JP5667435B2 - Cogeneration nuclear power generation system - Google Patents

Description

  The present invention relates to a cogeneration nuclear power generation system, and more particularly to a cogeneration nuclear power generation system suitable for application to a boiling water nuclear power plant.

  A boiling water nuclear power plant supplies steam generated in a nuclear reactor to a high-pressure turbine and a low-pressure turbine through main steam piping. These turbines are rotated by steam and rotate a generator connected to the turbine. Thereby, power generation is performed. The steam exhausted from the low-pressure turbine is condensed into water by the condenser. This water passes through the feed water pipe as feed water, is heated by a plurality of feed water heaters provided in the feed water pipe, and is supplied to the nuclear reactor.

  A cogeneration system has been proposed in which a boiling water nuclear power plant is used not only as a power generator but also as a heat source to improve the substantial thermal efficiency of the entire plant (Japanese Patent Laid-Open Nos. 59-48696 and 63-63). 3298, JP-A-4-140406, and JP-A-2007-51565).

  In the cogeneration nuclear power generation system described in Japanese Patent Application Laid-Open No. 59-48696, hot water supplied to high-pressure side heat utilization equipment with a heat exchanger using steam extracted from the main steam pipe upstream of the high-pressure turbine. Is generated. Moreover, using the steam extracted from the exhaust side of the high-pressure turbine, hot water supplied to the low-pressure side heat utilization facility is generated by another heat exchanger.

  Japanese Patent Laid-Open No. 63-3298 describes a cogeneration nuclear power generation system in which water discharged from a condenser is heated with steam extracted from a steam turbine, and the water flowing in the hot water supply pipe is heated using the superheated waste water. doing.

  A combined heat and power generation system that heats the drainage of a condenser that condenses steam exhausted from a steam turbine with steam extracted from the steam turbine to generate hot water and supplies the hot water to hot water utilization equipment is disclosed in Japanese Patent Laid-Open No. Hei 4 -140406.

  Japanese Patent Application Laid-Open No. 2007-51565 also describes a cogeneration system. In this cogeneration system, steam extracted from the high-pressure turbine and low-pressure turbine of the power plant is led to separate heat exchangers, and water is sequentially heated by these heat exchangers, and hot water supplied to the hot water utilization facility is supplied. Is generated.

JP 59-48696 A JP-A 63-3298 JP-A-4-140406 JP 2007-51565 A

  In JP-A-59-48696 and JP-A-2007-51565, when generating hot water to be supplied to hot water utilization equipment, hot water is generated by heating with steam used for power generation. The heat utilization rate is improved. However, the heat efficiency of the power generation portion is not improved because the heat of the hot drainage discharged from the condenser that condenses the steam exhausted from the turbine is not used.

  In the combined heat and power generation system described in Japanese Patent Laid-Open Nos. 63-3298 and 4-140406, hot water supplied to the hot water utilization facility is discharged from a condenser that condenses steam exhausted from the turbine. Therefore, the substantial thermal efficiency of the power generation portion is improved. In Japanese Patent Laid-Open No. 63-3298 and Japanese Patent Laid-Open No. 4-140406, hot water having a temperature of 50 ° C. to 100 ° C., which is meaningful as heat utilization in a hot water utilization facility, is generated using the warm drainage of a condenser. Therefore, it is necessary to use much of the heat generated in the steam generator (reactor etc.) for the production of the hot water, and it is necessary to keep the power generation amount in the power plant low.

  An object of the present invention is to provide a cogeneration nuclear power generation system that can absorb either the reactor heat output or load fluctuation and improve the thermal efficiency.

A feature of the present invention that achieves the above-mentioned object is a steam generator, a turbine that is supplied with steam generated by the steam generator and rotates a generator, a condenser that condenses steam exhausted from the turbine, A water supply pipe connecting the condenser and the steam generator, a hot water generator, and a controller;
The hot water generating device includes a first pipe that guides water, a plurality of second pipes that guide steam extracted from the turbine, and water that is connected to each second pipe and is provided in the first pipe and is guided by the first pipe A plurality of heating devices for heating with steam supplied by the second pipe, a first flow control valve provided in the second pipe, and a plurality of second flow control valves provided for each of the plurality of second pipes Have
It has the said control apparatus which controls each opening degree of a 1st flow control valve and a several 2nd flow control valve based on a nuclear reactor heat output.

  By controlling the opening degree of each of the first flow rate control valve and the plurality of second flow rate control valves, the amount of heat used in the hot water generator and the nuclear power plant Since the ratio of the amount of power generation can be adjusted, fluctuations in the reactor thermal output of the nuclear power plant can be absorbed by using heat in the hot water generator, and the thermal efficiency of the entire system in the cogeneration nuclear power generation system can be improved. improves.

A steam generator, a steam generated by the steam generator, a turbine that rotates the generator, a condenser that condenses the steam exhausted from the turbine, a hot water generator, and a controller,
The hot water generating device includes a first pipe for guiding water, a plurality of second pipes for guiding steam extracted from the turbine, and the first pipe connected to each second pipe and guided by the first pipe. A plurality of heating devices for heating water with steam supplied through a second pipe, a first flow control valve provided in the second pipe, and a plurality of second flow control valves provided for each of the plurality of second pipes And
The object described above can also be achieved by including the control device that controls the opening degree of each of the first flow rate control valve and the plurality of second flow rate control valves based on the power generation amount request signal.

  By controlling the opening of each of the first flow control valve and the plurality of second flow control valves, the amount of heat used in the hot water generator using the power generation request signal corresponding to the load fluctuation of the nuclear power plant Since the ratio of the amount of power generated in the nuclear power plant can be adjusted, load fluctuations in the nuclear power plant can be absorbed by the heat utilization in the hot water generator, and the thermal efficiency of the entire system in the cogeneration nuclear power generation system Will improve.

  According to the present invention, since the ratio between the heat utilization amount and the power generation amount is adjusted, either the reactor heat output or the load fluctuation can be absorbed, and the thermal efficiency can be improved.

BRIEF DESCRIPTION OF THE DRAWINGS It is a block diagram of the cogeneration nuclear power generation system of Example 1 which is one suitable Example of this invention. It is explanatory drawing which shows an example of the process which produces | generates the control signal of the control performed with the control apparatus shown in FIG. It is explanatory drawing which shows an example of the change of the reactor heat output and generator output in the conventional boiling water nuclear power plant. It is explanatory drawing which shows an example of the change of the reactor heat output corresponding to the load fluctuation in a conventional boiling water nuclear power plant, and a generator output. It is explanatory drawing which shows an example of the change of the reactor heat output and generator output in the boiling water nuclear power plant included in the cogeneration nuclear power generation system of Example 1. It is a block diagram of the cogeneration nuclear power generation system of Example 2 which is another Example of this invention. It is a block diagram of the cogeneration nuclear power generation system of Example 3 which is another Example of this invention. It is a block diagram of the cogeneration nuclear power generation system of Example 4 which is another Example of this invention. It is explanatory drawing which shows an example of the change of the reactor heat output and generator output in the boiling water nuclear power plant included in the cogeneration nuclear power generation system of Example 4. It is explanatory drawing which shows an example of the process which produces | generates the control signal of the control performed with the control apparatus shown in FIG. It is a block diagram of the cogeneration nuclear power generation system of Example 5 which is another Example of this invention.

  The inventors examined a measure for substantially improving the thermal efficiency of the entire plant in the combined heat and nuclear power generation system including the nuclear power plant.

  As a result, as described in Japanese Patent Application Laid-Open No. 4-140406, the inventors heated a part of the drainage from the condenser, which condensed the steam exhausted from the turbine, with the extracted steam from the turbine. The conclusion was reached that hot water should be generated. This makes it possible to use heat that has been thrown away into the sea (or river), and to improve the substantial thermal efficiency of the entire nuclear power plant. Furthermore, the inventors determine the amount of drainage that is taken out from the drain pipe connected to the heat transfer pipe of the condenser and becomes hot water, and the amount of extraction steam from a turbine or the like that heats the drainage, and the reactor heat output (or It was newly found out that by controlling according to the load fluctuation), the fluctuation (or load fluctuation) of the reactor heat output can be absorbed by the heat utilization in the hot water generator.

  As a result, it is possible to improve the substantial thermal efficiency of the entire nuclear power plant that constitutes the cogeneration nuclear power generation system, and to configure the cogeneration nuclear power generation system that can cope with the reactor heat output (or load fluctuation).

  An embodiment of the present invention reflecting the above examination results will be described below.

  A cogeneration nuclear power generation system according to embodiment 1, which is a preferred embodiment of the present invention, will be described with reference to FIG. The cogeneration nuclear power generation system 1 of the present embodiment uses a boiling water nuclear power plant as the nuclear power plant. The cogeneration nuclear power generation system 1 includes a boiling water nuclear power plant 2, a hot water generation device 17, and a control device 24.

  The boiling water nuclear power plant 2 includes a nuclear reactor 3, a main steam pipe 4, a high-pressure turbine 5, a low-pressure turbine 6, a condenser 7, and a water supply pipe 11 that are steam generators. A nuclear reactor 3 having a core 26 loaded with a plurality of fuel assemblies is connected to a high-pressure turbine 5 and a low-pressure turbine 6 by a main steam pipe 4. The condenser 7 to which the low pressure turbine 6 is connected has a plurality of heat transfer tubes 8 inside. Cooling water supply pipes 9 having suction ports arranged in the sea (or river) are connected to the respective inlets of the heat transfer pipes 8. A cooling water discharge pipe 10 is communicated with each outlet of the heat transfer pipes 8. A water supply pipe 11 connects the condenser 7 and the nuclear reactor 3. The low-pressure feed water heater 12, the feed water pump 13, and the high-pressure feed water heater 14 are installed in the feed water pipe 11 in this order from the upstream. The extraction steam pipe 15 connected to the high pressure turbine 5 is connected to the high pressure feed water heater 14. A bleed steam pipe 16 connected to the low pressure turbine 6 is connected to the low pressure feed water heater 12.

  A pressure gauge 27 is installed in the main steam pipe 4 upstream of the high-pressure turbine 5. A flow meter 28 and a thermometer 29 are provided in the feed water pipe 11 downstream of the high pressure feed water heater 14. A pressure gauge 27, a flow meter 28, and a thermometer 29 are connected to the reactor thermal output monitor 25, and the reactor thermal output monitor 25 is connected to the control device 24.

  The hot water generator 17 includes a hot water supply pipe 18, a heat utilization water flow control valve 19, heaters 21A and 21B, heating steam supply pipes 22A and 22B, and heating steam flow control valves 23A and 23B. A hot water supply pipe 18 provided with a heat utilization water flow rate control valve 19 is connected to the cooling water discharge pipe 10. The heater 21 </ b> A, the pump 20, and the heater 21 </ b> B are provided in the hot water supply pipe 18 in this order from the heat utilization water flow rate control valve 19 toward the downstream. The hot water supply pipe 18 is connected to a hot water utilization facility (not shown). A heating steam supply pipe 22A provided with a heating steam flow rate control valve 23A connects the low-pressure turbine 6 and the heater 21A. A heating steam supply pipe 22B provided with a heating steam flow rate control valve 23B connects the high-pressure turbine 5 and the heater 21B. Installation of the pump 20 is indispensable when the temperature of the hot water flowing out from the final stage heater (heater 21B in this embodiment) exceeds 100 ° C. However, when the temperature is lower than 100 ° C., the pump 20 may not be installed. Moreover, the order of the heat utilization water flow control valve 19, the pump 20, and the heater 21 may be changed from the arrangement order shown in FIG.

  The connection positions of the heating steam supply pipes 22A and 22B to the low pressure turbine 6 and the high pressure turbine 5 are determined by the temperature of the hot water supplied from the hot water generator 17 to the hot water utilization facility. When the temperature of the hot water is high, the connection position of the heating steam supply pipe 22A to the low pressure turbine 6 and the connection position of the heating steam supply pipe 22B to the high pressure turbine 5 are closer to the steam inlets of the respective turbines. Become.

  Steam generated in the core 26 in the nuclear reactor 3 is supplied to the high-pressure turbine 5 and the low-pressure turbine 6 through the main steam pipe 4 to rotate these turbines. A generator connected to the turbine is rotated to generate electricity. The steam exhausted from the low pressure turbine 6 is exhausted to the condenser 7. The cooling water guided by the cooling water supply pipe 9 is supplied to each heat transfer pipe 8 provided in the condenser 7. This cooling water cools and condenses the steam flowing outside the heat transfer tubes 8 in the condenser 7. The cooling water used for condensing the steam is discharged from each heat transfer pipe 8 to the cooling water discharge pipe 10.

  The water generated by the condensation of the steam is boosted by the feed water pump 13 as feed water and supplied to the reactor 3 through the feed water pipe 11. The feed water flowing in the feed water pipe 11 is composed of a low pressure feed water heater 12 to which the steam extracted from the low pressure turbine 6 is guided through the extraction steam pipe 16 and the steam extracted from the high pressure turbine 5 through the extraction steam pipe 15. It is heated by the high-pressure feed water heater 14 that is guided. Feed water whose temperature has risen due to heating by these feed water heaters is supplied to the reactor 3. The extracted steam supplied to the low-pressure feed water heater 12 and the high-pressure feed water heater 14 is condensed by heating the feed water to become drain water. These drain waters are supplied to the condenser 7 by a drain pipe (not shown) connected to the low pressure feed water heater 12 and the high pressure feed water heater 14 and mixed with the feed water.

  A part of the cooling water whose temperature rises due to condensation of steam in the heat transfer pipe 8 of the condenser 7 and is discharged to the cooling water discharge pipe 10 flows into the hot water supply pipe 18. The cooling water is heated by the heaters 21 </ b> A and 21 </ b> B to become warm water whose temperature is further increased, and is pressurized by the pump 20 and supplied to the hot water utilization facility through the warm water supply pipe 18. Steam extracted from the low-pressure turbine 6, which is a heating source of hot water, is supplied to the heater 21A through the heating steam supply pipe 22A. The steam extracted from the high-pressure turbine 5, which is a heating source of hot water, is supplied to the heater 21B through the heating steam supply pipe 22B. The cooling water that has flowed into the hot water supply pipe 18 from the cooling water discharge pipe 10 is heated by the steam. The extraction steam supplied to each of the heaters 21A and 21B is condensed by heating the cooling water flowing in the hot water supply pipe 18, becomes drain water, and is connected to the heaters 21A and 21B. It is led to the condenser 7 by a drain pipe (not shown) different from the drain pipe and mixed with the feed water.

  Below, absorption of the fluctuation | variation of the reactor thermal output in a present Example is demonstrated.

  FIG. 3 shows an example of changes in the operation time of the reactor heat output and generator output of a conventional boiling water nuclear power plant. In the operation example shown here, the reactor heat output and the generator output are kept constant. In actual operation of a boiling water nuclear power plant, even if the reactor heat output is kept constant, the seawater temperature changes depending on the season, and the thermal efficiency of the boiling water nuclear power plant changes. Slightly fluctuates. For this reason, either the reactor heat output or the generator output is controlled to be kept constant. However, since the change in thermal efficiency due to seasonal changes is slight, each characteristic shown in FIG. 2 shows an example in which the change in thermal efficiency due to seasons is ignored.

  FIG. 4 shows an example of changes in the reactor heat output and the generator output when the operation corresponding to the load fluctuation of the boiling water nuclear power plant is performed. There is an example of changing the generator output of a nuclear power plant in response to a load change when the ratio of the power generation amount of the nuclear power plant in the total power generation is high overseas. In a conventional boiling water nuclear power plant, the generator output is changed by changing the reactor heat output.

  FIG. 5 shows an example of changes in reactor thermal output and generator output corresponding to changes in reactor thermal output in the present embodiment. In FIG. 5, the reference heat output (reference reactor heat output) used in the present embodiment is indicated by a one-dot chain line, and the reactor heat output indicated by the solid line is the reactor heat power determined by the core design. Is the output. Examples corresponding to load fluctuations will be described in the fourth and subsequent embodiments.

  In a conventional boiling water nuclear power plant, the reactor heat output is constant as shown in FIG. 3, but this is because the reactor heat output is kept constant by controlling the core flow rate and adjusting the position of the control rod. This is because. If such control is not performed, the reactor heat output varies with the burning of nuclear fuel material in the core. The change trend of reactor heat output varies with fuel design and core design.

  In the cogeneration nuclear power generation system 1 of the present embodiment, the reactor heat output is allowed to fluctuate, and in order to keep the reactor heat output constant at the time of rated output operation as in a conventional boiling water nuclear power plant. There is no need to frequently control the core flow rate (for example, increase the core flow rate) and change the control rod position. That is, in the present embodiment, in one operation cycle from the start of the boiling water nuclear power plant 2 to the stop of the operation of the boiling water nuclear power plant 2 for replacement of the fuel assembly, particularly during rated output operation. Several times, the reactor heat output is adjusted by changing the core flow rate and the control rod position.

  The temperature control of the hot water produced | generated with the hot water production | generation apparatus 17 in a present Example is demonstrated.

  The pressure gauge 27 measures the pressure of the steam supplied from the nuclear reactor 3 to the high pressure turbine 5. The flow meter 28 measures the flow rate of the feed water supplied to the reactor 3, and the thermometer 29 measures the temperature of the feed water discharged from the high-pressure feed water heater 14 at the final stage and supplied to the reactor 3. The measured steam pressure, feed water flow rate, and feed water temperature are input to the reactor thermal output monitor 25. The reactor thermal output monitor 25 calculates the reactor thermal output based on the input steam pressure, feed water flow rate, and feed water temperature. Specifically, the amount of heat input to the reactor 3 by water supply is obtained using the water supply temperature and the water supply flow rate. The saturation temperature of the steam is determined using the steam pressure, and the amount of heat input to the high pressure turbine 5 by the steam is determined based on the flow rate of steam supplied to the high pressure turbine 5 and the saturation temperature. The steam flow rate supplied to the high-pressure turbine 5 is the same as the feed water flow rate because the water level in the nuclear reactor 3 is adjusted to be constant. The steam flow rate may be measured by installing a flow meter in the main steam pipe 3. The reactor heat output monitor 25 calculates the reactor heat output by subtracting the amount of heat input to the reactor 3 by water supply from the amount of heat input to the high-pressure turbine 5 by steam. It can be said that such a reactor thermal output monitor 25 for calculating the reactor thermal output measures the reactor thermal output of the boiling water nuclear power plant 2. The reactor heat output obtained (measured) by the reactor heat output monitor 25 is input to the control device 24.

  Control of each flow control valve 19, 23A, 23B of the hot water generator 17 will be described with reference to FIG. Control of these flow control valves is performed by the control device 24. The control device 24 obtains the opening degree of the heat utilization water flow rate control valve 19 by each calculation in steps S1 and S2, and obtains the respective opening amounts of the heating steam flow rate control valves 23A and 23B by each calculation in steps S3 and S4. . A reference heat output (reference reactor heat output) indicated by a one-dot chain line in FIG. 5 is stored in a storage device (not shown) of the control device 24. The reference heat output may be, for example, the minimum heat output in one operation cycle (except when the boiling water nuclear power plant 2 is started and stopped). When a certain amount of heat is used even at the minimum heat output (when hot water is supplied from the hot water generator 17 to the hot water use facility), the heat use water flow rate and the heating steam flow rate are also provided at the reference heat output. To be greater than zero. When only power generation is performed at the reference heat output, the heat utilization water flow rate and the heating steam flow rate at the reference heat output are both set to zero.

  Further, the storage device of the control device 24 stores information on the four characteristics shown in FIG. The first characteristic is a characteristic indicating the relationship between the nuclear reactor heat output and the flow rate of heat utilization water. The second characteristic is a characteristic indicating the relationship between the heat utilization water flow rate and the valve opening, the third characteristic is a characteristic indicating the relationship between the reactor thermal output and the heating steam flow rate, and the fourth characteristic is the reactor thermal output. It is a characteristic which shows the relationship of valve opening. When the reactor heat output increases, the flow rates of the heat utilization water and the heating steam respectively increase. When the flow rate of the heat utilization water increases, the opening degree of the heat utilization water flow rate control valve 19 increases, and when the flow rate of the heating steam increases, the respective opening amounts of the heating steam flow rate control valves 23A and 23B increase. .

  In step S1, the difference in the reactor heat output from the reference heat output is obtained by subtracting the reference heat output from the reactor heat output input from the reactor heat output monitor 25. Using the first characteristic, the flow rate of the heat utilization water with respect to the difference in the reactor heat output (reactor heat output on the horizontal axis of the first characteristic) is obtained. In step S2, the opening degree of the heat use water flow rate control valve 19 with respect to the heat use water flow rate obtained in step S1 is obtained using the second characteristic. The control device 24 controls the opening degree of the heat utilization water flow rate control valve 19 based on the opening degree obtained in step S2.

  In step S3, the difference of the reactor thermal output from the reference thermal output is obtained by subtracting the reference thermal output from the reactor thermal output input from the reactor thermal output monitor 25. Using the third characteristic, the steam flow rate for heating with respect to the difference in the reactor heat output (reactor heat output on the horizontal axis of the third characteristic) is obtained. In step S4, the opening degree of each of the heating steam flow rate control valves 23A and 23B with respect to the heating steam flow rate obtained in step S3 is obtained using the fourth characteristic. The control device 24 controls the respective opening degrees of the heating steam flow rate control valves 23A and 23B based on the respective opening degrees obtained in step S4.

  With the control by the control device 24 described above, the opening degree of the heat utilization water flow rate control valve 19 is controlled so that the cooling water corresponding to the heat utilization water flow rate obtained in step S1 flows, and the steam flow control for heating is performed. The valves 23A and 23B are controlled in their opening degrees so that extracted steam corresponding to the heating steam flow rate obtained in step S3 is supplied to the heaters 21A and 21B, respectively. By controlling the opening degree of the heat use water flow rate control valve 19 as described above, the hot water supply pipe 18 is supplied from the cooling water discharge pipe 10 to the coolant at a flow rate corresponding to the heat use water flow rate obtained in step S1. Flows in.

  Opening of each of the steam flow control valves for heating 23A and 23B based on the reactor heat output of the difference between the reactor heat output of the boiling water nuclear power plant 2 and the reference heat output (hereinafter referred to as the reactor heat output difference). The generator output is held at the generator output corresponding to the reference heat output since the degree of extraction and the amount of extracted steam supplied to the heaters 21A and 21B are adjusted. For example, when the reactor heat output of the boiling water nuclear power plant 2 increases, the flow rate of steam flowing through the high-pressure turbine 4 and the low-pressure turbine 5 increases, and the generator output increases. In the present embodiment, in such a case, the difference in the reactor thermal output increases, so that the respective openings of the heating steam flow control valves 23A and 23B increase, and the heater 21A is heated by the heating steam supply pipes 22A and 22B. , 21B respectively increases the amount of extracted steam. The flow rate of the steam flowing through the high-pressure turbine 4 and the low-pressure turbine 5 is reduced, and the generator output is held at the generator output (broken line shown in FIG. 5) corresponding to the reference heat output.

  When each opening degree of the steam flow control valves 23A and 23B for heating is increased, the hot water temperature at the outlet of the final stage heater 21 (heater 21B in this embodiment) provided in the hot water supply pipe 18 is increased. To rise. The control device 24 increases the opening degree of the heat utilization water flow rate control valve 19 to the opening degree obtained based on the reactor heat output difference as described above. Thereby, the flow rate of the cooling water flowing into the hot water supply pipe 18 from the cooling water discharge pipe 10 is increased, and the temperature rise of the hot water supplied to the hot water utilization facility through the hot water supply pipe 18 is suppressed. As a result, the hot water temperature is maintained at a predetermined temperature.

  According to the present embodiment, the heating steam flow rate control valves 23A and 23B are controlled based on the reactor heat output difference that is the difference between the measured reactor heat output of the boiling water nuclear power plant 2 and the reference heat output. Since the flow rate of the extracted steam supplied to the heaters 21A and 21B is adjusted by controlling the respective opening degrees, the cooling water flowing into the hot water supply pipe 18 from the cooling water discharge pipe 10 corresponds to the reactor heat output difference. It can heat with heater 21A, 21B with calorie | heat amount, and can produce | generate warm water. Since the flow rate of the extracted steam supplied to the heaters 21A and 21B is controlled based on the difference in reactor heat output, the generator output corresponds to the reference heat output at least during rated output operation of one operation cycle. It can be held at the output and becomes almost constant. In this embodiment, since the opening degree of the heat utilization water flow rate control valve 19 is controlled based on the reactor heat output difference, at least during rated output operation, the flow rate of the extracted steam supplied to the heaters 21A and 21B is increased or decreased. Regardless, the temperature of the hot water supplied from the hot water generator 17 to the hot water utilization facility through the hot water supply pipe 18 can be maintained at a predetermined temperature.

  In the present embodiment, hot water using the reactor heat output of the boiling water nuclear power plant 2 by controlling the respective opening degrees of the heat utilization water flow rate control valve 19 and the heating steam flow rate control valves 23A, 23B. It is possible to adjust the ratio between the heat utilization amount in the generator 17 and the power generation amount in the boiling water nuclear power plant 2. As a result, the fluctuation of the reactor heat output of the boiling water nuclear power plant 2 can be absorbed by the heat utilization in the hot water generator 17, and the generator output in the boiling water nuclear power plant 2 is as described above. , Can be held at a predetermined output.

  In this embodiment, a part of the warmed cooling water discharged from the heat transfer pipe 8 in the condenser 7 to the sea (or river) through the cooling water discharge pipe 10 is heated by the hot water supply pipe 17. Since the hot water is generated from the cooling water, a part of the exhaust heat of the boiling water nuclear power plant 2 discharged from the condenser 7 to the outside is used as the hot water in the hot water generator 17. It can be effectively used for generation. Therefore, the thermal efficiency of the whole system in the cogeneration nuclear power generation system 1 is improved.

  In a conventional boiling water nuclear power plant, when the reactor heat output decreases from the rated output during rated power operation, the core flow rate is constantly increased to maintain the reactor heat output at the rated output. On the other hand, in the present embodiment, during the rated power operation, the increase in the reactor thermal output due to the increase in the core flow rate is performed, for example, once a week. For this reason, in this embodiment, control of the reactor power is simplified, and the operability of the boiling water nuclear power plant is improved. As a result, the reactor power control system can be simplified.

  From the viewpoint of improving the overall efficiency of the cogeneration nuclear power generation system 1 in the present embodiment, the number of stages of the heaters 21 provided in the hot water supply pipe 18 in the hot water generator 17 is increased as much as possible and supplied to the heaters 21. As extraction steam for heating, it is better to use steam as low as possible (steam extracted from a position close to the blade of the final stage of the low-pressure turbine 6). For example, in order to raise the temperature to the target temperature (for example, 150 ° C.) of the hot water used in the hot water utilization facility, it is necessary to use high-temperature and high-pressure steam at 150 ° C. or higher. In this case, first, in order to raise the temperature of the hot water to 100 ° C., the water flowing in the hot water supply pipe 18 is heated with low-temperature steam, and then the temperature of the hot water is raised from 100 ° C. to 150 ° C. Therefore, it may be further heated with steam having a high temperature. In this way, the hot water generator 17 is provided with a plurality of stages of heaters 21, and the heater 21 (for example, the heater 21A) disposed in the region where the temperature of the hot water is low uses low-temperature and low-pressure steam, thereby boiling water. In the process of decreasing the temperature and pressure of steam in the nuclear power plant 2, the energy can be recovered by the high-pressure turbine 5 and the low-pressure turbine 6 and used for power generation. For this reason, the efficiency of the whole system in the cogeneration nuclear power generation system 1 improves. Considering this, from the viewpoint of improving system efficiency, it is desirable to extract at least a part of the heating steam supplied to the heater 21 from at least the low-pressure turbine 6.

  From the viewpoint of increasing the number of stages of the heater 21 as much as possible, what is the extraction point of each extraction steam supplied to each of the low-pressure feed water heater 12 and the high-pressure feed water heater 14 (installed in the low-pressure turbine 6 and the high-pressure turbine 5 respectively)? It is possible to newly install another bleed point. However, the installation of another extraction point has a great impact on the increase in equipment costs. In order to avoid an increase in equipment cost, the heating steam supply pipe 22A is connected to the extraction steam pipe 16, the heating steam supply pipe 22B is connected to the extraction steam pipe 15, and the extraction points provided in the high-pressure turbine 5 and the low-pressure turbine 6 are connected. Should be reduced. Furthermore, the temperature rise width of the feed water per feed water heater provided in the feed water pipe 11 is generally 55 ° C. or less. For this reason, the temperature rise width of the warm water guided by the warm water supply pipe 18 per one stage of the plurality of heaters 21 provided in the warm water supply pipe 18 in the warm water generation apparatus 17 may be 55 ° C. or less. Thereby, the raise in equipment cost can be suppressed, improving the stem efficiency in the cogeneration nuclear power generation system 1. FIG.

  In this embodiment, the opening degree of the heat utilization water flow rate control valve 19 is controlled, but instead of opening degree control of the heat utilization water flow rate control valve 19, the rotation speed of the pump 20 provided in the hot water supply pipe 18 is changed. Control may be performed using the reactor heat output measured by the reactor heat output monitor 25. In this case, the rotational speed of the pump 20 is controlled so as to increase as the reactor heat output increases. Moreover, you may control both the opening degree of the heat utilization water flow control valve 19 and the rotation speed of the pump 29. FIG.

  A cogeneration nuclear power generation system according to embodiment 2, which is another embodiment of the present invention, will be described with reference to FIG. The cogeneration nuclear power generation system 1A of the present embodiment has a configuration in which the reactor thermal output monitor 25 in the cogeneration nuclear power generation system 1 of the first embodiment is replaced with a reactor thermal output monitor 25A. Other configurations of the cogeneration nuclear power generation system 1A are the same as those of the cogeneration nuclear power generation system 1. The reactor thermal output monitor 25 </ b> A is connected to a plurality of neutron detectors 30 installed in the core 26 in the reactor 3 of the boiling water nuclear plant 2.

  Each neutron detector 30 installed in the core 26 detects a neutron flux generated by fission of nuclear fuel material contained in a plurality of fuel assemblies loaded in the core 26. Each neutron flux detected by each neutron detector 30 is input to the reactor thermal output monitor 25A, and the reactor thermal output monitor 25A obtains the reactor thermal output using each neutron flux. In this way, the reactor thermal output monitor 25A also measures the reactor thermal output.

  In the present embodiment, power generation by the boiling water nuclear power plant 2 and hot water generation by the hot water generator 17 are performed in the same manner as in the first embodiment. In the present embodiment, the control device 24 opens each of the flow control valves 19, 23A, and 23B based on the reactor thermal power input from the reactor thermal power monitor 25A, as in the control device 24 in the first embodiment. Control the degree.

  Also in this embodiment, each effect produced in the first embodiment can be obtained.

  A cogeneration nuclear power generation system according to embodiment 3, which is another embodiment of the present invention, will be described with reference to FIG. The cogeneration nuclear power generation system 1B of the present embodiment has a configuration in which the boiling water nuclear power plant 2 is replaced with a pressurized water nuclear power plant 31 in the cogeneration nuclear power generation system 1 of the first embodiment. Other configurations of the co-heat nuclear power generation system 1B are the same as those of the co-heat nuclear power generation system 1.

  The pressurized water nuclear power plant 31 has a configuration in which, in the boiling water nuclear power plant 2 used in the first embodiment, the reactor 3 as a steam generator is replaced with a reactor 3A and a steam generator 32 as a steam generator. Have The steam generator 32 includes a plurality of heat transfer tubes (not shown) therein. A cooling water discharge nozzle (not shown) provided in the nuclear reactor 3 </ b> A having the core 26 is connected to the shell side inlet portion in the steam generator 32 by the primary cooling system pipe 33. The return pipe portion of the primary cooling system pipe 33 connected to the shell-side outlet in the steam generator 32 is connected to a cooling water inflow nozzle (not shown) provided in the nuclear reactor 3A. The water supply pipe 11 is connected to the inlet side end portions of a plurality of heat transfer tubes (not shown) installed in the steam generator 32, and the main steam pipe 4 is the outlet side end of those heat transfer tubes in the steam generator 32. Contacted the department. Other configurations of the pressurized water nuclear power plant 31 are the same as those of the boiling water nuclear power plant 2.

  The generation of steam in the pressurized water nuclear power plant 31 will be described. The cooling water is heated to a high temperature by the heat generated by the fission of the nuclear fuel material contained in the fuel assembly loaded in the core 26. This high-temperature cooling water is supplied to the shell side of the steam generator 32 through the primary piping 33. The feed water supplied to each heat transfer tube of the steam generator 32 by the feed water pipe 11 is heated by the high-temperature cooling water flowing through the shell side to become steam. The generated steam is supplied to the high-pressure turbine 5 and the low-pressure turbine 6 through the main steam pipe 4 to rotate these turbines. The shell-side cooling water whose temperature has been lowered by heating the feed water in the heat transfer tubes is returned to the reactor 3 </ b> A through the return pipe portion of the primary cooling system piping 33.

  Also in the present embodiment, as in the first embodiment, a part of the cooling water in the cooling water discharge pipe 10 is guided to the hot water supply pipe 18 and is heated by the heaters 21A and 21B of the hot water generator 17 with the heating steam supply pipe. Heated by the bleed steam introduced by 22A and 22B. The opening degree of each flow control valve 19, 23A, 23B is controlled by the control device 24 as in the first embodiment.

  Also in this embodiment, each effect produced in the first embodiment can be obtained.

  A cogeneration nuclear power generation system according to embodiment 4, which is a preferred embodiment of the present invention, will be described with reference to FIG. The cogeneration nuclear power generation system 1 </ b> C of the present embodiment includes the boiling water nuclear power plant 2 and the hot water generator 17, as in the first embodiment. In the cogeneration nuclear power generation system 1C of the present embodiment, in the cogeneration nuclear power generation system 1 of the present embodiment, the reactor heat output monitor 25 is replaced with a power generation amount request signal output device 34, and the control device 24 is replaced with a control device 24A. It has a configuration. Other configurations of the cogeneration nuclear power generation system 1C are the same as those of the cogeneration nuclear power generation system 1. Power generation by the boiling water nuclear power plant 2 and hot water generation by the hot water generator 17 are performed in the same manner as in the first embodiment.

  The power generation amount request signal output device 34 stores a power generation amount request signal indicating a set value of the power generation amount set by the operator in consideration of load fluctuations in the power system. The control device 24A receives the power generation amount request signal set in the power generation amount request signal output device 34, and based on the power generation amount request signal, the opening degree of the heat utilization water flow rate control valve 19 and the heating steam flow rate control valve 23A, Each opening degree of 23B is controlled. In the present embodiment, the opening degree of the heat-utilizing water flow rate control valve 19 and the opening degrees of the heating steam flow rate control valves 23A and 23B are controlled in accordance with load fluctuations of the power system.

  FIG. 9 shows an example of changes in the reactor heat output and the generator output corresponding to load fluctuations when this embodiment is used. The generator output basically matches the power generation request signal. In a conventional boiling water nuclear power plant, as shown in FIG. 4, the reactor heat output is changed in accordance with the change in the generator output. However, in the present embodiment, in the boiling water nuclear power plant 2, the reactor heat output is not controlled, and only the generator output is changed. The reactor heat output changes with time in FIG. 4, but is substantially constant in FIG. This is because the load fluctuation is a fluctuation with a short period of one day, and thus the reactor heat output can be regarded as a substantially constant value in the short period shown in FIG.

  Furthermore, information related to the four characteristics shown in FIG. 10 is stored in the storage device of the control device 24A. The fifth characteristic is a characteristic indicating the relationship between the power generation requirement and the flow rate of the heat water. The sixth characteristic is a characteristic indicating the relationship between the heat utilization water flow rate and the valve opening, the seventh characteristic is a characteristic indicating the relationship between the power generation requirement and the heating steam flow rate, and the eighth characteristic is the heating steam flow rate and the valve. It is the characteristic which shows the relationship of an opening degree. As the power generation requirement increases, the flow rates of heat utilization water and heating steam respectively decrease. When the flow rate of the heat utilization water increases, the opening degree of the heat utilization water flow rate control valve 19 increases, and when the flow rate of the heating steam increases, the respective opening amounts of the heating steam flow rate control valves 23A and 23B increase. .

  Control of the opening degree of the heat utilization water flow rate control valve 19 and the opening degree of the heating steam flow rate control valves 23A and 23B by the control device 24A will be described with reference to FIG.

  In step S1A, the power generation amount request (power generation amount request signal) input from the power generation amount request signal output device 34 is subtracted from the reference power generation amount request to obtain a difference in the power generation amount request from the reference power generation amount request. Using the fifth characteristic, the flow rate of the heat utilization water with respect to the difference in the power generation request (the power generation request on the horizontal axis of the fifth characteristic) is obtained. In step S2A, the opening degree of the heat utilization water flow rate control valve 19 with respect to the flow rate of heat utilization water obtained in step S1A is obtained using the sixth characteristic. The control device 24A controls the opening degree of the heat using water flow rate control valve 19 based on the opening degree obtained in step S2A.

  In step S3A, the difference between the power generation amount request and the reference power generation amount request is obtained by subtracting the power generation amount request (power generation amount request signal) input from the power generation amount request signal output device 34 from the reference power generation amount request. Using the seventh characteristic, the heating steam flow rate with respect to the difference in the power generation request (the power generation request on the horizontal axis of the third characteristic) is obtained. In step S4A, the opening degree of each of the heating steam flow rate control valves 23A and 23B with respect to the heating steam flow rate obtained in step S3A is obtained using the eighth characteristic. The control device 24A controls the respective opening degrees of the heating steam flow rate control valves 23A and 23B based on the respective opening degrees obtained in step S4A.

  As the power generation requirement increases, it is necessary to increase the flow rate of steam flowing in the high-pressure turbine 4 and the low-pressure turbine 5. For this reason, in this embodiment, when the power generation requirement increases, the control device 24A uses the seventh characteristic in step S3A to reduce the heating steam flow rate, and uses the eighth characteristic in step S4A. Each opening degree of the heating steam flow rate control valves 23A and 23B corresponding to the small amount of the heating steam flow rate is reduced. As a result, the amount of extracted steam supplied to the heaters 21A and 21B by the heated steam supply pipes 22A and 22B decreases, and the flow rate of steam flowing through the high-pressure turbine 4 and the low-pressure turbine 5 increases. Even if the reactor heat output is constant, the flow rate of the steam flowing through the high-pressure turbine 4 and the low-pressure turbine 5 is increased, so that the generator output can be increased in response to the increasing power generation requirement.

  Further, when the power generation demand decreases, the control device 24A controls the heating steam flow rate control valves 23A and 23B so that the respective opening degrees are increased, and the extracted steam supplied to the heaters 21A and 21B, respectively. Increase the amount. Thereby, even if the reactor heat output is constant, the flow rate of the steam flowing through the high-pressure turbine 4 and the low-pressure turbine 5 is reduced, and the generator output can be reduced in response to the decreasing power generation requirement.

  In the present embodiment, even when the power generation request in consideration of the load fluctuation of the power system fluctuates, it can be absorbed by the heat utilization in the hot water generation device 17, and the generator output is easily adjusted according to the power generation amount request. be able to.

  In the present embodiment, hot water using the reactor heat output of the boiling water nuclear power plant 2 by controlling the respective opening degrees of the heat utilization water flow rate control valve 19 and the heating steam flow rate control valves 23A, 23B. It is possible to adjust the ratio between the heat utilization amount in the generator 17 and the power generation amount in the boiling water nuclear power plant 2. As a result, the load fluctuation of the electric power system can be absorbed by the heat utilization in the hot water generator 17, and the generator output in the boiling water nuclear power plant 2 is adjusted according to the power generation requirement as described above. be able to.

  When the respective opening degrees of the steam flow control valves for heating 23A and 23B are reduced by the control device 24A, the flow rate of the extracted steam supplied to the heaters 21A and 21B is reduced, and hot water is supplied from the hot water generator 17 The temperature of the hot water supplied to the hot water utilization facility through the pipe 18 is lowered. In order to suppress the temperature drop of the hot water, the control device 24A decreases the flow rate of the cooling water flowing into the hot water supply pipe 18 from the cooling water discharge pipe 10 by reducing the opening degree of the heat utilization water flow control valve 19. The temperature drop of the hot water supplied to the hot water use facility through the hot water supply pipe 18 is suppressed. As a result, the hot water temperature is maintained at a predetermined temperature.

  Also in the present embodiment, a part of warmed cooling water discharged from the heat transfer pipe 8 in the condenser 7 through the cooling water discharge pipe 10 to the sea (or river) is heated by the hot water supply pipe 18 to generate the hot water generator 17. Since the hot water is generated from this cooling water, the thermal efficiency of the entire system in the cogeneration nuclear power generation system 1C is improved as in the first embodiment. Further, similarly to the first embodiment, the control of the reactor power is simplified, and the drivability of the boiling water nuclear power plant is improved.

  In the present embodiment, each effect produced in the first embodiment can be obtained.

  In the present embodiment, similarly to the first embodiment, the rotational speed of the pump 20 may be controlled instead of controlling the respective opening degrees of the steam flow control valves 23A and 23B for heating.

  A cogeneration nuclear power generation system according to embodiment 5, which is another embodiment of the present invention, will be described with reference to FIG. The cogeneration nuclear power generation system 1D of the present embodiment has a configuration in which the boiling water nuclear power plant 2 is replaced with a pressurized water nuclear power plant 31 in the cogeneration nuclear power generation system 1C of the fourth embodiment. The other configuration of the cogeneration nuclear power generation system 1D is the same as that of the cogeneration nuclear power generation system 1C.

  In the present embodiment, each effect produced in the fourth embodiment can be obtained.

  1, 1A, 1B, 1C, 1D ... co-generation nuclear power generation system, 2 ... boiling water nuclear power plant, 3, 3A ... nuclear reactor, 4 ... main steam piping, 5 ... high pressure turbine, 6 ... low pressure turbine, 7 ... Condenser, 8 ... Heat transfer pipe, 10 ... Cooling water discharge pipe, 11 ... Water supply pipe, 17 ... Hot water generator, 18 ... Hot water supply pipe, 19 ... Heat utilization water flow control valve, 20 ... Pump, 21A, 21B ... Heater, 22A, 22B ... heating steam supply pipe, 23A, 23B ... heating steam flow rate control valve, 24, 24A ... control device 24, 25A ... reactor thermal output monitor, 26 ... core, 32 ... steam generator, 34 ... Power generation request signal output device.

Claims (7)

Reactor, steam generated in the reactor is supplied turbines to rotate the generator, condenser for condensing steam exhausted from the turbine, and the water supply pipe connecting the reactor and the condenser A nuclear power plant , a hot water generator, and a controller,
A hot water generator is provided in the first pipe connected to the first pipe for guiding water, a plurality of second pipes for guiding the steam extracted from the turbine, and the second pipe. Provided for each of the plurality of second pipes, a plurality of heating devices for heating the water guided by the pipes with the steam supplied by the second pipe, a first flow rate control valve provided in the first pipe, A plurality of second flow rate control valves,
The flow rate of the water flowing in the first pipe is obtained based on the difference between the measured reactor thermal output and the reference reactor thermal output, and the opening degree of the first flow control valve is obtained based on the flow rate of the water. Controlling the opening of the first flow rate control valve based on the determined opening, determining the flow rate of steam flowing in the second pipe based on the difference, and determining the first flow rate based on the flow rate of the steam. 2. A cogeneration nuclear power generation comprising the control device that obtains the respective opening amounts of the two flow rate control valves and controls the respective opening amounts of the second flow rate control valves based on the obtained opening amounts. system. Reactor, steam generated in the reactor is supplied turbines to rotate the generator, condenser for condensing steam exhausted from the turbine, and the water supply pipe connecting the reactor and the condenser A nuclear power plant , a hot water generator, and a controller,
A hot water generator is provided in the first pipe connected to the first pipe for guiding water, a plurality of second pipes for guiding the steam extracted from the turbine, and the second pipe. Provided for each of the plurality of second pipes, a plurality of heating devices for heating the water guided by the pipes with the steam supplied by the second pipe, a first flow rate control valve provided in the second pipe, A plurality of second flow rate control valves,
The flow rate of the water flowing in the first pipe is obtained based on the difference between the reference power generation amount request signal and the power generation amount request signal output from the power generation amount request signal output device, and the first flow rate is calculated based on the flow rate of the water. Obtaining an opening of the control valve, obtaining an opening of the first flow control valve based on the flow rate of the water, controlling an opening of the first flow control valve based on the obtained opening; The flow rate of the steam flowing through the second pipe is determined based on the difference, the respective opening amounts of the second flow rate control valves are determined based on the flow rate of the steam, and the first flow rate is determined based on the determined opening amounts. A cogeneration nuclear power generation system comprising the control device for controlling the opening degree of each of the two flow rate control valves . A reactor, a steam generator to which a coolant heated in the reactor is supplied, a steam to be generated by the steam generator, a turbine that rotates a generator, and a condenser that condenses the steam exhausted from the turbine A nuclear power plant having a water supply, a water supply pipe connecting the condenser and the steam generator, a hot water generation device, and a control device;
A hot water generator is provided in the first pipe connected to the first pipe for guiding water, a plurality of second pipes for guiding the steam extracted from the turbine, and the second pipe. Provided for each of the plurality of second pipes, a plurality of heating devices for heating the water guided by the pipes with the steam supplied by the second pipe, a first flow rate control valve provided in the second pipe, A plurality of second flow rate control valves,
The flow rate of the water flowing in the first pipe is obtained based on the difference between the measured reactor thermal output and the reference reactor thermal output, and the opening degree of the first flow control valve is obtained based on the flow rate of the water. Controlling the opening of the first flow rate control valve based on the determined opening, determining the flow rate of steam flowing in the second pipe based on the difference, and determining the first flow rate based on the flow rate of the steam. 2. A cogeneration nuclear power generation comprising the control device that obtains the respective opening amounts of the two flow rate control valves and controls the respective opening amounts of the second flow rate control valves based on the obtained opening amounts. system. A reactor, a steam generator to which a coolant heated in the reactor is supplied, a steam to be generated by the steam generator, a turbine that rotates a generator, and a condenser that condenses the steam exhausted from the turbine A nuclear power plant having a water supply, a water supply pipe connecting the condenser and the steam generator, a hot water generation device, and a control device;
A hot water generator is provided in the first pipe connected to the first pipe for guiding water, a plurality of second pipes for guiding the steam extracted from the turbine, and the second pipe. Provided for each of the plurality of second pipes, a plurality of heating devices for heating the water guided by the pipes with the steam supplied by the second pipe, a first flow rate control valve provided in the second pipe, A plurality of second flow rate control valves,
The flow rate of the water flowing in the first pipe is obtained based on the difference between the reference power generation amount request signal and the power generation amount request signal output from the power generation amount request signal output device, and the first flow rate is calculated based on the flow rate of the water. Obtaining an opening of the control valve, obtaining an opening of the first flow control valve based on the flow rate of the water, controlling an opening of the first flow control valve based on the obtained opening; The flow rate of the steam flowing through the second pipe is determined based on the difference, the respective opening amounts of the second flow rate control valves are determined based on the flow rate of the steam, and the first flow rate is determined based on the determined opening amounts. A cogeneration nuclear power generation system comprising the control device for controlling the opening degree of each of the two flow rate control valves. The cogeneration nuclear power generation system according to any one of claims 1 to 4 , wherein the turbine includes a high-pressure turbine and a low-pressure turbine, and the plurality of second pipes are connected to the low-pressure turbine. The turbine includes a high-pressure turbine and a low-pressure turbine, the plurality of heating devices include a first heating device and a second heating device, and the first heating device is more than the second heating device in the first pipe. Are also arranged upstream, the plurality of second pipes are connected to the low-pressure turbine and connected to the first heating device, and the third pipe is connected to the high-pressure turbine and connected to the second heating device. The cogeneration nuclear power generation system according to any one of claims 1 to 4 , comprising a fourth pipe. The condenser is an inlet connected to the cooling water supply pipe outlet has a connected heat transfer tubes to the cooling water discharge pipe, to the first pipe claims 1 and is connected to the cooling water discharge pipe 5. The cogeneration nuclear power generation system according to any one of 4 above.

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