be explained. The boundaries of the operation permission region (region boundaries) are tabulated in an operation map of core flow rate and core power, and the effects of swing and tilt on the operation permission region boundaries are also tabulated. The speed and tilt due to swing input from the swing tilt sensor 9 are input to calculate the operating region boundaries during normal operation. The core power input from the core power monitor 3 and the core flow rate input from the flow sensor 5 are compared with the calculated operating region boundaries (operating region boundaries during normal operation), and if the core flow rate exceeds the operating region boundaries, a signal is output to the control rod positioner 26.

[0051] As described above, according to the first embodiment, stable operation of the boiling water reactor as a floating reactor is possible by measuring the sway and tilt and adjusting the output power control unit 23 based on the measurements. In addition, appropriate operation monitoring of the boiling water reactor as a floating reactor is possible by adjusting the operating range based on the measured values ​​of the sway and tilt.

[0052] In the nuclear power plant 100, vertical and horizontal accelerations are applied to the floating platform 31 as the floating platform 31 moves and tilts in the vertical and horizontal directions. This causes changes in the steam volume fraction of the core 2 and the pressure loss in the recirculation flow path, resulting in changes in the flow rate of the cooling water and the output of the core 1. Therefore, in this embodiment, in step S55, the pump rotation speed setter 25 (see FIG. 1) of the output control unit 23 generates a pump rotation speed change command for changing the rotation speed of the recirculation pump 7 so as to reduce the change in output, and outputs the command to the recirculation pump control device 18, thereby adjusting the rotation speed of the recirculation pump 7.

[0053] The nuclear power plant 100 according to the first embodiment uses a floating reactor power control device to mitigate the effects of swaying and tilting, enabling stable and safe operation of the reactor even during swaying and tilting.

[0054] (1) The nuclear power plant 100 according to the first embodiment adjusts the output of the reactor 1 by adjusting the rotation speed of the recirculation pump 7. In the case of a forced circulation boiling water reactor, the nuclear power plant 100 according to the first embodiment adjusts the reactivity insertion command and the rotation speed change command of the recirculation pump based on the measured sway and tilt, thereby mitigating the impact of sway and tilt on the core output and core flow rate. (2) The nuclear power plant 100 according to the first embodiment adjusts the boundary of the operation permission region (region boundary) to enable safe operation, taking into consideration changes in the critical power at which boiling transition occurs due to rocking and tilting of the reactor 1. Note that boiling transition is the transition of the boiling state from annular flow to atomized flow, and at this time the liquid film on the fuel rod surface disappears, causing a sudden rise in the fuel rod surface temperature. (3) The nuclear power plant 100 according to the first embodiment can improve the stability of the reactor power control device by predicting the fluctuations and tilt of the reactor 1 from measurement data of the fluctuations and tilt of the reactor 1.

[0055] The nuclear power plant 100 according to the first embodiment can stably operate the reactor 1, which is a floating reactor.

[0056] [Second embodiment] The configuration of a nuclear power plant 100A according to the second embodiment will be described below with reference to Fig. 4. Fig. 4 is an overall configuration diagram of the nuclear power plant 100A according to the second embodiment. Compared to the nuclear power plant 100 according to the first embodiment, the nuclear power plant 100A according to the second embodiment does not use a recirculation pump 7 but provides a chimney 30 to obtain a desired core flow rate, and the operation method of the power control unit 23A is different.

[0057] As shown in FIG. 4, the nuclear power plant 100A according to the second embodiment differs from the nuclear power plant 100 according to the first embodiment (see FIG. 1) in the following respects. (1) Chimney 30 has been added. (2) The recirculation pump 7, the plant control device 17, the recirculation pump control device 18, the pump rotation speed setting device 25, and the first adjusting means 42a are omitted. (3) The power control unit 23A of the nuclear power plant 100A has a fourth evaluation means 41d and a fourth adjustment means 42d. (4) The reactivity insertion command generated and adjusted by the power control unit 23A is input to the control rod positioner 26, and the core power is adjusted by moving the control rods 8.

[0058] The fourth evaluation means 41d is a means for evaluating the core flow rate due to natural circulation and predicting the amount of change in the core flow rate. The fourth evaluation means 41d evaluates the amount of change in the core flow rate of a natural circulation reactor without a recirculation pump based on the pressure loss characteristics of the recirculation flow path of the reactor 1 evaluated by the first evaluation means 41a. The fourth adjustment means 42d is a means for adjusting a command (reactivity change command) for changing the reactivity in the core. The fourth adjustment means 42d adjusts the reactivity change command based on the change in the core flow rate predicted by the fourth evaluation means 41d.

[0059] The nuclear power plant 100A according to the second embodiment does not have a recirculation pump 7, a recirculation pump control device 18, or a pump rotation speed setting device 25, and therefore adjusts the core power output according to the position of the control rod 8. The control rod control device 19 outputs an insertion or withdrawal command to the control rod 8 to control the position of the control rod 8. At that time, the power control unit 23A inputs an insertion command signal for the control rod 8 from the control rod position setting device 26 of the power control unit 23, the control room 24, and an operation monitoring function not shown in FIG.

[0060] The operation of the nuclear power plant 100A will be described below with reference to Fig. 5. Fig. 5 is an explanatory diagram of the operation of a power control unit 23A of the nuclear power plant 100A. The power control unit 23 of the nuclear power plant 100A according to the second embodiment differs in the operation of the control rod positioner 26. In this embodiment, even when rocking and tilting of the reactor 1 occurs, the position of the control rod 8 is adjusted particularly by the control rod positioner 26 to control the core power according to the reactivity of the control rod 8, thereby mitigating the effects of rocking and tilting and ensuring stable operation of the reactor 1, which is a floating reactor.

[0061] 5, in a nuclear power plant 100A according to the second embodiment, a generator output set value I50 is input to a power control unit 23A from an indicating device 50, a generator output I13 is input from a generator output sensor 13, and a core pressure I4 is input from a pressure sensor 4. In addition, a core output I3 is input to a core power monitor 3, a core flow rate I5 is input from a flow rate sensor 5, a water level I6 is input from a water level sensor 6, and a swing tilt amount I9 is ​​input from a swing tilt sensor 9.

[0062] The output control unit 23A of the nuclear power plant 100A according to the second embodiment performs the processes from step S10 to step S30, similarly to the nuclear power plant 100 according to the first embodiment.

[0063] Moreover, the output control unit 23A of the nuclear power plant 100A according to the second embodiment performs the processes of steps S40 to S44 in the same manner as the nuclear power plant 100 according to the first embodiment.

[0064] However, after step S44, the power control unit 23A of the nuclear power plant 100A according to the second embodiment predicts the amount of change in the core flow rate (step S46) and adjusts the reactivity insertion command according to the amount of change in the core flow rate (step S60).

[0065] Next, the power control unit 23A generates a control rod insertion command instructing the insertion of the control rod 8 in the control rod position setter 26 (see FIG. 1) based on the corrected reactivity insertion command adjusted in step S60 and the control rod position 19 (step S66). The power control unit 23A outputs the control rod insertion command generated in step S66 to the control rod control device 19 (see FIG. 1) to cause the control rod control device 19 (see FIG. 1) to adjust the position of the control rod 8.

[0066] Here, we will provide additional information about the operation of the power control unit 23A shown in Fig. 5. The control rod position setter 26 controls the core power based on the reactivity of the control rod 8. In doing so, the power control unit 23A receives signals from the core power monitor 3, pressure sensor 4, flow rate sensor 5, water level sensor 6, swing inclination sensor 9, and generator output sensor 13. The control rod position setter 26 outputs a command signal to the control rod control device 19 to insert the control rod 8.

[0067] The power control unit 23A calculates a power change command by subtracting the power set value from the power input from the power sensor 13. The power control unit 23A adjusts the pressure input from the pressure sensor 4, performs proportional-integral-derivative control, and calculates a core power command. The power control unit 23A calculates a reactivity insertion command from the difference between the core power command and the core power input from the core power monitor 3. The power control unit 23A prepares a table showing the relationship between the amount of change in acceleration and inclination in the axial direction and in the plane perpendicular to the axis of the reactor 1 input from the swing inclination sensor 9 and the amount of change in reactivity under the condition that the core flow rate does not change, and adjusts the reactivity insertion command.

[0068] Furthermore, the power control unit 23A uses the core power input from the core power monitor 3, the core flow rate input from the flow rate sensor 5, the core water level input from the water level sensor 6, and the axial acceleration of the core 2 input from the rocking inclination sensor 9 to calculate the amount of change in the static pressure difference between the outside and inside of the shroud and the amount of change in the pressure loss characteristics of the recirculation flow path, predict the amount of change in the core flow rate, and adjust the reactivity insertion command according to the amount of change in the core flow rate. The power control unit 23A creates a table of the relationship between the position of the control rod 8 and the reactivity (control rod worth) due to the control rod 8, calculates a control rod insertion command from the corrected reactivity insertion command and the position of the control rod 8, and outputs an insertion command signal for the control rod 8 to the control rod control device 19.

[0069] In the nuclear power plant 100A, vertical and horizontal accelerations are applied to the floating platform 31 as the floating platform 31 moves and tilts in the vertical and horizontal directions. This causes the reactivity of the core to change. Therefore, in this embodiment, in step S66, the control rod position setter 26 (see FIG. 1) of the power control unit 23A generates a control rod insertion command for changing the reactivity of the core so as to reduce the change in reactivity in the core, and outputs the command to the control rod control device 19, thereby adjusting the position of the control rod 8.

[0070] Like the nuclear power plant 100 according to the first embodiment, the nuclear power plant 100A according to the second embodiment uses a floating reactor power control device to mitigate the effects of rocking and tilting, enabling stable and safe operation of the reactor even during rocking and tilting.

[0071] The nuclear power plant 100A according to the second embodiment adjusts the reactor power output by the amount of control rod insertion. In the case of a natural circulation boiling water reactor, the nuclear power plant 100A according to the second embodiment adjusts the reactivity insertion command based on the measured rocking and tilting, thereby mitigating the effects of rocking and tilting on the core power output and core flow rate.

[0072] The nuclear power plant 100A according to the second embodiment can stably operate the reactor 1, which is a floating nuclear reactor, similarly to the nuclear power plant 100 according to the first embodiment.

[0073] [Third embodiment] The configuration of a nuclear power plant 100B according to the third embodiment will be described below with reference to Fig. 6. The nuclear power plant 100B according to the third embodiment is different from the nuclear power plant 100A according to the second embodiment described above in the operation method of the output control unit 23B (particularly, the operation methods of the control rod position setter 26, the extraction steam rate setter 28, and the feedwater flow rate setter 29).

[0074] As shown in FIG. 6, the nuclear power plant 100B according to the third embodiment differs from the nuclear power plant 100A according to the second embodiment (see FIG. 4) in the following respects. The output control unit 23B of the nuclear power plant 100B does not include the fourth evaluation means 41d, but includes a fifth adjustment means 42e, and the feedwater flow rate setter 29 and the extraction air amount setter 28 are changed.

[0075] The feedwater flow rate setter 29 adjusts the recirculation flow rate of the reactor 1 based on a command (reactivity insertion command) to change the reactivity in the reactor core 2. The feedwater flow rate setter 29 calculates a water level adjustment command for the reactor 1 based on the core flow rate adjustment command, and evaluates a feedwater flow rate change command. The feedwater flow rate setter 29 also predicts the core water level change of the reactor 1 based on the adjusted actual feedwater flow rate, and predicts the core flow rate change in the same way as the fourth evaluation means 41d. The fourth adjusting means 42d adjusts the reactivity change command based on the core flow rate change predicted by the feedwater flow rate setter 29, not based on the fourth evaluating means 41d.

[0076] The extraction amount setter 28 adjusts the inlet temperature of the core 2 based on a command (reactivity insertion command) for changing the reactivity in the core 2. The extraction amount setter 28 generates a command for the amount of change in the inlet temperature of the core 2, calculates a command for the feedwater temperature, and evaluates the extraction amount. The fifth adjustment means 42e is a means for adjusting a command (reactivity insertion command) for changing the reactivity in the core. The fifth adjustment means 42e evaluates the inlet temperature of the core 2 based on the feedwater temperature predicted by the extraction amount setter 28, and adjusts the reactivity change amount command.

[0077] The operation of the nuclear power plant 100B will be described below with reference to Fig. 7. Fig. 7 is a diagram illustrating the operation of the output control unit 23B of the nuclear power plant 100B.

[0078] 7, in a nuclear power plant 100B according to the third embodiment, a generator output set value I50 is input to a power control unit 23B from an indicating device 50, a generator output I13 is input to a power control unit 23B from a generator output sensor 13, and a core pressure I4 is input to a pressure sensor 4. Further, a core output I3 is input to a power control unit 23B from a core power monitor 3, a core flow rate I5 is input to a flow rate sensor 5, a water level I6 is input to a water level sensor 6, and a swing tilt amount I9 is ​​input to a swing tilt sensor 9. Further, a control rod position I19 is input to a power control unit 23B from a control rod control device 19.

[0079] The output control unit 23B of the nuclear power plant 100B according to the third embodiment performs the processes from step S10 to step S35, similarly to the nuclear power plant 100 according to the first embodiment.

[0080] Moreover, the output control unit 23B of the nuclear power plant 100B according to the third embodiment performs the processes from step S40 to step S44, similarly to the nuclear power plant 100 according to the first embodiment described above.

[0081] However, after steps S35 and S44, the power control unit 23B of the nuclear power plant 100B according to the third embodiment adjusts the core flow rate by changing the water level rather than by using a recirculation pump, based on the core flow rate adjustment command generated in step S35. The feedwater flow rate setter 29 adjusts the core water level command that instructs an increase or decrease in the core water level based on the core flow rate adjustment command generated in step S35, the change in the pressure loss characteristics of the core recirculation flow path evaluated in step S44, and the swing tilt amount I9 of the reactor 1 measured by the swing tilt sensor 9 (step S82). The power control unit 23B identifies a mismatch between the main steam flow rate and the feedwater flow rate (step S84). The feedwater flow rate setter 29 adjusts the feedwater flow rate command that instructs an increase or decrease in the feedwater flow rate, based on the core water level command adjusted in step S82 and the mismatch identified in step S84 (step S86). The feedwater flow rate control device 22 generates a signal to the feedwater pump 15 based on the feedwater flow rate command adjusted in step S86, and feeds back the feedwater flow rate to be actually changed to the feedwater flow rate setter 29. The feedwater flow rate setter 29 predicts the amount of change in the core water level based on the fed-back feedwater flow rate (step S88), and predicts the amount of change in the core flow rate in the same way as the fourth evaluation means 41d of the second embodiment (step S46a). The power control unit 23B adjusts the reactivity insertion command according to the amount of change in the core flow rate predicted in step S46a (step S60).

[0082] When the adjustment of the feedwater flow rate is started, the power control unit 23B evaluates the command for the core 2 inlet temperature by tabulating the relationship between the steam volume fraction in the core 2 and the temperature of the cooling water at the core 2 inlet based on the reactivity insertion command adjusted in step S60 (step S90). The extraction steam amount setter 28 adjusts a feedwater temperature command that instructs an increase or decrease in the feedwater temperature based on the core inlet temperature command evaluated in step S90 and the feedwater flow rate input from the feedwater flow rate control device 22 (step S92) and adjusts an extraction steam amount command that instructs an increase or decrease in the extraction steam amount (step S94). The extraction steam control device 21 generates an opening signal to the extraction valve 16 based on the extraction steam amount command adjusted in step S94 and feeds back the actual extraction steam amount to be changed to the extraction steam amount setter 28. The extraction steam amount setter 28 predicts a change in the feedwater temperature based on the feedback extraction steam amount (step S95).

[0083] Next, the power control unit 23B finally adjusts the reactivity insertion command based on the feedwater temperature predicted in step S95 (step S96).

[0084] Next, the power control unit 23B generates a control rod insertion command in the control rod position setter 26 (see FIG. 6) based on the final reactivity insertion command and the control rod position 19 adjusted in step S96 (step S98). The power control unit 23B outputs the control rod insertion command generated in step S98 to the control rod control device 19 (see FIG. 6) to cause the control rod control device 19 (see FIG. 6) to adjust the position of the control rod 8.

[0085] Here, the operation of the power control unit 23B shown in Fig. 7 will be explained in more detail. After adjusting the generated reactivity insertion command by swinging and tilting, the power control unit 23B outputs a corrected reactivity insertion command to the feedwater flow rate setter 29.

[0086] The feedwater flow rate setter 29 uses the corrected reactivity insertion command, the swing tilt amount input from the swing tilt sensor 9, the core power input from the core power monitor 3, the core flow rate input from the flow rate sensor 5, and the water level of the reactor 1 input from the water level sensor 6 to calculate a core flow rate command equivalent to the corrected reactivity insertion command, and calculates a reactor water level command equivalent to this core flow rate command. The feedwater flow rate setter 29 also calculates a water level change amount command adjusted so as not to exceed the operational water level limit. Based on the mismatch between the main steam flow rate and the feedwater flow rate evaluated by the power control unit 23B and the operational water level limit, the feedwater flow rate setter 29 calculates a feedwater flow rate command equivalent to the water level change amount command, and outputs a feedwater flow rate command signal.

[0087] When the water level approaches the operating limit or the command for the amount of change in the feedwater flow rate exceeds the limit value for the rate of change in the feedwater flow rate, a discrepancy occurs between the command for changing the core water level and the actual amount of change in the water level. Therefore, the power control unit 23B predicts the core water level and the core flow rate. The power control unit 23B predicts the amount of reactivity insertion due to the amount of change in the core flow rate, and outputs an additional reactivity insertion command to the extraction amount setter 28 to compensate for the shortfall in the corrected reactivity insertion command.

[0088] The power control unit 23B calculates a core inlet temperature command based on the additional reactivity insertion command. The extraction amount setter 28 uses the core inlet temperature command, the core power input from the core power monitor 3, the core flow rate input from the flow rate sensor 5, and the water level and feedwater flow rate of the reactor 1 input from the water level sensor 6 to calculate a feedwater temperature command, calculate the extraction amount, and output a extraction amount signal to the extraction control device 21.

[0089] When the bleed rate signal exceeds the limit value, a discrepancy occurs between the bleed rate command and the actual bleed rate, so the bleed rate setter 28 predicts the feedwater temperature. The power control unit 23B predicts the core inlet temperature, predicts the amount of reactivity insertion depending on the amount of change in the core inlet temperature, and outputs a final reactivity insertion command to the control rod position setter 26 to compensate for the insufficiency of the additional reactivity insertion command.

[0090] The power control unit 23B calculates a control rod insertion command from the final reactivity insertion command and the position of the control rod 8 using the control rod position setter 26, and outputs a control rod 8 insertion command signal to the control rod control device 19.

[0091] The nuclear power plant 100B according to the third embodiment, like the nuclear power plants 100 and 100A according to the other embodiments, uses a floating reactor power control device to mitigate the effects of swaying and tilting, enabling stable and safe operation of the reactor even during swaying and tilting.

[0092] Furthermore, the nuclear power plant 100B according to the third embodiment adjusts the reactor power by the feedwater flow rate and feedwater temperature in addition to the control rod insertion amount. In the case of a natural circulation boiling water reactor, the nuclear power plant 100B according to the third embodiment can reduce control rod movement by partially inserting the reactivity based on the feedwater flow rate and feedwater temperature in response to a reactivity insertion command.

[0093] The nuclear power plant 100B according to the third embodiment can stably operate the reactor 1, which is a floating nuclear reactor, similarly to the nuclear power plants 100 and 100A according to the other embodiments.

[0094] The present invention is not limited to the above-described embodiments and includes various modifications. For example, the above-described embodiments have been described in detail to clearly explain the present invention, and the present invention is not necessarily limited to those including all of the described configurations. It is possible to replace part of the configuration of the embodiments with other configurations, and it is also possible to add other configurations to the configuration of the embodiments. Furthermore, it is possible to add, delete, or replace part of each configuration with other configurations.

[0095] For example, the nuclear power plant 100 according to the first embodiment (see Fig. 1) can be modified into a nuclear power plant 100C shown in Fig. 8. Fig. 8 is an overall configuration diagram of the modified nuclear power plant 100C.

[0096] As shown in FIG. 8, the output control unit 23C of the modified nuclear power plant 100C differs from the nuclear power plant 100 according to the first embodiment (see FIG. 1) in that a predicted value calculation unit 32 that calculates a predicted value from an actual measurement value is added.

[0097] The power control unit 23C executes the process shown in FIG. 2. At that time, the power control unit 23C uses the predicted value calculation unit 32 to predict predicted values ​​of the rocking and tilt of the floating platform 31 after a given time has elapsed from the instantaneous measurement results measured by the rocking and tilt sensor 9. The power control unit 23C outputs the instantaneous measurement results and the prediction. Then, the power control unit 23C uses the instantaneous and predicted data when making the evaluation in step S44 and the adjustment in step S48. Therefore, the nuclear power plant 100C of the modified example uses not only the actual measured values ​​measured by the rocking and tilt sensor 9 but also the predicted values ​​for the adjustments made by the power control unit 23C. Such a power control unit 23C can adjust the power output of the reactor 1 using the predicted values ​​calculated by the predicted value calculation unit 32.

[0098] Furthermore, for example, the output control unit 23A of the nuclear power plant 100A according to the second embodiment described above may be configured to mutually feed back the change in core flow path pressure loss characteristic predicted by the first evaluation means 41a and the change in reactivity predicted by the second evaluation means 41b, or the core flow rate predicted by the fourth evaluation means 41d, and adjust the reactivity change command and the core flow rate change command. [Explanation of symbols]

[0099] 1. Nuclear reactor (floating reactor) 2. Reactor core 3 Core power monitor 4 Pressure Sensors 5 Flow Sensor 6 Water level sensor 7 Recirculation Pump 8 control rods 9. Swing tilt sensor (measurement unit) 10 Steam control valve 11 Turbine 12. Generator 13 Generator output sensor 14 Condenser 15 Water supply pump 16 Bleed valve 17 Plant control device (output control device) 18 Recirculation pump control device 19 Control rod control device 20 Pressure control device 21 Bleeding control device 22 Water supply flow control device 23, 23A, 23B Output control section 24 Control Room 25 Pump rotation speed setting device 26 Control rod positioner 27 Regulating valve opening setting device 28 Bleed air amount setting device 29 Water supply flow rate setting device 30 Chimney 31 Floating Platform 32 Prediction value calculation section 41a First evaluation means 41b Second evaluation method 41c Third evaluation method 41d Fourth evaluation method 42a First adjustment means 42b Second adjustment means 42c Third adjustment means 42d Fourth adjustment means 42e Fifth adjustment means 50 Indicating device 100, 100A, 100B, 100C Nuclear Power Plant I3 Core power I4 Core pressure I5 Core flow rate I6 water level I9 Swing tilt amount I13 Generator output I19 Control rod position I50 Generator output setting value

Claims

1. A measurement unit that measures the amount of tilt of a floating nuclear reactor; a power control unit that adjusts the power output of the floating reactor in accordance with the tilt amount of the floating reactor measured by the measurement unit, The output control unit a first evaluation means for calculating a static pressure difference between the inside and outside of the core, calculating a change in the static pressure difference based on the rocking tilt amount of the floating reactor measured by the measurement unit, and evaluating a change in the pressure loss characteristic of the core recirculation flow path; and a first adjusting means for adjusting the rotation speed of the recirculation pump based on the pressure loss characteristic of the core recirculation flow path evaluated by the first evaluating means and the core flow rate adjustment command generated by the power control unit. A power control device for a floating nuclear reactor.

2. 2. The power control device for a floating nuclear reactor according to claim 1, The output control unit a second evaluation means for evaluating a change in reactivity due to changes in various parameters including boiling characteristics, steam volume fraction, and steam volume fraction distribution in the core fuel based on the rocking tilt amount of the floating reactor measured by the measurement unit, and predicting the change in reactivity; and second adjusting means for adjusting a reactivity change command based on the reactivity change evaluated by the second evaluating means. A power control device for a floating nuclear reactor.

3. 3. The power control device for a floating nuclear reactor according to claim 2, The output control unit The change amount of the pressure loss characteristic of the core recirculation flow path predicted by the first evaluation means and the change amount of the reactivity evaluated by the second evaluation means are fed back to each other, and the reactivity change amount command and the core flow rate change amount command are adjusted. A power control device for a floating nuclear reactor.

4. 2. The power control device for a floating nuclear reactor according to claim 1, The output control unit a third evaluation means for evaluating a change in critical power at which boiling transition occurs and predicting a margin for critical power based on the tilt amount of the floating reactor measured by the measurement unit; and third adjustment means for suppressing the output when the margin for the critical output predicted by the third evaluation means becomes smaller than a preset threshold value. A power control device for a floating nuclear reactor.

5. A measurement unit that measures the amount of rocking tilt of a floating nuclear reactor; a power control unit that adjusts the power output of the floating reactor in accordance with the tilt amount of the floating reactor measured by the measurement unit, The output control unit a first evaluation means for calculating a static pressure difference between the inside and outside of the core, calculating a change in the static pressure difference based on the rocking tilt amount of the floating reactor measured by the measurement unit, and evaluating a change in the pressure loss characteristic of the core recirculation flow path; a fourth evaluation means for predicting a change in core flow rate of a natural circulation reactor without a recirculation pump based on the pressure loss characteristics of the core recirculation flow path evaluated by the first evaluation means; The reactivity change command is adjusted based on the core flow rate change predicted by the fourth evaluation means. and a fourth adjusting means A power control device for a floating nuclear reactor.

6. 6. The power control device for a floating nuclear reactor according to claim 5, The output control unit a second evaluation means for evaluating a change in reactivity due to changes in various parameters including boiling characteristics, steam volume fraction, and steam volume fraction distribution in the core fuel based on the rocking tilt amount of the floating reactor measured by the measurement unit, and predicting the change in reactivity; and second adjusting means for adjusting a reactivity change command based on the reactivity change evaluated by the second evaluating means. A power control device for a floating nuclear reactor.

7. 7. The power control device for a floating nuclear reactor according to claim 6, The output control unit a feedwater flow rate setter that calculates a core water level command based on a core flow rate command calculated in response to a reactivity insertion command and the amount of tilt of the floating reactor measured by the measurement unit, and when a feedwater flow rate command is adjusted, predicts a change in the core water level in response to an actual feedwater flow rate, and predicts a change in the core flow rate in response to the change in the core water level and the amount of tilt of the floating reactor measured by the measurement unit. A power control device for a floating nuclear reactor.

8. A power control process for adjusting the output of the floating reactor in accordance with the measured tilt amount of the floating reactor, The method further includes a recirculation pump rotation speed adjusting step of adjusting the rotation speed of the recirculation pump according to the change in the static pressure difference between the inside and outside of the core. A method for operating a floating nuclear reactor, comprising:

9. 9. The method for operating a floating nuclear reactor according to claim 8, a core flow rate change prediction step of predicting a core flow rate change; and a control rod position adjusting step of adjusting the control rod position in accordance with the change in the core flow rate. A method for operating a floating nuclear reactor, comprising:

10. 9. The method for operating a floating nuclear reactor according to claim 8, a critical power change evaluation step for evaluating the amount of change in critical power at which boiling transition occurs; an output suppression control step of suppressing the output when the margin of margin with respect to the critical output becomes smaller than a preset threshold value. A method for operating a floating nuclear reactor, comprising:

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