JP2008304264A - Nuclear power plant and its operation method - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To increase output of a nuclear power plant without changing constitution of plant equipment in the nuclear power plant. <P>SOLUTION: In an operation method of the nuclear power plant 10, assuming that a period from start of a reactor 11 until operation stop of the reactor for fuel exchange is one operation period, the thermal output of the reactor in the (N+1)-th operation cycle is increased furthermore than the thermal output of the reactor in the N-th operation cycle before the (N+1)-th operation cycle, and the ratio of a bleeding steam flow rate bled from a high-pressure steam system 12 by a high-pressure bleeding system and guided to a high-pressure heater 26 to a main steam flow rate at the reactor outlet is reduced in the (N+1)-th operation cycle in comparison with the case in the N-th operation cycle, and thereby the enthalpy or temperature at the reactor inlet of supply water to be supplied from a water supply system 14 into the reactor 11 is controlled in the (N+1)-th operation cycle equally to the case in the N-th operation cycle. <P>COPYRIGHT: (C)2009,JPO&INPIT

Description

  The present invention relates to a nuclear power plant and a method for operating the nuclear power plant, and more particularly to a nuclear power plant suitable for increasing power generation capacity by increasing a reactor heat output and a method for operating the nuclear power plant.

In a conventional new nuclear power plant, as described in, for example, Patent Document 1, the mass flow rate of main steam at the core outlet is improved by improving the fuel configuration or the shape configuration of the fuel assembly in order to increase the electric output. (Hereinafter, “mass flow rate” is simply referred to as “flow rate” in the present specification), and it has been proposed to increase the electrical output.
JP-A-9-264983

  When the above-described conventional technique is applied to an existing nuclear power plant, the feed water temperature rises incidentally as the heat output increases. At this time, the enthalpy of the feed water also rises at the same time, so the amount of enthalpy required to boil by receiving heat in the reactor or steam generator, that is, the difference in enthalpy between the inlet and outlet of the reactor or steam generator is reduced. To do.

  In general, the thermal output Q of the reactor, the enthalpy difference H at the inlet and outlet of the reactor or steam generator, and the main steam flow generated in the reactor or steam generator (feed water flow to the reactor or steam generator) ) Since there is a relationship of Q = H × G with G, the enthalpy difference H at the inlet and outlet of the reactor or steam generator decreases at the time of increased output, so that the reactor or steam generator When the heat output Q of the nuclear reactor is increased, the main steam flow rate or the feed water flow rate increases at a rate greater than the increase amount of the thermal output Q. This increase in main steam flow or feed water flow is caused by almost all plant equipment such as feed water pipes, feed water heaters, feed water pumps, steam dryers and other in-furnace structures, main steam pipes, high pressure turbines, low pressure turbines and condensers. The design margin will be reduced.

  One method for suppressing this is to decrease the feed water temperature by reducing the flow rate of the extracted steam to the feed water heater. However, if the extraction steam flow rate for the feed water heater is simply reduced as a whole, the amount of steam in the low-pressure turbine is insufficient, the thermal efficiency is greatly deteriorated, and the electrical output hardly increases, which is not realistic. In addition, drastically reducing the flow rate of the extracted steam, particularly in a low-pressure turbine designed on the premise of upstream extraction, increases the extreme flow rate and increases the possibility of reducing the design margin. .

  In nuclear power plants using ordinary boiling water reactors, steam turbines are one of the plant equipment where the reduction in design margin due to the increase in the main steam flow may cause the most trouble when planning to increase heat output. There is. In nuclear power generation systems other than boiling water reactors, there is a similar problem for plants with relatively small design margins for steam turbines. When applying conventional technology to existing nuclear power plants, large-scale plant equipment is required. Improvement and replacement were necessary.

  An object of the present invention is made in consideration of the above-described circumstances, and provides a nuclear power plant capable of realizing an increased output of the nuclear power plant without changing the configuration of the plant equipment in the nuclear power plant and an operation method thereof. There is.

  An operation method of a nuclear power plant according to the present invention includes a nuclear reactor, a high-pressure steam system to which steam generated in the nuclear reactor is supplied, from a reactor outlet to a low-pressure turbine inlet through a high-pressure turbine, and from the low-pressure turbine inlet A low-pressure steam system to a condenser inlet for condensing steam discharged from the low-pressure turbine, the condenser, and a feed water heater for heating the feed water supplied from the condenser, the feed water heating And a water supply system for supplying water from the reactor to the nuclear reactor, and a period from the start of the nuclear reactor until the operation of the nuclear reactor is stopped for fuel replacement in one operating cycle. The reactor heat output in the subsequent operation cycle is increased from the reactor heat output in the previous operation cycle before the subsequent operation cycle, and extracted from the high-pressure steam system. The ratio of the mass flow rate of the extracted steam leading to the feed water heater to the mass flow rate of the main steam at the reactor outlet is reduced in the subsequent operation cycle compared to the previous operation cycle, thereby allowing the water supply system to The enthalpy or temperature at the reactor inlet of the feed water supplied to the reactor is controlled in the subsequent operation cycle in the same manner as the previous operation cycle.

  The nuclear power plant operating method according to the present invention includes a reactor, a steam system including a high-pressure turbine and a low-pressure turbine to which steam generated in the reactor is supplied, and condensing steam discharged from the low-pressure turbine. A condenser, a low-pressure feed water heater that is installed downstream of the condenser and upstream of the main feed pump and that heats feed water supplied from the condenser, and downstream of the main feed pump And a high-pressure feed water heater installed upstream from the nuclear reactor, and a feed water system for supplying feed water from the high-pressure feed water heater to the nuclear reactor. When the period from the start-up of the reactor until the operation of the reactor is stopped for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is The temperature of the high-pressure feed water heater is increased in the subsequent operation cycle, and the amount of temperature increase in the later operation cycle is smaller than that in the previous operation cycle. It is characterized in that the enthalpy or temperature at the reactor inlet of supplied water is controlled in the subsequent operation cycle in the same manner as in the previous operation cycle.

  The nuclear power plant operating method according to the present invention further comprises a nuclear reactor, a steam system including a high-pressure turbine and a low-pressure turbine to which steam generated in the nuclear reactor is supplied, and condensing the steam discharged from the low-pressure turbine. And a feed water heater for heating the feed water supplied from the condenser, a feed water system for feeding the feed water from the feed water heater to the reactor, and extracting steam from the high pressure turbine side And an extraction steam system having at least one high-pressure extraction pipe leading to the feed water heater, and the operation of the nuclear reactor is stopped for fuel replacement from the start-up of the nuclear reactor. When the period up to is one operation cycle, the reactor heat output in the subsequent operation cycle is increased from the reactor heat output in the previous operation cycle before the subsequent operation cycle. In the subsequent operation cycle, the extraction steam is stopped or the mass flow rate of the extraction steam is adjusted to be equal to or less than the predetermined flow rate in at least one system of the high-pressure extraction pipe in which the extraction steam was flowing at the predetermined flow rate in the previous operation cycle. Thus, the enthalpy or temperature at the reactor inlet of the feed water supplied from the feed water system to the reactor is controlled in the later operation cycle in the same manner as the previous operation cycle.

  A nuclear power plant according to the present invention includes a nuclear reactor, a steam system including a high-pressure turbine and a low-pressure turbine to which steam generated in the nuclear reactor is supplied, and condensate for condensing the steam discharged from the low-pressure turbine. And a feed water heater that heats feed water supplied from the condenser, a feed water system that feeds feed water from the feed water heater to the reactor, and steam is extracted from the high-pressure turbine side to extract the steam An extraction steam system having at least one high-pressure extraction pipe leading to the feed water heater, and an extraction steam adjusting means provided in the extraction steam system for adjusting the extraction steam. Therefore, when the period until the operation of the reactor is stopped is defined as one operation cycle, the nuclear reactor heat output in the subsequent operation cycle is changed to the atom in the operation cycle prior to the subsequent operation cycle. When the heat output is increased, the extraction steam is adjusted by the extraction steam adjusting means, and the enthalpy or temperature at the reactor inlet of the feed water supplied from the feed water system to the reactor is changed in a later operation cycle. It is characterized by being configured to control in the same manner as the previous operation cycle.

  According to the nuclear power plant and the operation method thereof according to the present invention, the enthalpy or temperature at the reactor inlet of feed water supplied from the feed water system to the reactor, or the steam generator of feed water supplied from the feed water system to the steam generator Since the enthalpy or temperature at the inlet is controlled to be equal to that in the previous operation cycle in the subsequent operation cycle, the enthalpy difference between the inlet and the outlet of the reactor or the steam generator can be kept constant. For this reason, the rate of increase in the main steam flow rate or feed water flow rate in the reactor or steam generator is higher than that in the previous operation cycle in the subsequent operation cycle. It can be made equal to the increasing rate of increase. As a result, the increase in the main steam flow rate or feed water flow rate in the reactor or steam generator can be suppressed against the increase in reactor heat output, so the steam turbine, steam control valve, feed water pipe, main steam pipe, The design margin of the plant equipment such as a structure can be suitably maintained, and the increased output of the nuclear power plant can be realized without changing the configuration of these plant equipment.

  The best mode for carrying out the present invention will be described below with reference to the drawings. However, the present invention is not limited to these embodiments.

[A] First embodiment (FIGS. 1 to 5)
FIG. 1 is a system configuration diagram of a nuclear power plant showing a configuration at the time of increased output in the first embodiment in which the operation method of the nuclear power plant according to the present invention is applied to a nuclear power plant having a boiling water reactor. is there. FIG. 2 is a system configuration diagram showing the configuration of the nuclear power plant of FIG.

  A nuclear power plant 10 shown in FIGS. 1 and 2 is a direct cycle plant equipped with a boiling water reactor, and includes a nuclear reactor 11, a high-pressure steam system 12, a low-pressure steam system 13, a feed water system 14, and a high-pressure extraction system 15. And a low-pressure extraction system 16. The nuclear reactor 11 is a boiling water reactor (light water reactor), and a reactor core and a jet pump are provided in the reactor pressure vessel 19 and a recirculation pump is provided outside the reactor pressure vessel 19, and coolant (water) is heated by the reactor core. And generate steam.

  The high-pressure steam system 12 is a system from the outlet of the reactor 11 to the inlet of the low-pressure turbine 21 through the high-pressure turbine 20, and the steam generated in the reactor 11 is high-pressure via the main steam pipe 22 and the steam control valve 23. Power is supplied to the turbine 20. The high-pressure steam system 12 includes any one of a moisture separator, a moisture separation heater, and a moisture separation reheater between the high-pressure turbine 20 and the low-pressure turbine 21. In the present embodiment, a moisture separator 32 is provided. The low-pressure steam system 13 is a system from the inlet of the low-pressure turbine 21 to the inlet of the condenser 25, and introduces steam discharged from the high-pressure turbine 20 into the low-pressure turbine 21 to obtain power.

  The feed water system 14 includes a condenser 25, a low pressure feed water heater 27, a high pressure feed water heater 26, and a main feed water pump 28. The condenser 25 condenses the steam discharged from the low pressure turbine 21. The low-pressure feed water heater 27 is disposed downstream of the condenser 25 and upstream of the main feed pump 28, and heats the feed water from the condenser 25 by the extracted steam from the low-pressure extraction system 16. The high-pressure feed water heater 26 is installed downstream of the main feed water pump 28 and upstream of the reactor 11, and the feed water from the low-pressure feed water heater 27 is further supplied by the extracted steam from the high-pressure extraction system 15. Heat. The main feed water pump 28 supplies the feed water heated by the high pressure feed water heater 26 and the low pressure feed water heater 27 to the reactor 11 through the feed water pipe 29.

  The high-pressure extraction system 15 includes at least one system (one) of high-pressure extraction pipes 30 that extract steam from the high-pressure turbine 20 side, that is, in the middle of the high-pressure turbine 20 or from the outlet, and guide the steam to the high-pressure feed water heater 26. Here, the outlet of the high-pressure turbine 20 means the interval from the outlet of the high-pressure turbine 20 to the inlet of the moisture separator 32, the moisture separation heater, or the moisture separation reheater. Further, the low-pressure extraction system 16 extracts steam from the low-pressure turbine 21 side, for example, in the middle of the low-pressure turbine 21, the moisture separator 32, the moisture separator / heater or the moisture separator / reheater, and the low-pressure feed heater 27. The low-pressure bleed pipe 31 leading to is provided with at least one system (one).

  By the way, in the nuclear power plant 10 (FIG. 1, FIG. 2) comprised as mentioned above, when implementing the heat output increase operation of the reactor 11, the feed water flow volume supplied to the reactor 11 from the feed water system 14 is increased. At the same time, as shown in FIG. 1, flow rate adjustment valves 17 and 18 as extraction steam adjustment means and flow rate adjustment means are installed in the high pressure extraction pipe 30 and the low pressure extraction pipe 31, respectively. The flow rate adjusting valve 17 adjusts the flow rate of the bleed steam extracted from the high pressure bleed system 15 and introduced into the high pressure feed water heater 26. Further, the flow rate adjustment valve 18 adjusts the flow rate of the extracted steam extracted from the low pressure extraction system 16 and introduced into the low pressure feed water heater 27.

  The extraction steam adjustment means is not limited to the flow adjustment valves 17 and 18, but may be a flow adjustment means such as an office. Further, in each of the high pressure extraction pipe 30 and the low pressure extraction pipe 31, at least in one system. It may be a closing valve or a closing plug (plugging) as closing means for stopping the flow of the extracted steam.

  That is, the period from the start of the reactor 11 to the stop of the operation of the reactor 11 for fuel replacement is defined as one operation period, and the reactor heat output in the (N + 1) th operation cycle, which is the subsequent operation cycle, The reactor heat output is increased in the Nth operation cycle, which is the previous operation cycle before the (N + 1) th operation cycle. At this time, (a) the extraction steam introduced from the high-pressure extraction system 15 to the high-pressure feed water heater 26 is adjusted by the flow rate adjusting valve 17. Alternatively, (b) the flow control valves 17 and 18 adjust the extraction steam introduced from the high pressure extraction system 15 to the high pressure feed water heater 26 and from the low pressure extraction system 16 to the low pressure feed water heater 27, respectively. According to one of these (a) and (b), the enthalpy or temperature at the inlet of the feed water supplied from the feed water system 14 to the reactor 11 is changed to the Nth operation cycle in the (N + 1) th operation cycle. Control equally. Here, it is assumed that the Nth operation cycle is, for example, a rated operation with a reactor heat output of 100%.

  In the case of (a), for example, in the (N + 1) th operation cycle for increasing the reactor heat output, at least one system of the high-pressure extraction pipe 30 in the high-pressure extraction system 15 in which the extraction steam flows at a predetermined flow rate in the N-th operation cycle. Thus, the flow rate of the extracted steam is adjusted to a predetermined flow rate or less by the flow rate adjusting valve 17, or the flow of the extracted steam is stopped by a closing valve or the like. As a result, the ratio of the extraction steam flow extracted from the high-pressure extraction system 15 and introduced into the high-pressure feed water heater 26 to the main steam flow at the reactor 11 outlet is changed to the Nth operation cycle in the (N + 1) th operation cycle. Compared to decrease. Thereby, the temperature rise amount of the feed water in the high-pressure feed water heater 26 is smaller in the (N + 1) th operation cycle than in the Nth operation cycle. As a result, it becomes possible to control the enthalpy or temperature at the inlet of the reactor 11 of the feed water supplied from the feed water system 14 to the reactor 11 in the (N + 1) th operation cycle, equivalent to the Nth operation cycle.

  In the case of (b), for example, in the (N + 1) th operation cycle for increasing the reactor heat output, at least the high-pressure extraction pipe 30 in the high-pressure extraction system 15 in which the extraction steam flows at a predetermined flow rate in the Nth operation cycle. In one system, the flow rate of the extracted steam is adjusted to a predetermined flow rate or less by the flow rate adjusting valve 17, or the flow of the extracted steam is stopped by a closing valve or the like. Further, in this (N + 1) th operation cycle, the flow rate of the extraction steam is set to a predetermined flow rate by the flow rate adjusting valve 18 in at least one system of the low pressure extraction pipe 31 in the low pressure extraction system 16 in which the extraction gas flows at a predetermined flow rate in the Nth operation cycle. Adjust the following or stop the flow of extracted steam with a stop valve.

  Accordingly, the flow rate of the extraction steam extracted from the high-pressure extraction system 15 and led to the high-pressure feed water heater 26 and the extraction steam flow rate extracted from the low-pressure extraction system 16 and guided to the low-pressure feed water heater 27 at the respective reactor 11 outlets. The ratio to the main steam flow rate is reduced in the (N + 1) th operation cycle as compared with the Nth operation cycle. Among these, the degree of decrease in the ratio of the extraction steam flow rate from the high pressure extraction system 15 to the main steam flow rate at the outlet of the reactor 11 decreases in the (N + 1) th operation cycle as compared with the Nth operation cycle. The ratio of the extracted steam flow rate from the system 16 to the main steam flow rate at the outlet of the reactor 11 is set larger than the degree of decrease in the (N + 1) th operation cycle compared to the Nth operation cycle.

  Thereby, the total temperature increase amount of the feed water by the high-pressure feed water heater 26 and the low-pressure feed water heater 27 is smaller than the N-th operation cycle in the (N + 1) th operation cycle, and the high-pressure feed water heater for the total temperature rise amount The ratio of the temperature rise amount of the water supply by 26 is smaller in the (N + 1) th operation cycle than in the Nth operation cycle. As a result, it is possible to control the enthalpy or temperature at the inlet of the feed water supplied from the feed water system 14 to the reactor 11 in the (N + 1) th operation cycle in the same manner as the Nth operation cycle.

  Here, the control of the feed water temperature at the inlet of the feed water supplied from the feed water system 14 to the reactor 11 in the (N + 1) th operation cycle is equivalent to the Nth operation cycle. The temperature of the water supplied to the reactor 11 is controlled within a temperature range of ± 1 ° C. with respect to the temperature of the water supplied to the reactor 11 in the Nth operation cycle.

  In particular, the extraction point at which the flow rate of the extraction steam from the high-pressure extraction system 15 is reduced is the extraction point in the middle of the high-pressure turbine 20 or at the outlet of the high-pressure turbine 20. Choosing a bleed point is most effective. In this case, as described above, the flow rate adjusting valve 17 that controls the flow rate of the extracted steam may be provided to reduce the flow rate of the extracted steam. The flow rate adjusting valve 17 may be controlled by the feed water temperature monitoring device 33 so that the feed water temperature in the (N + 1) th operation cycle is the same as the feed water temperature in the Nth operation cycle (at the rated time).

  Further, at least one system of the high pressure bleed pipe 30 may be completely closed. As a method of closing, a closing valve may be installed in the middle of the high pressure bleed pipe 30 or the high pressure bleed pipe 30 may be plugged. When the high-pressure bleed pipe 30 is completely closed, the control device for the bleed steam flow rate can be eliminated and the operation control can be simplified. Whether the extraction steam flow rate is controlled or the high pressure extraction pipe 30 is completely closed depends on the thermal balance and the increased output width of the nuclear power plant 10. For example, when the increased output width is small, if one system of the high-pressure extraction pipe 30 is completely closed, the feed water temperature at the inlet of the reactor 11 is too low. In this case, the extraction steam flow rate is adjusted. Alternatively, design / planning is performed so that the feed water temperature at the inlet of the reactor 11 becomes equal to that at the rated time under the blockage of the high-pressure extraction pipe 30.

  FIG. 3 shows the relationship between the reactor heat output, the main steam flow rate (the main steam flow rate flowing from the reactor 11 to the main steam pipe 22), the extraction steam flow rate, and the operation cycle. And it is a graph shown about the case of the conventional increase output.

  In the operation cycle shown in FIG. 3, the Nth operation cycle and the (N−1) th operation cycle are operation cycles before the power increase method of the present embodiment is applied. At this time, the reactor heat output is Q = 100% (rated). The system configuration of the nuclear power plant 10 before this increased output is shown in FIG. In the (N + 1) th) operation cycle, the reactor heat output is increased by α% to Q = 100 + α%, and the system configuration of the nuclear power plant 10 at this increased output is shown in FIG. Each operation cycle shows a normal operation state, and excluding the operation state at the time of starting, stopping, transient state, and accident.

  As means for increasing the reactor heat output, the control rod extraction amount in the (N + 1) th operation cycle is made larger than that in the Nth cycle, or the coolant flow rate in the core in the (N + 1) th operation cycle is recirculated. It can be realized by increasing the number of rotations of the pump to make it larger than the Nth operation cycle or changing the type of the fuel assembly.

  Depending on the nuclear power plant 10, the extraction steam flow rate and the main steam flow rate during one operation cycle may be changed as shown in FIG. 4. Such an operation cycle reduces the temperature of the feed water supplied from the feed water system 14 to the reactor 11 by reducing the flow rate of the extracted steam in order to keep the reactor heat output constant even at the end of combustion of the fuel. Is. Also in the case of the nuclear power plant 10 having the operation cycle as shown in FIG. 4, the heat balance, the extraction steam flow rate, the main steam flow rate, the feed water heating amount, etc. are the operation cycle excluding start / stop, accident / transient event occurrence, and trial operation. Among them, the operation with the maximum main steam flow shall be compared.

  By the way, in general, when the reactor heat output is increased, in order to acquire the increased heat, the flow rate of the feed water supplied from the feed water system 14 to the reactor 11 is increased, or the reactor 11 It is necessary to expand the enthalpy difference between the coolant at the inlet and the outlet of the (reactor pressure vessel 19). In the latter case, in the conventional power increase method, the temperature and enthalpy of water supplied from the water supply system 14 to the reactor 11 at the time of increased power increase, so the main steam enthalpy at the reactor 11 outlet is increased to increase the enthalpy. The difference will be expanded. However, it is difficult to increase the main steam enthalpy at the outlet of the reactor 11 from the technical and safety aspects such as high pressure and high temperature in the reactor pressure vessel 19. The enthalpy difference between the inlet and the outlet of the coolant in FIG. 9 is rather reduced, and therefore the main steam flow rate must be increased from the equation (A) described below. When the main steam flow rate is increased in this way, the design margin of the plant equipment such as the in-furnace structure and the steam turbine decreases, so these equipment must be changed. Therefore, the former method is exclusively used to increase the water supply flow rate.

Here, the reactor heat output assumes that the main steam flow rate and the feed water flow rate are the same, and the difference in the enthalpy of the coolant at the inlet and outlet of the reactor 11 (reactor pressure vessel 19) and the feed water It is defined as the product of the flow rate or the main steam flow rate. Therefore, the reactor thermal output Q at the time of rating is as follows, where H is the enthalpy difference between the coolant at the inlet and outlet of the reactor 11 at the time of rating, and G is the feed water flow rate or main steam flow rate at the time of rating. Indicated.
[Equation 1]
Q = H × G (A)

Further, in the conventional power increase method, the increase amount ΔQ (Q>ΔQ> 0) of the heat output after the α% increase power is set to ΔH (H>ΔH> 0) as the enthalpy increase amount of the feed water after the increase power If the amount of increase in the main steam flow rate or feed water flow rate before and after the α% increase output is ΔG,
[Equation 2]
ΔQ = {(H−ΔH) × (G + ΔG)} − (H × G)
When the above equation is arranged with respect to ΔG, the following equation (B) is obtained.

[Equation 3]
ΔG = (ΔQ + ΔH × G) ÷ (H−ΔH) (B)

  Therefore, from this equation (B), it can be seen that the increase amount ΔG of the main steam flow rate or the feed water flow rate is not directly proportional to the increase amount ΔQ of the heat output, but has nonlinearity that depends on the planned design of the increase output. . Furthermore, as shown by the above formula (B) and the thin line in FIG. 5, this nonlinearity becomes more significant as the feed water enthalpy increase ΔH is larger, and from H> ΔH> 0, the main steam flow rate or feed water flow rate The rate of increase has the property of increasing more sharply with respect to the rate of increase in heat output. On the other hand, the increase amount ΔH of the water supply enthalpy changes depending on the increase amount ΔQ depending on the setting of the increase amount ΔQ of the heat output.

  On the other hand, in the present embodiment, as shown in FIG. 1, the flow rate of the feed water supplied from the feed water system 14 to the reactor 11 is increased and the reactor 11 (reactor) as shown in FIG. Pressure vessel 19) By deliberately planning and designing the coolant enthalpy at the inlet (feed water enthalpy) as at the rated time, the difference in coolant enthalpy at the inlet and outlet of the reactor 11 can be kept constant. Yes.

  In order to maintain the water supply enthalpy at the inlet of the nuclear reactor 11 equally before and after the increased output, the extraction is conducted from the high-pressure steam system 12 and the low-pressure steam system 13 and led to the feed water heaters 26 and 27 in the plan of the increased output. What is necessary is to reduce the steam flow rate. However, if the amount of extracted steam is simply reduced as a whole, the amount of steam introduced to the low-pressure turbine 21 is insufficient, the thermal efficiency is greatly reduced, and the amount of power generation cannot be increased. Therefore, the flow rate of the extracted steam from the middle of the high-pressure turbine 20 or the outlet (where the outlet of the high-pressure turbine 20 is actually between the outlet of the high-pressure turbine and the inlet of the moisture separator 32) through the high-pressure extraction system 15. The pressure is selectively reduced with respect to the low-pressure turbine 21 side. As a result, the amount of steam flowing through the low-pressure turbine 21 can be increased to increase the amount of power generation.

  Most of the extracted steam from the middle or outlet of the high-pressure turbine 20 is used in the high-pressure feed water heater 26 installed on the downstream side of the main feed water pump 28. Therefore, the power increase method according to this embodiment changes the way of viewing. And a method of controlling the amount of heating of the feed water downstream of the main feed water pump 28.

  In the case of a nuclear power plant where the flow rate of the extracted steam extracted from the middle or the outlet of the high-pressure turbine 20 is originally low, the low-pressure turbine 21 generates a low pressure to make the enthalpy of water supply to the reactor 11 equal to that at the time of rating. It is necessary to reduce the flow rate of the extracted steam extracted through the extraction system 16 and sufficiently reduce the temperature of the water supplied to the reactor 11. Even when this embodiment is applied to such a nuclear power plant, the flow rate of the bleed steam from the middle or outlet of the high pressure turbine 20 is reduced more than the flow rate of the bleed steam from the low pressure turbine 21 side. Some effect can be obtained.

FIG. 5 shows changes in various quantities related to the heat balance around the reactor 11 and changes in characteristics due to the increased output of this embodiment. The result of application of the present embodiment is a change in various quantities indicated by a thick line, and the conventional increase output method is a change in various quantities indicated by a thin line. In this embodiment, as a result of maintaining the enthalpy of the feed water supplied from the feed water system 14 to the reactor 11 at the same level as that at the rated time, the difference in the enthalpy of the coolant at the inlet and the outlet of the reactor 11 can be secured constant, The rate of increase of the steam flow rate or the feed water flow rate is equivalent to the rate of increase of the reactor heat output. This is shown by the result when ΔH = 0 in the above formula (B) as follows. That is, in the formula (B), if ΔH = 0 is substituted,
[Equation 4]
ΔG = (ΔQ + 0 × G) ÷ (H−0)
ΔG = ΔQ ÷ H (C)
Therefore, the increase amount ΔG of the main steam flow rate or the feed water flow rate is directly proportional to the increase amount ΔQ of the heat output.

  That is, the conventional sharp non-linearity in the characteristics of the increase amount ΔG of the main steam flow rate or the feed water flow rate is eliminated by the application of this embodiment, and the reactor heat output is increased by α% with respect to the Nth cycle. Even so, the increase amount ΔG of the main steam flow rate or the feed water flow rate is suppressed to α% or so on the basis of the proportional relationship with the increase amount ΔQ of the heat output (see formula (C)). The main steam flow rate or feed water flow rate in the heat output operation of 100 + α% in the conventional power increase method corresponds to the main steam flow rate or feed water flow rate of 100 + α + β% (β> 0) under the application of this embodiment.

  Therefore, in this embodiment, since the increase in the main steam flow rate or the feed water flow rate can be suppressed with respect to the same increase plan of the reactor heat output as in the conventional case, the nuclear power plant 10 can be changed without greatly changing the configuration of the nuclear power plant 10. Can be increased. Alternatively, in this embodiment, in the same main steam flow rate or feed water flow rate increase plan as before, it is possible to improve the amount of increase in heat output and increase the rate of increase in reactor heat output.

  This means that the planable range of the increased output that falls within the range of the design margin of the plant equipment including the steam turbines 20 and 21 is substantially expanded, and thus the lifetime power generation amount of the nuclear power plant 10 is improved. From a different perspective, steam turbines, one of the plant equipment that has the greatest potential to interfere with the increase in heat output, can be planned without increasing the main plant equipment. 20, 21 and by extension means the possibility of drafting an increased output plan (ie, reducing the investment amount) that eliminates the need to replace the steam control valve 23 which is dominant with respect to the maximum swallowable steam amount of the steam turbine. .

  Therefore, according to the present embodiment, the following effects (1) to (3) are obtained.

  (1) The enthalpy or temperature at the inlet of the reactor 11 (reactor pressure vessel 19) of the feed water supplied from the feed water system 14 to the reactor 11 is controlled in the (N + 1) th operation cycle in the same manner as the Nth operation cycle. Therefore, the enthalpy difference between the coolant at the inlet and the outlet of the reactor 11 can be ensured to be constant. For this reason, the main steam flow rate at the outlet of the nuclear reactor 11 and the feed water flow rate supplied to the nuclear reactor 11 increase at a rate of increase in the (N + 1) th operation cycle as compared with the Nth operation cycle. N + 1) It can be made equal to the increase rate that increases in the operation cycle compared to the Nth operation cycle. As a result, it is possible to suppress both an increase in the main steam flow rate at the outlet of the reactor 11 and an increase in the feed water flow rate supplied to the reactor 11 with respect to the increase in the reactor thermal output performed to increase the power generation amount. , 21, the steam control valve 23, the feed water pipe 29, the main steam pipe 22, and the load acting on the plant equipment such as the reactor internal structure can be reduced. For this reason, the design margin of these plant equipment can be maintained suitably, and since it is not necessary to change these plant equipment to a thing with a big design margin, the increase output of the nuclear power plant 10 is not carried out, without changing the structure of a plant equipment. Can be realized.

  (2) The ratio of the extraction steam flow extracted from the middle of the high-pressure turbine 20 in the high-pressure steam system 12 or from the outlet through the high-pressure extraction system 15 to the main steam flow at the outlet of the reactor 11 in the (N + 1) th operation cycle. The degree of reduction to be reduced compared to the N operation cycle is the ratio of the extraction steam flow extracted from the low pressure steam system 13 through the low pressure extraction system 16 to the main steam flow at the reactor 11 outlet in the (N + 1) th operation cycle. Then, it is larger than the degree of reduction to be reduced compared to the Nth operation cycle. From this, the amount of steam guided from the high-pressure steam system 12 including the high-pressure turbine 20 to the low-pressure turbine 21 is ensured satisfactorily. For this reason, in the (N + 1) th operation cycle, the flow rate of the extracted steam extracted from the high pressure steam system 12 and the low pressure steam system 13 is simply reduced as compared with the Nth operation cycle, or from the feed water system 14 to the reactor. Compared with the case where the temperature of the feed water supplied to 11 is made lower in the (N + 1) th operation cycle than in the Nth operation cycle, a decrease in thermal efficiency can be suppressed, and a larger electrical output can be obtained.

  (3) Since the temperature of the feed water supplied from the feed water system 14 to the reactor 11 (reactor pressure vessel 19) is kept substantially equal to the Nth operation cycle (rated time) in the (N + 1) th operation cycle, When the output of the nuclear reactor 11 is increased, there is no influence on the coolant density due to the temperature change of the core inlet coolant, and the elements that hinder the increase in the heat output can be eliminated.

  In the first embodiment, the case of a direct cycle nuclear power plant equipped with a boiling water reactor has been described. However, the present invention is applied to another direct cycle type plant, for example, a thermal power plant. It is possible.

[B] Second embodiment (FIGS. 6 and 7)
FIG. 6 is a system configuration diagram of a nuclear power plant showing a configuration at the time of increased output in the second embodiment in which the operation method of the nuclear power plant according to the present invention is applied to a nuclear power plant having a pressurized water reactor. . FIG. 7 is a system configuration diagram showing the configuration of the nuclear power plant of FIG. 6 before the increased output. In the second embodiment, the same parts as those in the first embodiment are denoted by the same reference numerals, and the description will be simplified or omitted.

  The present embodiment differs from the first embodiment in that an indirect cycle type nuclear power plant in which the nuclear power plant 40 of the present embodiment includes a pressurized water reactor 41 as shown in FIGS. 6 and 7. The main steam pipe 22 of the high-pressure steam system 12 and the feed water pipe 29 of the feed water system 14 are connected together to the steam generator 43 instead of the reactor vessel 42 of the reactor 41.

  The steam generator 43 generates steam (water vapor) using the coolant heated in the reactor 41 as a heat source, and constitutes a primary system with the reactor 41. The steam generated in the steam generator 43 is sequentially led to the high-pressure steam system 12 and the low-pressure steam system 13 to work by the high-pressure steam turbine 20 and the low-pressure steam turbine 21, condensed in the condenser 24, and heated by the low-pressure feed water The heater 27 and the high-pressure feed water heater 26 are sequentially heated to return to the steam generator 43, and a secondary system is configured by this path. In the present embodiment, a moisture separator / heater 44 is installed between the high-pressure steam turbine 20 and the low-pressure steam turbine 21. The moisture separator / heater 44 is a moisture separator / reheater. Or a moisture separator.

  The exchange heat quantity in the steam generator 43 is obtained by subtracting the heat leak amount in the primary system from the heat output of the reactor 41. Usually, the heat leak quantity is sufficient as compared with the reactor heat output. Small. For this reason, in this embodiment, it is assumed that the exchange heat quantity in the steam generator 43 and the heat output of the reactor 41 are equal.

  By the way, in the nuclear power plant 40 (FIG. 6, FIG. 7) comprised as mentioned above, when implementing the thermal output increase operation of the nuclear reactor 41, the feed water flow volume supplied to the steam generator 43 from the feed water system 14 is increased. In addition, as shown in FIG. 6, flow rate adjustment valves 17 and 18 as extraction steam adjustment means and flow rate adjustment means are installed in the high pressure extraction pipe 30 and the low pressure extraction pipe 31, respectively. The flow rate adjusting valve 17 adjusts the flow rate of the bleed steam extracted from the high pressure bleed system 15 and introduced into the high pressure feed water heater 26. Further, the flow rate adjustment valve 18 adjusts the flow rate of the extracted steam extracted from the low pressure extraction system 16 and introduced into the low pressure feed water heater 27.

  The extraction steam adjustment means is not limited to the flow adjustment valves 17 and 18, but may be a flow adjustment means such as an office. Further, in each of the high pressure extraction pipe 30 and the low pressure extraction pipe 31, at least in one system. It may be a closing valve or a closing plug (plugging) as closing means for stopping the flow of the extracted steam.

  That is, the period from the start of the reactor 41 to the stop of the operation of the reactor 41 for fuel replacement is defined as one operation period, and the reactor heat output in the (N + 1) th operation cycle, which is the subsequent operation cycle, The reactor heat output is increased in the Nth operation cycle, which is the previous operation cycle before the (N + 1) th operation cycle. At this time, (a) the extraction steam introduced from the high-pressure extraction system 15 to the high-pressure feed water heater 26 is adjusted by the flow rate adjusting valve 17. Alternatively, (b) the flow control valves 17 and 18 adjust the extraction steam introduced from the high pressure extraction system 15 to the high pressure feed water heater 26 and from the low pressure extraction system 16 to the low pressure feed water heater 27, respectively. According to one of (a) and (b), the enthalpy or temperature at the inlet of the steam generator 43 supplied to the steam generator 43 from the feed water system 14 is changed to the Nth operation in the (N + 1) th operation cycle. Control the same as the cycle. Here, it is assumed that the Nth operation cycle is, for example, a rated operation with a reactor heat output of 100%.

  In the case of (a), for example, in the (N + 1) th operation cycle for increasing the reactor heat output, at least one system of the high-pressure extraction pipe 30 in the high-pressure extraction system 15 in which the extraction steam flows at a predetermined flow rate in the N-th operation cycle. Thus, the flow rate of the extracted steam is adjusted to a predetermined flow rate or less by the flow rate adjusting valve 17, or the flow of the extracted steam is stopped by a closing valve or the like. Thus, the ratio of the flow rate of the extracted steam extracted from the high-pressure extraction system 15 and introduced into the high-pressure feed water heater 26 to the main steam flow rate at the outlet of the steam generator 43 is the Nth operating cycle in the (N + 1) th operating cycle. Reduce compared to. Thereby, the temperature rise amount of the feed water in the high-pressure feed water heater 26 is smaller in the (N + 1) th operation cycle than in the Nth operation cycle. As a result, the enthalpy or temperature at the inlet of the steam generator 43 supplied from the feed water system 14 to the steam generator 43 can be controlled in the (N + 1) th operation cycle to be equivalent to the Nth operation cycle. .

  In the case of (b), for example, in the (N + 1) th operation cycle for increasing the reactor heat output, at least the high-pressure extraction pipe 30 in the high-pressure extraction system 15 in which the extraction steam flows at a predetermined flow rate in the Nth operation cycle. In one system, the flow rate of the extracted steam is adjusted to a predetermined flow rate or less by the flow rate adjusting valve 17, or the flow of the extracted steam is stopped by a closing valve or the like. Further, in this (N + 1) th operation cycle, the flow rate of the extraction steam is set to a predetermined flow rate by the flow rate adjusting valve 18 in at least one system of the low pressure extraction pipe 31 in the low pressure extraction system 16 in which the extraction gas flows at a predetermined flow rate in the Nth operation cycle. Adjust the following or stop the flow of extracted steam with a stop valve.

  Accordingly, the steam generator 43 outlets of the extraction steam flow extracted from the high pressure extraction system 15 and led to the high pressure feed water heater 26 and the extraction steam flow extracted from the low pressure extraction system 16 and led to the low pressure feed water heater 27 are extracted. In the (N + 1) th operation cycle, the ratio with respect to the main steam flow rate is reduced together in comparison with the Nth operation cycle. Among these, the degree of decrease in which the ratio of the extraction steam flow rate from the high pressure extraction system 15 to the main steam flow rate at the outlet of the steam generator 43 decreases in the (N + 1) th operation cycle compared to the Nth operation cycle is low pressure. The ratio of the extraction steam flow rate from the extraction system 16 to the main steam flow rate at the outlet of the steam generator 43 is set to be greater than the degree of decrease in the (N + 1) th operation cycle compared to the Nth operation cycle.

  Thereby, the total temperature increase amount of the feed water by the high-pressure feed water heater 26 and the low-pressure feed water heater 27 is smaller than the N-th operation cycle in the (N + 1) th operation cycle, and the high-pressure feed water heater for the total temperature rise amount The ratio of the temperature rise amount of the water supply by 26 is smaller in the (N + 1) th operation cycle than in the Nth operation cycle. As a result, the enthalpy or temperature at the inlet of the steam generator 43 introduced into the steam generator 43 from the feed water system 14 can be controlled in the (N + 1) th operation cycle in the same manner as the Nth operation cycle. .

  Here, controlling the feed water temperature at the inlet of the steam generator 43 supplied to the steam generator 43 from the feed water system 14 in the (N + 1) th operation cycle is equivalent to the Nth operation cycle. Controlling the feed water temperature to the steam generator 43 in the operation cycle to a temperature range of ± 1 ° C. with respect to the feed water temperature to the steam generator 43 in the Nth operation cycle.

  FIG. 3 shows the relationship between the reactor heat output, the main steam flow rate (main steam flow rate flowing from the steam generator 43 to the main steam pipe 22), the extraction steam flow rate, and the operation cycle. It is a graph shown about the case at the time of a form and the conventional increase output.

  In the operation cycle shown in FIG. 3, the Nth operation cycle and the (N−1) th operation cycle are operation cycles before the power increase method of the present embodiment is applied. At this time, the reactor heat output is Q = 100% (rated). The system configuration of the nuclear power plant 40 before the increased output is shown in FIG. In the (N + 1) th operation cycle, the reactor heat output is increased by α% to Q = 100 + α%, and the system configuration of the nuclear power plant 40 at this increased output is shown in FIG. Each operation cycle shows a normal operation state, and excluding the operation state at the time of starting, stopping, transient state, and accident.

  The means for increasing the reactor heat output can be realized by a method in which the amount of control rods to be extracted in the (N + 1) th operation cycle is larger than that in the Nth cycle or the type of fuel assembly is changed. is there.

  Depending on the nuclear power plant 40, the extraction steam flow rate and the main steam flow rate during one operation cycle may be changed as shown in FIG. Such an operation cycle reduces the temperature of the feed water supplied from the feed water system 14 to the steam generator 43 by reducing the flow rate of the extracted steam in order to keep the reactor heat output constant even at the end of combustion of the fuel. It is something to be made. Also in the case of the nuclear power plant 40 having an operation cycle as shown in FIG. 4, the heat balance, the extraction steam flow rate, the main steam flow rate, the feed water heating amount, etc. are the operation cycle excluding start / stop, occurrence of an accident / transient event, and trial operation. Among them, the operation with the maximum main steam flow shall be compared.

  By the way, in general, when the reactor heat output is increased, in order to acquire the increased amount of heat, the flow rate of the feed water supplied from the feed water system 14 to the steam generator 43 is increased, or steam is generated. It is necessary to enlarge the enthalpy difference between the inlet and outlet of the vessel 43. In the latter case, in the conventional power increase method, the temperature and enthalpy of water supplied from the feed water system 14 to the steam generator 43 at the time of increased power increase, so the main steam enthalpy at the outlet of the steam generator 43 is increased. The enthalpy difference will be expanded. However, since the rise of the main steam enthalpy at the outlet of the steam generator 43 is difficult from the technical and safety aspects such as high pressure and high temperature in the primary reactor vessel 42, the water supply to the steam generator 43 is difficult. As the enthalpy rises, the enthalpy difference between the inlet and the outlet in the steam generator 43 is rather reduced, and from the above formula (A), the main steam flow rate or the feed water flow rate in the steam generator 43 must be increased. I don't get it. Therefore, the former method is exclusively used to increase the feed water flow rate to the steam generator 43.

  Here, the reactor heat output assumes that the main steam flow rate and the feed water flow rate in the steam generator 43 are the same, the enthalpy difference between the inlet and the outlet of the steam generator 43, and the steam generator 43 Since it is defined as the product of the feed water flow rate or the main steam flow rate in FIG.

  In general, in the nuclear power plant 40 having the steam generator 43, the main steam pressure and the enthalpy in the steam generator 43 tend to decrease as the reactor heat output increases. Compared to the rate of change in quantities related to the heat balance around the steam generator 43, including the feed water temperature and enthalpy to the generator 43. Therefore, in the present embodiment, the main steam enthalpy in the steam generator 43 is assumed to be equal before and after the increased output of the reactor 41.

  In the conventional power increase method, the characteristic of the increase ΔG in the feed water flow rate or the main steam flow rate in the steam generator 43 before and after the power increase is the same as in the case of the boiling water reactor. It can be seen that the non-linearity is not directly proportional to the increase amount ΔQ of the heat output but depends on the plan / design of the increase output. Further, as shown in the equation (B) and the thin line in FIG. 5, this non-linearity becomes more prominent as the amount of increase ΔH of the water supply enthalpy to the steam generator 43 increases, and from H> ΔH> 0, The rate of increase in the main steam flow rate or the feed water flow rate in the steam generator 43 has a characteristic of increasing more sharply with respect to the rate of increase in heat output.

  On the other hand, in the present embodiment, as shown in FIG. 6, the flow rate of the feed water supplied from the feed water system 14 to the steam generator 43 is increased and the steam generator 43 The enthalpy difference between the inlet and the outlet of the steam generator 43 is kept constant by intentionally planning and designing the water supply enthalpy at the inlet in the same manner as at the time of rating.

  In order to maintain the same water supply enthalpy at the inlet of the steam generator 43 before and after the increased output, when the increased output is planned, the high pressure steam system 12 and the low pressure steam system 13 are extracted and led to the feed water heaters 26 and 27. What is necessary is just to reduce the extraction steam flow rate. However, if the amount of extracted steam is simply reduced as a whole, the amount of steam introduced to the low-pressure turbine 21 is insufficient, the thermal efficiency is greatly reduced, and the amount of power generation cannot be increased. Therefore, the flow rate of the extracted steam from the middle of the high-pressure turbine 20 or from the outlet through the high-pressure extraction system 15 is selectively reduced with respect to the low-pressure turbine 21 side. As a result, the amount of steam flowing through the low-pressure turbine 21 can be increased to increase the amount of power generation.

  Most of the extracted steam from the middle or outlet of the high-pressure turbine 20 is used in the high-pressure feed water heater 26 installed on the downstream side of the main feed water pump 28. Therefore, the power increase method according to this embodiment changes the way of viewing. And a method of controlling the amount of heating of the feed water downstream of the main feed water pump 28.

  In the case of a nuclear power plant where the flow rate of the extracted steam extracted from the middle or the outlet of the high-pressure turbine 20 is originally small, the low-pressure turbine 21 is used in order to make the enthalpy of water supply to the steam generator 43 equal to the rated value. It is necessary to reduce the flow rate of the extracted steam extracted through the low-pressure extraction system 16 and sufficiently lower the temperature of the feed water to the steam generator 43. Even when this embodiment is applied to such a nuclear power plant, the flow rate of the bleed steam from the middle or outlet of the high pressure turbine 20 is reduced more than the flow rate of the bleed steam from the low pressure turbine 21 side. Some effect can be obtained.

  The transition of various quantities related to the heat balance around the reactor 41 and the change in characteristics due to the increased output of this embodiment are the same as those in the boiling water reactor as in FIG. The result of application of the present embodiment is a change in various quantities indicated by a thick line, and the conventional increase output method is a change in various quantities indicated by a thin line. In this embodiment, as a result of maintaining the enthalpy of the feed water supplied from the feed water system 14 to the steam generator 43 at the same level as the rated value, a constant enthalpy difference between the inlet and the outlet of the steam generator 43 can be secured, and steam generation The increase rate of the main steam flow rate or the feed water flow rate in the reactor 43 is equivalent to the increase rate of the reactor heat output from the above formula (C) and FIG.

  That is, the conventional sharp non-linearity in the characteristic of the increase amount ΔG of the main steam flow rate or the feed water flow rate in the steam generator 43 disappears by the application of the present embodiment, and the reactor heat output is reduced by α to the Nth cycle. Even if the plan is to increase by%, the increase amount ΔG of the main steam flow rate or the feed water flow rate is suppressed to be equivalent to α% based on the proportional relationship with the increase amount ΔQ of the heat output (see formula (C)). As shown in FIG. 5, the main steam flow rate or feed water flow rate in the heat output operation of 100 + α% in the conventional power increase method is 100 + α + β% (β> 0) main steam flow rate or feed water as shown in FIG. Corresponds to the flow rate.

  Therefore, in the present embodiment, the increase in the main steam flow rate or the feed water flow rate in the steam generator 43 can be suppressed with respect to the same increase plan of the reactor heat output as in the past, so that the configuration of the nuclear power plant 40 can be significantly changed. Therefore, the increased output of the nuclear power plant 40 can be realized. Alternatively, in the present embodiment, in the plan for increasing the main steam flow rate or the feed water flow rate in the same steam generator 43 as in the prior art, it is possible to improve the amount of increase in heat output and increase the rate of increase in reactor heat output. It becomes possible.

  This means that the planable range of the increased output that falls within the design margin of the plant equipment including the steam turbines 20 and 21 is substantially expanded, and thus the lifetime power generation of the nuclear power plant 40 is improved. From a different perspective, steam turbines, one of the plant equipment that has the greatest potential to interfere with the increase in heat output, can be planned without increasing the main plant equipment. 20, 21 and by extension means the possibility of drafting an increased output plan (ie, reducing the investment amount) that eliminates the need to replace the steam control valve 23 which is dominant with respect to the maximum swallowable steam amount of the steam turbine. .

  According to the present embodiment, the following effects (4) and (5) are obtained.

  (4) Since the enthalpy or temperature at the inlet of the steam generator 43 of the feed water supplied from the feed water system 14 to the steam generator 43 is controlled in the (N + 1) th operation cycle in the same way as the Nth operation cycle, A constant enthalpy difference between the inlet and the outlet of the generator 43 can be secured. For this reason, the ratio of the main steam flow rate at the outlet of the steam generator 43 and the feed water flow rate supplied to the steam generator 43 increases in the (N + 1) th operation cycle as compared with the Nth operation cycle, In the (N + 1) operation cycle, the increase rate can be made equal to that of the Nth operation cycle. As a result, it is possible to suppress both an increase in the main steam flow rate at the outlet of the steam generator 43 and an increase in the feed water flow rate supplied to the steam generator 43 with respect to the increase in the reactor heat output performed to increase the power generation amount. Loads acting on plant equipment such as the turbines 20 and 21, the steam control valve 23, the water supply pipe 29, and the main steam pipe 22 can be reduced. For this reason, the design margin of these plant equipment can be suitably maintained, and since it is not necessary to change these plant equipment to a thing with a large design margin, it is possible to increase the output of the nuclear power plant 40 without changing the configuration of the plant equipment. Can be realized.

  (5) The ratio of the extraction steam flow extracted from the middle of the high-pressure turbine 20 in the high-pressure steam system 12 through the high-pressure extraction system 15 to the main steam flow at the outlet of the steam generator 43 in the (N + 1) th operation cycle. The degree of decrease compared to the Nth operation cycle is the ratio of the extraction steam flow rate extracted from the low pressure steam system 13 through the low pressure extraction system 16 to the main steam flow rate at the outlet of the steam generator 43 (N + 1). In the operation cycle, the degree of reduction is smaller than that in the Nth operation cycle. From this, the amount of steam guided from the high-pressure steam system 12 including the high-pressure turbine 20 to the low-pressure turbine 21 is ensured satisfactorily. Therefore, when the flow rate of the extracted steam extracted from the high-pressure steam system 12 and the low-pressure steam system 13 in the (N + 1) -th operation cycle is simply reduced as compared with the N-th operation cycle, steam is generated from the water supply system 14. Compared with the case where the temperature of the feed water supplied to the vessel 43 is lower in the (N + 1) th operation cycle than in the Nth operation cycle, it is possible to suppress a decrease in thermal efficiency and obtain a larger electrical output.

  In the second embodiment, the case of an indirect cycle type nuclear power plant equipped with a pressurized water reactor has been described. However, the present invention is applied to other plants of the indirect cycle type, such as a geothermal power plant. May be.

The nuclear power plant system block diagram which shows the structure at the time of increase output in 1st Embodiment which applied the operating method of the nuclear power plant which concerns on this invention to the nuclear power plant which comprises a boiling water reactor. The line | wire system block diagram which shows the structure before the increase output of the nuclear power plant of FIG. The graph which shows the relationship between an operation cycle, a nuclear reactor heat output, a main steam flow rate, and an extraction steam flow rate. The graph which shows the other relationship of an operation cycle, nuclear reactor heat output, main steam flow, and extraction steam flow. FIG. 7 is a graph showing a comparison of changes in heat balance-related quantities around a nuclear reactor after increasing power in the nuclear power plant of FIGS. 1 and 6 and a conventional nuclear power plant. The nuclear power plant system block diagram which shows the structure at the time of increased output in 2nd Embodiment which applied the operating method of the nuclear power plant which concerns on this invention to the nuclear power plant which comprises a pressurized water reactor. The system | strain block diagram which shows the structure before the increase output of the nuclear power plant of FIG.

Explanation of symbols

DESCRIPTION OF SYMBOLS 10 Nuclear power plant 11 Reactor 12 High-pressure steam system 13 Low-pressure steam system 14 Water supply system 15 High-pressure extraction system 16 Low-pressure extraction system 17, 18 Flow control valve 20 High-pressure turbine 21 Low-pressure turbine 25 Condenser 26 High-pressure feed water heater 27 Low-pressure feed water heating 28 Main feed pump 30 High pressure extraction pipe 40 Nuclear plant 41 Reactor 43 Steam generator

Claims (12)

A nuclear reactor,
A high-pressure steam system from the reactor outlet to the low-pressure turbine inlet to which steam generated in the reactor is supplied;
A low pressure steam system from the low pressure turbine inlet to a condenser inlet for condensing steam discharged from the low pressure turbine;
In a method for operating a nuclear power plant, comprising the condenser and a feed water heater that heats feed water supplied from the condenser, and a feed water system that supplies feed water from the feed water heater to the nuclear reactor. ,
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
The ratio of the mass flow rate of the extracted steam extracted from the high-pressure steam system and led to the feed water heater to the mass flow rate of the main steam at the reactor outlet is reduced in the subsequent operation cycle compared to the previous operation cycle. so,
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a reactor inlet of feed water supplied from the feed water system to the reactor is controlled in a later operation cycle in the same manner as the previous operation cycle. A nuclear reactor,
A high-pressure steam system from the reactor outlet to the low-pressure turbine inlet to which steam generated in the reactor is supplied;
A low pressure steam system from the low pressure turbine inlet to a condenser inlet for condensing steam discharged from the low pressure turbine;
In a method for operating a nuclear power plant, comprising the condenser and a feed water heater that heats feed water supplied from the condenser, and a feed water system that supplies feed water from the feed water heater to the nuclear reactor. ,
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
The ratio of the mass flow rate of the bleed steam extracted from the high pressure steam system and the low pressure steam system to the feed water heater, with respect to the mass flow rate of the main steam at the reactor outlet, Both compared to the cycle,
Of these, the ratio of the mass flow rate of the extracted steam from the high-pressure steam system to the mass flow rate of the main steam at the reactor outlet decreases in the later operation cycle as compared with the previous operation cycle, By making the ratio of the mass flow rate of the bleed steam from the steam system to the mass flow rate of the main steam at the reactor outlet larger than the degree of decrease compared to the previous operation cycle in the later operation cycle,
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a reactor inlet of feed water supplied from the feed water system to the reactor is controlled in a later operation cycle in the same manner as the previous operation cycle. A nuclear reactor,
A steam system including a high-pressure turbine and a low-pressure turbine to which steam generated in the nuclear reactor is supplied;
A condenser for condensing the steam discharged from the low-pressure turbine, and a low-pressure feed water heater installed on the downstream side of the condenser and the upstream side of the main feed pump for heating the feed water supplied from the condenser. And a high-pressure feed water heater installed downstream from the main feed pump and upstream from the reactor, and a feed water system for supplying feed water from the high-pressure feed water heater to the reactor In the operation method of a nuclear power plant,
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
By reducing the amount of temperature increase in the high-pressure feed water heater in the subsequent operation cycle compared to the previous operation cycle,
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a reactor inlet of feed water supplied from the feed water system to the reactor is controlled in a later operation cycle in the same manner as the previous operation cycle. A nuclear reactor,
A steam system including a high-pressure turbine and a low-pressure turbine to which steam generated in the nuclear reactor is supplied;
A condenser for condensing the steam discharged from the low-pressure turbine, and a low-pressure feed water heater installed on the downstream side of the condenser and the upstream side of the main feed pump for heating the feed water supplied from the condenser. And a high-pressure feed water heater installed downstream from the main feed pump and upstream from the reactor, and a feed water system for supplying feed water from the high-pressure feed water heater to the reactor In the operation method of a nuclear power plant,
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
By making the ratio of the temperature increase amount in the high pressure feed water heater to the total temperature increase amount in the low pressure feed water heater and the high pressure feed water heater smaller in the later operation cycle than in the previous operation cycle,
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a reactor inlet of feed water supplied from the feed water system to the reactor is controlled in a later operation cycle in the same manner as the previous operation cycle. A nuclear reactor,
A steam system including a high-pressure turbine and a low-pressure turbine to which steam generated in the nuclear reactor is supplied;
A water supply system comprising a condenser for condensing steam discharged from the low-pressure turbine, and a feed water heater for heating feed water supplied from the condenser, and supplying feed water from the feed water heater to the reactor When,
In a method for operating a nuclear power plant, comprising: an extraction steam system including at least one high-pressure extraction pipe that extracts steam from the high-pressure turbine side and guides the steam to the feed water heater;
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
In the subsequent operation cycle, the extraction steam is stopped or the mass flow rate of the extraction steam is adjusted to be equal to or less than the predetermined flow rate in at least one of the high-pressure extraction pipes in which the extraction steam was flowing at the predetermined flow rate in the previous operation cycle. so,
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a reactor inlet of feed water supplied from the feed water system to the reactor is controlled in a later operation cycle in the same manner as the previous operation cycle. A nuclear reactor,
A steam generator that generates steam using a coolant heated in the reactor as a heat source;
A high-pressure steam system from the steam generator outlet to the low-pressure turbine inlet to which steam generated by the steam generator is supplied;
A low pressure steam system from the low pressure turbine inlet to a condenser inlet for condensing steam discharged from the low pressure turbine;
An operation method of a nuclear power plant comprising: the condenser, and a feed water heater that heats the feed water supplied from the condenser, and a feed water system that feeds the feed water from the feed water heater to the steam generator In
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
The ratio of the mass flow rate of the extracted steam that is extracted from the high-pressure steam system and led to the feed water heater to the mass flow rate of the main steam at the outlet of the steam generator is reduced in the subsequent operation cycle compared to the previous operation cycle. With that
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a steam generator inlet of feed water supplied from the feed water system to the steam generator is controlled in a later operation cycle to be equal to a previous operation cycle. A nuclear reactor,
A steam generator that generates steam using a coolant heated in the reactor as a heat source;
A high-pressure steam system from the steam generator outlet to the low-pressure turbine inlet to which steam generated by the steam generator is supplied;
A low pressure steam system from the low pressure turbine inlet to a condenser inlet for condensing steam discharged from the low pressure turbine;
An operation method of a nuclear power plant comprising: the condenser, and a feed water heater that heats the feed water supplied from the condenser, and a feed water system that feeds the feed water from the feed water heater to the steam generator In
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
The ratio of the mass flow rate of the bleed steam extracted from the high pressure steam system and the low pressure steam system to the feed water heater with respect to the mass flow rate of the main steam at the outlet of the steam generator, Reduce both compared to the driving cycle,
Of these, the degree of reduction in which the ratio of the mass flow rate of the extracted steam from the high-pressure steam system to the mass flow rate of the main steam at the outlet of the steam generator decreases in the later operation cycle compared to the previous operation cycle, By making the ratio of the mass flow rate of the extracted steam from the low-pressure steam system to the mass flow rate of the main steam at the outlet of the steam generator larger than the degree of decrease that is reduced in the subsequent operation cycle compared to the previous operation cycle,
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a steam generator inlet of feed water supplied from the feed water system to the steam generator is controlled in a later operation cycle to be equal to a previous operation cycle. A nuclear reactor,
A steam generator that generates steam using a coolant heated in the reactor as a heat source;
A steam system including a high-pressure turbine and a low-pressure turbine to which steam generated by the steam generator is supplied;
A condenser for condensing the steam discharged from the low-pressure turbine, and a low-pressure feed water heater installed on the downstream side of the condenser and the upstream side of the main feed pump for heating the feed water supplied from the condenser. And a high-pressure feed water heater installed downstream from the main feed pump and upstream from the steam generator, and a feed water system for supplying feed water from the high-pressure feed water heater to the steam generator; In a method of operating a nuclear power plant having
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
By reducing the amount of temperature increase in the high-pressure feed water heater in the subsequent operation cycle compared to the previous operation cycle,
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a steam generator inlet of feed water supplied from the feed water system to the steam generator is controlled in a later operation cycle to be equal to a previous operation cycle. A nuclear reactor,
A steam generator that generates steam using a coolant heated in the reactor as a heat source;
A steam system including a high-pressure turbine and a low-pressure turbine to which steam generated by the steam generator is supplied;
A condenser for condensing the steam discharged from the low-pressure turbine, and a low-pressure feed water heater installed on the downstream side of the condenser and the upstream side of the main feed pump for heating the feed water supplied from the condenser. And a high-pressure feed water heater installed downstream from the main feed pump and upstream from the steam generator, and a feed water system for supplying feed water from the high-pressure feed water heater to the steam generator; In a method of operating a nuclear power plant having
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
By making the ratio of the temperature increase amount in the high pressure feed water heater to the total temperature increase amount in the low pressure feed water heater and the high pressure feed water heater smaller in the later operation cycle than in the previous operation cycle,
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a steam generator inlet of feed water supplied from the feed water system to the steam generator is controlled in a later operation cycle to be equal to a previous operation cycle. A nuclear reactor,
A steam generator that generates steam using a coolant heated in the reactor as a heat source;
A steam system including a high-pressure turbine and a low-pressure turbine to which steam generated by the steam generator is supplied;
A water supply comprising a condenser for condensing the steam discharged from the low-pressure turbine, and a feed water heater for heating the feed water supplied from the condenser, and supplying the feed water from the feed water heater to the steam generator The system,
In a method for operating a nuclear power plant, comprising: an extraction steam system including at least one high-pressure extraction pipe that extracts steam from the high-pressure turbine side and guides the steam to the feed water heater;
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. Increased from the reactor thermal power in the operating cycle of
In the subsequent operation cycle, the extraction steam is stopped or the mass flow rate of the extraction steam is adjusted to be equal to or less than the predetermined flow rate in at least one of the high-pressure extraction pipes in which the extraction steam was flowing at the predetermined flow rate in the previous operation cycle. so,
An operation method of a nuclear power plant, wherein an enthalpy or temperature at a steam generator inlet of feed water supplied from the feed water system to the steam generator is controlled in a later operation cycle to be equal to a previous operation cycle. A nuclear reactor,
A steam system including a high-pressure turbine and a low-pressure turbine to which steam generated in the nuclear reactor is supplied;
A water supply system comprising a condenser for condensing steam discharged from the low-pressure turbine, and a feed water heater for heating feed water supplied from the condenser, and supplying feed water from the feed water heater to the reactor When,
An extraction steam system comprising at least one high-pressure extraction pipe for extracting steam from the high-pressure turbine side and guiding it to the feed water heater;
An extraction steam adjusting means provided in the extraction steam system for adjusting the extraction steam;
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. When increased from the reactor thermal power in the operating cycle of
The extraction steam is adjusted by the extraction steam adjusting means, and the enthalpy or temperature at the reactor inlet of the feed water supplied from the feed water system to the reactor is controlled in the later operation cycle in the same manner as the previous operation cycle. A nuclear power plant configured as described above. A nuclear reactor,
A steam generator that generates steam using a coolant heated in the reactor as a heat source;
A steam system including a high-pressure turbine and a low-pressure turbine to which steam generated by the steam generator is supplied;
A water supply comprising a condenser for condensing the steam discharged from the low-pressure turbine, and a feed water heater for heating the feed water supplied from the condenser, and supplying the feed water from the feed water heater to the steam generator The system,
An extraction steam system comprising at least one high-pressure extraction pipe for extracting steam from the high-pressure turbine side and guiding it to the feed water heater;
An extraction steam adjusting means provided in the extraction steam system for adjusting the extraction steam;
When the period from the start-up of the reactor to the stop of the operation of the reactor for fuel replacement is defined as one operation cycle, the reactor heat output in the subsequent operation cycle is determined prior to the subsequent operation cycle. When increased from the reactor thermal power in the operating cycle of
The extraction steam is adjusted by the extraction steam adjusting means, and the enthalpy or temperature at the steam generator inlet of the feed water supplied from the feed water system to the steam generator is made equal to the previous operation cycle in the subsequent operation cycle. A nuclear power plant configured to control.

Images