JP4516438B2 - Operation method of nuclear power plant - Google Patents

Description

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

  In a conventional new nuclear power plant, for example, the fuel configuration or the shape configuration of the fuel assembly is improved to increase the electrical output, and the main steam flow rate at the core outlet is increased to increase the electrical output. It was.

  Such a conventional technique is disclosed in Japanese Patent Laid-Open No. 9-264983.

JP-A-9-264983

  When the above-mentioned conventional technology is applied to an existing nuclear power plant, the core flow rate flowing through the reactor core is almost the same as before the increase in electrical output, and the thermal output of the core increases. The void ratio (ratio of steam in the channel volume) increases. This increases the flow rate of the coolant and increases the core pressure loss. Further, since the amount of steam generated in the core increases, the pressure loss in the two-phase flow portion of water-steam increases and the margin of stability of the core tends to decrease. Furthermore, when the core average void ratio increases, the amount of steam that condenses during so-called pressure transients, which increases the pressure when the load of the generator is interrupted, increases, and the reduction range of the core average void ratio increases. In general, a boiling water light water reactor has a negative void feedback coefficient so that the reactor power decreases as the void ratio increases. However, during such pressure transients, the reactor core output increases because the average void fraction of the core decreases. According to the conventional technique, after the electrical output is increased, the decrease in the core average void ratio during the pressure transient increases as described above, and the design margin for the pressure transient may also decrease.

  Apart from that, the main steam flow increases almost in proportion to the increase in electrical output. For this reason, due to the increase in main steam flow, water supply system equipment such as feed water piping, feed water heaters, feed water pumps, in-furnace structures such as steam dryers, main steam pipes, high pressure turbines, low pressure turbines, condensers, etc. The design margin of all equipment is reduced. In a nuclear power plant using a normal boiling water light water reactor, one of the devices that may lose its design margin due to an increase in the main steam flow rate is a high-pressure turbine. In nuclear power generation systems other than boiling water reactors, there is a similar problem for plants with relatively small design margins for high-pressure turbines. When applying conventional technology to existing nuclear power plants, large-scale plant equipment is required. Improvement and exchange were necessary. In order to suppress the increase in the main steam flow rate, it is only necessary to lower the feed water temperature. However, if the extraction for heating the feed water is simply reduced as a whole, the thermal efficiency is greatly deteriorated and the electrical output hardly increases. Not right.

  The purpose of the present invention is to increase the power output of the plant without making a major change in the configuration of the plant equipment, while making the pressure loss characteristics, stability margin, and design margin during transients almost the same as before the power increase. It is to provide a method for operating a nuclear power plant.

  In order to achieve the above object, according to the present invention, when the period from the start of the nuclear power plant to the stop of the operation of the nuclear power plant for fuel replacement is defined as one operation cycle, the second reactor heat output in the second operation cycle is obtained. , More than the first reactor heat output in the first operation cycle of at least one operation cycle before the second operation cycle, extracted from the steam system and led to the feed water heater, especially in the middle of the high-pressure turbine and The ratio of the extracted steam from the high pressure turbine outlet (actually from the high pressure turbine outlet to the inlet of either the moisture separator or moisture separator superheater or moisture separator reheater) to the main steam flow rate. The second operation cycle is smaller than the first operation cycle, and the temperature of the water discharged from the feed water heater is decreased by 1 ° C. or more and 40 ° C. or less.

  In order to achieve the above object, the present invention provides the second reactor heat output in the second operation cycle of the nuclear reactor with the first reactor heat in the first operation cycle at least one operation cycle before the second operation cycle. The mass flow rate of the extracted steam, which is increased from the output, extracted from the steam system and led to the feed water heater, particularly during the high-pressure turbine and from the outlet of the high-pressure turbine, is set to the second operating cycle with respect to the first operating cycle. Then, reduce the temperature of the water discharged from the feed water heater by 1 ° C. or more and 40 ° C. or less.

In order to achieve the above object, the present invention provides the second reactor heat output in the second operation cycle of the nuclear reactor with the first reactor heat in the first operation cycle at least one operation cycle before the second operation cycle. The mass flow rate of the extracted steam that is extracted from the steam system and led to the feed water heater is reduced in the second operation cycle with respect to the first operation cycle. It is to reduce the temperature rise in the high-pressure feed water heater installed downstream from the feed water pump and to reduce the temperature of the water discharged from the feed water heater by 1 ° C. or more and 40 ° C. or less.

  In order to achieve the above object, the present invention provides the second reactor heat output in the second operation cycle of the nuclear reactor with the first reactor heat in the first operation cycle at least one operation cycle before the second operation cycle. Discharge from the feed water heater by stopping at least one of the extracted steam pipes from the steam system and extracting from the steam system to the feed water heater, especially during the high pressure turbine and from the high pressure turbine outlet. The temperature of the water to be reduced is 1 ° C. or more and 40 ° C. or less.

In order to achieve the above object, according to the present invention, the second reactor heat output (Q2) in the second operation cycle of the reactor is changed to the first in the first operation cycle at least one operation cycle before the second operation cycle. Increase the reactor thermal output (Q1) by A%, extract steam from the steam system and lead it to the feed water heater, especially during the high-pressure turbine and at the high-pressure turbine outlet (actually moisture separation from the high-pressure turbine outlet) In the second operating cycle, the ratio of the extracted steam to the main steam flow rate from the steamer or the moisture separation superheater or the moisture separation reheater to the inlet of any one of the second operation cycle is reduced. The temperature of the first operation cycle of the feed water discharged from the feed water heater is T1 (° C.), the temperature in the second operation cycle is T 2 (° C.), and the reactor in the second operation cycle Flow into When the heart rate and W (kg / s),
0 <A ≦ 5,
T2 ≦ T1-7.7 × (Q2-Q1) / (4.5 × W),
It is to do.

In order to achieve the above object, according to the present invention, the second reactor heat output (Q2) in the second operation cycle of the reactor is changed to the first in the first operation cycle at least one operation cycle before the second operation cycle. Increase the reactor thermal output (Q1) by A%, extract steam from the steam system and lead it to the feed water heater, especially during the high-pressure turbine and at the high-pressure turbine outlet (actually moisture separation from the high-pressure turbine outlet) In the second operating cycle, the ratio of the extracted steam to the main steam flow rate from the steamer or the moisture separation superheater or the moisture separation reheater to the inlet of any one of the second operation cycle is reduced. The temperature of the first operation cycle of the feed water discharged from the feed water heater is T1 (° C.), the temperature in the second operation cycle is T 2 (° C.), and the reactor in the second operation cycle Flow into When the heart rate and W (kg / s),
5 <A ≦ 10,
T1-40 ≦ T2 ≦ T1-7.7 × (Q2 × (A + 95) / 100-Q1) / (4.5 × W)
It is to do.

In order to achieve the above object, according to the present invention, the second reactor heat output (Q2) in the second operation cycle of the reactor is changed to the first in the first operation cycle at least one operation cycle before the second operation cycle. Increase the reactor thermal output (Q1) by A%, extract steam from the steam system and lead it to the feed water heater, especially during the high-pressure turbine and at the high-pressure turbine outlet (actually moisture separation from the high-pressure turbine outlet) In the second operating cycle, the ratio of the extracted steam to the main steam flow rate from the steamer or the moisture separation superheater or the moisture separation reheater to the inlet of any one of the second operation cycle is reduced. The temperature of the first operation cycle of the feed water discharged from the feed water heater is T1 (° C.), the temperature in the second operation cycle is T 2 (° C.), and the reactor in the second operation cycle Flow into When the heart rate and W (kg / s),
10 <A <30,
T2 ≦ T1-7.7 × (Q2 × (A + 90) / 100-Q1) / (4.5 × W)
It is to do.

  According to the present invention, regarding the increased output of the existing nuclear power plant, the core pressure loss characteristics, stability margin, thermal margin, and design margin during transients are substantially the same as before the increased output, while greatly increasing the configuration of the nuclear plant. It will be possible to increase the output of the nuclear power plant without changing to.

  An embodiment when the present invention is applied to a boiling water light water reactor system which is one of direct cycle nuclear power plants will be described.

  FIG. 1 shows the overall configuration of a boiling water light water reactor system according to the present embodiment together with an example of a heat balance at the time of increased output of the operation method according to the present embodiment, and FIG. 2 shows the same boiling water light water reactor before the increased output. An example of the heat balance of the system is shown. FIG. 3 shows an example of the heat balance of a conventional boiling water light water reactor system at the time of increased output. 4 and 5 are conceptual diagrams of operation cycles according to the operation method of the present embodiment. 1, 2, and 3, the reactor thermal output is represented by Q, the mass flow rate of water and steam is represented by G, the enthalpy of water and steam is represented by H, and the thermal output Q and mass flow rate are represented by G represents a ratio (%) to the reactor thermal power before the increase in power output shown in FIG. 2 and the steam flow rate at the reactor pressure vessel outlet, and enthalpy is represented by a numerical value in units of (kJ / kg). In addition, each embodiment of the present invention shows a normal operation state, and excludes an operation state at the time of starting and stopping, a transient state, an operation state in which the core heat output is changed by the control rod, and an operation state at the time of an accident.

  First, the whole structure of the boiling water type light water reactor system concerning a present Example is demonstrated using FIG.

  In FIG. 1, reference numeral 1 denotes a reactor pressure vessel, which includes a recirculation pump and a jet pump in the reactor pressure vessel 1, and adjusts the core flow rate flowing through the core. The reactor pressure vessel 1 constitutes a reactor 21 together with its internal structure, and steam generated in the reactor 21 is supplied to the steam system 22. The steam system 22 includes a main steam pipe 2, a high-pressure turbine 3 and a low-pressure turbine 5 connected to the main steam pipe 2, and a moisture separator 4 positioned between the high-pressure turbine 3 and the low-pressure turbine 5. The portion from the reactor outlet including the high pressure turbine 3 to the inlet of the low pressure turbine 5 constitutes a high pressure steam system 22A, and the portion from the inlet of the low pressure turbine 5 to the inlet of the condenser 6 constitutes a low pressure steam system 22B. Yes. The condenser 6 condenses the steam discharged from the low pressure turbine 5. The condensate condensed in the condenser 6 is supplied to the feed water system 23 as feed water, and the feed water system 23 heats the feed water and returns it to the reactor 21. The water supply system 23 is installed on the downstream side of the main feed pump 8 and on the downstream side of the condenser 6 and on the upstream side of the main feed water pump 8, and the low pressure feed water heater 7 for heating the feed water supplied from the condenser 6; It includes a high-pressure feed water heater 9 installed downstream from the main feed pump 8 and upstream from the reactor 21, and the water discharged from the high-pressure feed water heater 9 is directed toward the reactor 21 via the feed water pipe 24. Led.

  Between the steam system 22 and the high pressure feed water heater 9 and the low pressure feed water heater 7, extraction pipes 25, 26, 27 are extracted from the steam system 22 and led to the high pressure feed water heater 9 and the low pressure feed water heater 7. , 28 are provided. The extraction pipe 25 extracts steam from the middle of the high-pressure turbine 3 and guides it to the high-pressure feed water heater 9. The extraction pipe 26 is connected to the outlet of the high-pressure turbine 3 (actually downstream from the outlet of the high-pressure turbine 3 and moisture separation). Steam is extracted from the upstream side of the inlet of the vessel 4 and led to the high-pressure feed water heater 9. The extraction pipe 27 extracts steam from the middle of the moisture separator 4 and guides it to the low-pressure feed water heater 7, and the extraction pipe 28 extracts steam from the middle of the low-pressure turbine 5 and guides it to the low-pressure feed water heater 7. The extraction pipe 25 is provided with an extraction pipe flow rate adjustment valve 10 which is an extraction steam amount adjusting means for adjusting the amount of extraction steam introduced from the middle of the high pressure turbine 3 to the high pressure feed water heater 9.

  Next, the operation method of the nuclear power plant according to this embodiment will be described.

  In the heat balance shown in FIGS. 1 to 3, as described above, the thermal output Q and the mass flow rate G correspond to the reactor thermal output before the increase in output shown in FIG. 2 and the steam flow rate at the reactor pressure vessel outlet. The ratio (%) is shown. As can be seen from the heat balance shown in FIGS. 1 to 3, in the operation method according to the present embodiment, the second reactor heat output in the second operation cycle (FIG. 1) of the reactor 21 is set before the second operation cycle. In the first operation cycle (FIG. 2), the heat output from the first reactor (FIG. 2) is increased (Q = 100 → 105), and the feed water heater 9 is extracted from the high-pressure steam system 22A in the first operation cycle. In contrast to the ratio (19/100) of the mass flow rate (G = 9 + 10 = 19) of the bleed steam leading to the main steam mass flow rate (G = 100) at the reactor outlet, the high-pressure steam system 22A in the second operating cycle. The ratio (13/100) of the mass flow rate (G = 3 + 10 = 13) of the extracted steam from the reactor to the mass flow rate (G = 100) of the main steam at the reactor outlet is reduced (19/100 → 13/100), Water heater It is low in the first operating cycle the temperature of the feedwater discharged (H = 924) than the second operation cycle (H = 832) from. As will be described later, the degree of lowering the temperature of the feed water in the second operation cycle from the first operation cycle is 1 ° C. to 40 ° C.

Further, when the operation method according to the present embodiment is changed from the viewpoint, the mass flow rate of the extraction steam extracted from the high-pressure steam system 22A in the first operation cycle (FIG. 2) and led to the feed water heater 9 (G = 9 + 10 = 19) On the other hand, in the second operation cycle (FIG. 1 ), the mass flow rate (G = 3 + 10 = 13) of the extracted steam from the high-pressure steam system 22A is decreased (G = 19 → 13) and discharged from the feed water heater 9. The temperature of the feed water is made lower in the second operation cycle than in the first operation cycle. In addition, by reducing the amount of temperature increase in the high-pressure feed water heater 9 in the second operation cycle (FIG. 1) relative to the amount of temperature increase in the high-pressure feed water heater 9 in the first operation cycle (FIG. 2), It can also be said that the temperature of the feed water discharged from the feed water heater 9 is lower in the second operation cycle (H = 832) than in the first operation cycle (H = 924).

  The operation method according to this embodiment will be described in more detail.

  FIG. 4 shows the relationship between the operation cycle, the reactor heat output, the main steam flow rate (the amount of steam flowing into the main steam pipe 2 from the reactor pressure vessel 1), and the amount of extracted steam when using this embodiment. It is shown in contrast with the output method. One operation cycle is defined as a period from when the reactor operation is stopped to when the reactor operation is stopped for fuel replacement.

  In the operation cycle shown in FIG. 4, the Nth operation cycle is before the power increase method of the present invention is applied, and at this time, the reactor heat output is Q = 100%. An example of the heat balance before this increased output is shown in FIG. In the (N + 1) th operation cycle, the reactor heat output is increased by 5% to Q = 105%. As a means for increasing the reactor thermal output, the control rod extraction amount in the (N + 1) th cycle is made larger than that in the Nth cycle, or the core flow rate in the (N + 1) th cycle is increased. Thus, it can be realized by a method of making it larger than the Nth cycle or changing the type of the fuel assembly. Moreover, since the temperature of the feed water supplied to the reactor pressure vessel 1 decreases when the present invention is applied, it can be expected that the reactor thermal output will naturally increase by coolant density feedback due to the decrease in the core inlet coolant temperature. Depending on the plant, the extraction flow rate and main steam flow rate during one operation cycle may be changed as shown in FIG. FIG. 5 shows an example in the (N + 1) th cycle when the reactor thermal output decreases during one operation cycle due to a decrease in the reactivity of the core and the extraction flow and main steam flow decrease accordingly (coast down). It is. Even when the control rod insertion degree in the core is changed other than the operation cycle as shown in FIG. 5, the output of the nuclear reactor temporarily decreases. Therefore, in this embodiment, the heat balance, extraction flow rate, main steam flow rate, core flow rate, feed water temperature, reactor heat output and feed water heating amount, etc. are started / stopped, and the core heat output is changed by operating the control rod. Comparison shall be made at the operating point at which the main steam flow rate is maximum during the operation cycle excluding accidents and transients and during test operation. In other words, this means the operating point at which the reactor thermal power is maximized during the operating cycle. In addition, as shown in FIG. 6, when the (N-1) cycle is 100% heat output and the output is greatly deviated from the 100% rated output for some reason in the Nth cycle, the (N-1) cycle is The cycle before the present invention is applied (first operation cycle), and the (N + 1) th operation cycle is the cycle to which the present invention is applied (second operation cycle).

  When the reactor heat output is increased, it is necessary to increase the feed water flow rate in order to take the increased amount of heat, or to increase the enthalpy difference of the coolant at the inlet and outlet of the reactor pressure vessel. The conventional method of increasing power uses the former method, and increases the feed water flow rate in proportion to the reactor heat output. FIG. 3 shows an example of heat balance by a conventional power increase method. As a result, in the conventional power increase method, the main steam flow rate in the (N + 1) th operation cycle shown in FIG. 4 is 105%. The present invention is characterized in that the latter method is adopted and the enthalpy difference between the reactor pressure vessel inlet and the outlet is expanded by intentionally lowering the coolant enthalpy at the reactor pressure vessel inlet. In order to lower the coolant enthalpy at the reactor vessel inlet, the amount of steam extracted from the steam system 22 and sent to the feed water heater 9 can be reduced. However, if the amount of extraction is simply reduced as a whole, the thermal efficiency is greatly reduced. Therefore, the amount of power generation cannot be increased so much. In order to suppress a decrease in thermal efficiency, the extraction from the high-pressure steam system 22 </ b> A including the high-pressure turbine 3 to the inlet of the low-pressure turbine 5 may be selectively reduced. This is because the energy of the steam of the high pressure steam system 22A is larger than the steam of the low pressure steam system 22B from the inlet of the low pressure turbine 5 to the inlet of the condenser 6, and the extraction from the high pressure steam system 22A is selectively reduced. This is because there is little thermal loss and it is possible to suppress a decrease in thermal efficiency at the time of increased output. In this embodiment, in order to selectively reduce the extraction from the high-energy portion of the high-pressure steam system 22A and suppress the decrease in thermal efficiency, in the middle of the high-pressure turbine 3 or the outlet of the high-pressure turbine 3 (actually, from the high-pressure turbine outlet) The amount of bleed from the inlet of the separator 4 is selectively reduced, and the amount of steam flowing to the low-pressure turbine 5 is increased to increase the amount of power generation. Since most of the extracted steam from the middle of the high-pressure turbine or from the outlet of the high-pressure turbine is used in the feed water heater 9 installed on the downstream side of the main feed water pump 8, the power increase pump according to the present invention changes the view. 8 is a method of reducing the heating amount of the feed water on the downstream side.

  In the case of a plant with a small amount of original extracted steam from the middle of the high-pressure turbine or from the outlet of the high-pressure turbine, it is necessary to reduce the amount of extracted steam from the low-pressure turbine 5 in order to sufficiently reduce the feed water temperature. Even when the present invention is applied to such a plant, a certain degree of effect can be obtained if the amount of extracted steam from the high-pressure turbine 3 and the outlet of the high-pressure turbine 3 is greatly reduced. In this embodiment, the main steam flow rate can be the same as that of the Nth cycle, although the reactor heat output is increased by 5% compared to the Nth cycle. Since the present embodiment shows an ideal power increase method, the main steam flow rates of the Nth operation cycle and the (N + 1) th operation cycle are the same, but they are not necessarily exactly the same. The main steam flow rate may be increased within the range of the design margin of the equipment to be included.

The extraction point for reducing the amount of extraction is the extraction point at the middle of the high-pressure turbine or at the outlet of the high-pressure turbine. When there are a plurality of extraction points, the most effective is when the most upstream extraction point is selected. In this case, the bleed pipe flow rate adjustment valve 10 for controlling the bleed amount may be installed to reduce the bleed amount, but at least one of the bleed pipes may be completely blocked. As a closing method, a closing valve may be installed in the middle of the bleed pipe or the bleed pipe may be plugged. When the bleed pipe is completely closed, the bleed amount control system equipment becomes unnecessary and the operation control is simplified. Whether the amount of extraction is controlled or whether the extraction pipe is completely closed depends on the heat balance and the increased output width of the plant. (If the amount of extraction per extraction tube is too large, the water supply temperature will be too low if it is completely closed, so the amount of extraction is adjusted in this case.)
According to the present embodiment, even when the reactor heat output is increased and the power generation amount of the nuclear power plant is increased, the increase in the feed water flow rate and the main steam flow rate can be suppressed. Therefore, it is possible to suppress an increase in load on the reactor internal structure. Compared with a case where the amount of extraction is simply reduced as a whole, a decrease in thermal efficiency can be suppressed, and a larger electrical output can be obtained. In addition, the high pressure turbine 3 generally needs to be replaced at the time of a large increase in power by the conventional power increasing method. However, if the present embodiment is used, the power increasing range that can be implemented without replacing the high pressure turbine 3 is the conventional method. Compared to enlarge.

  According to the operation method of the present embodiment, the feed water temperature decreases. Lowering the feed water temperature lowers the coolant temperature at the core inlet and increases the thermal margin of the core (corresponding to MCPR in the case of BWR), so there is also a safety advantage compared to the conventional power increase method. is there. At normal power increase, the core pressure loss increases and the stability margin decreases, but the power increase method according to the present invention also reduces the core void fraction and void reactivity coefficient due to a decrease in the coolant temperature at the core inlet. Moreover, the pressure loss in the core is reduced, and the decrease in the stability margin of the core is also suppressed. In addition, since the void ratio and void reactivity coefficient of the core are reduced, the design margin for pressure rise transients is also increased.

  Thus, the reduction in the feed water temperature has the effect of suppressing the reduction in the core characteristics and design margin of the boiling water light water reactor at the time of increased output. In general, in boiling water type light water reactors, the feed water temperature is not particularly controlled, so the change in the heat balance of the entire plant, specifically, the coolant that condenses steam in the condenser 6 of FIG. 1 (in many cases, seawater) Even with the same boiling water type light water reactor and the same core thermal power, the temperature can be changed within a range of less than 1 ° C. In this embodiment, the drop in the feed water temperature is about 20 ° C. However, if the feed water temperature is within a range to reduce the fall of the core characteristics at the time of increased output, it is greater than the fluctuation width of the feed water temperature during normal operation. If the feed water temperature is lowered by 1 ° C. or more, the effect described in this example can be obtained significantly. However, the feed water intersects with water at the saturation temperature in the reactor when entering the reactor pressure vessel 1. Therefore, there is a temperature difference between the water supply pipe 24 and the pressure vessel 1. If the feed water temperature is lowered too much, the temperature difference in this part becomes large, and the design limit may be exceeded from the viewpoint of thermal fatigue. From this point of view, the limit of the decrease in water supply temperature from the current operating temperature is 40 ° C.

  The decrease in the core pressure loss means that an increase in burden due to increased output to the jet pump and the recirculation pump for recirculation of the coolant can be suppressed. The increase in steam generated in the core is also smaller than the increase in heat output, so there is little impact on carry-under caused by the entrainment of steam in the recirculated water, and even when there is a significant increase in power It is easy to secure the window.

  Table 1 shows the relationship between the reactor heat output, the main steam flow rate, the extraction flow rate, and the enthalpy of water supply when the power increase method according to this embodiment is applied to various power increase amounts. The reactor heat output and main steam flow rate indicate the ratio with respect to the reactor heat output of 100%, and the extraction flow rate indicates the ratio with respect to the main steam flow rate with the reactor heat output of 100%. As can be seen from Table 1, the power increase method of the present invention is widely applicable even when the reactor heat output is 110%. The reason why only the output of 110% is shown in Table 1 is that the moisture separator 4 needs to be replaced when the output is further increased, so that the moisture separator 4 can be replaced or the core pressure can be changed. If combined with an increase or the introduction of a moisture separator superheater, it can be applied to a wider range.

  In general, in a boiling water type light water reactor, the reactor thermal output up to about 102% can be implemented only by improving the measurement accuracy of a feed water flow meter and the like, and the present invention exceeds the reactor thermal output of 102%. Great effect for increased output. Furthermore, when the reactor thermal output is increased to about 105%, generally no significant system equipment change such as replacement of the high-pressure turbine 3 is required. If the present invention is used, it is not necessary to replace the high-pressure turbine 3 even when the reactor thermal output exceeds 105%.

  Now, focusing on the fact that, as described above, with a normal increased output up to about 105% of the reactor thermal output, it is generally unnecessary to change the system equipment such as replacement of the high-pressure turbine 3, the reactor thermal output The application method of this embodiment may be changed with an increase of 5% or less and an increase of more than 5%. Specifically, if the reactor heat output increase width is 5% or less, it is not necessary to replace the high-pressure turbine 3, so the main purpose is to make the above-mentioned core characteristics (core average void ratio) the same as before power increase. To do. Now, assuming that the reactor thermal output before power increase is Q1 (kW), the reactor heat output after power increase is Q2 (kW), and the power output width is A (%), A ≦ 5 for power output below 5% It becomes. The core flow after power increase is W (kg / s), the operating pressure of a typical boiling water light water reactor is 7 MPa, and the constant pressure specific heat at around 200 ° C is about 4.5 (kJ / kg ・ K). Since the ratio of the feed water flow rate to the core flow rate in a typical boiling water light water reactor is about 13%, the conditions for making the core average void ratio the same as before the increased output are as follows.

The amount of heat change in the water supply is
W (kg / s) x 13 (%) / 100 (%) x 4.5 (kJ / kg · K) = W x 13/100 x 4.5 (kW / K)
It is. Now, let T1 be the feed water temperature before increasing output and T2 be the feed water temperature after increasing output. The feed water temperature T2 to reduce the feed water calorie equivalent to the increased output (Q2-Q1) (kW) is More.

Q2-Q1 = W × 13/100 × 4.5 × (T1-T2)
Here, in order to make the core characteristic equal to or higher than that before the increased output, it is only necessary to reduce the amount of heat of the feed water beyond the increased amount of heat. The condition for this is as follows.

Q2-Q1 ≦ W × 13/100 × 4.5 × (T1-T2)
If this equation is transformed,
T2 ≦ T1-7.7 × (Q2-Q1) / (4.5 × W)
It becomes. In other words, if the feed water temperature satisfies the above equation at an output increase of 5% or less, the core thermal margin, pressure loss characteristics, stability margin, transient design margin, etc. are basically the same as before the increase output. It will be better. In addition, since the main steam flow rate is the same as or less than that before the increased output, the design margin of the main steam system such as the high-pressure turbine and the steam dryer can be equal to or more than that before the increased output.

  Furthermore, the water supply temperature when the increase in reactor heat output exceeds 5% and below 10% is described below. When the reactor heat output increases by more than 5% with the conventional power increase, the main steam flow also increases by more than 5%, which generally exceeds the design margin of equipment such as high-pressure turbines. Necessary. In this case, it is possible to suppress the increase in the main steam flow rate to 5% or less by lowering the feed water temperature as in this embodiment.

Now, when an increase of 5% in the main steam flow rate is recognized, the amount of decrease in the amount of feed water required when the increased output width is A (%) corresponds to the increased output of (A-5) (%)
Q2 × (A-5 + 100) / 100-Q1 = Q2 × (A + 95/100) -Q1
It becomes. Assuming that the water supply temperature before the increased output is T1, and the water supply temperature after the increased output is T2, T1 and T2 may satisfy the following equations in order to cancel out this amount of heat.

Q2 × (A + 95/100) -Q1 = W × 13/100 × 4.5 × (T1-T2)
In order to increase the main steam flow rate to 5% or less,
Q2 × (A + 95/100) -Q1 ≦ W × 13/100 × 4.5 × (T1-T2)
You can change the above formula,
T2 ≦ T1-7.7 × (Q2 × (A + 95) / 100-Q1) / (4.5 × W)
Get. This is over 5%, and for increased output of 10% or less, if the water supply temperature T2 after the increased output satisfies the above equation, the increased output can be made within the design margin of a high-pressure turbine, etc. In addition, the design margin of the high-pressure turbine and core is equal to or higher than that of the conventional 5% output increase.

  When the output exceeds 10%, boiling water type light water reactors generally require replacement of a moisture separator in addition to a high pressure turbine. Therefore, as in the case described above, the feed water temperature may be lowered so that the increase in the main steam flow rate is 10% or less so that the moisture separator does not need to be replaced. The equation in this case is as follows for the increased output width A (%), considering the above-mentioned increased output exceeding 5% and 10% or less.

Q2 × (A-10 + 100) /100-Q1≦W×13/100×4.5× (T1-T2)
Transforming the above equation,
T2 ≦ T1-7.7 × (Q2 × (A + 90) / 100-Q1) / (4.5 × W)
It becomes. This indicates a decrease in the feed water temperature at which the moisture separator does not need to be replaced at an increased output exceeding 10%. In this case, the design margin of the high-pressure turbine and core is equal to or greater than that of the conventional 10% increase output.

  In addition, it is not realistic that the decrease in the feed water temperature exceeds 40 ° C from the viewpoint of thermal fatigue even when the output is 5% or less, the output is 5% or more and 10% or less and the output is over 10%. . In addition, as the output increases, the thermal margin of the core decreases. It is said that the core has a thermal margin of up to about 20% by adopting new fuel. Furthermore, improvement of the pump for increasing the core flow rate can be considered, but even if these are taken into consideration, the increased output width is considered to be a limit in terms of the core characteristics of about 30%. Moreover, since the increase in output of 30% or more from the equipment side exceeds the design limit of the low-pressure turbine or condenser higher than the high-pressure turbine, it is not practical because these equipments must be replaced.

  According to the present embodiment, an increase in output of 5% or less can be increased while maintaining a design margin of a high-pressure turbine or core equal to or higher than that before the increase. With an increase in output exceeding 5% and less than 10%, the design margin of the high-pressure turbine and core can be increased to the same level as the conventional 5% increase, and the replacement of equipment such as the high-pressure turbine can be reduced to an increase of 10%. Yes. When the output exceeds 10%, the design margin of the moisture separator and core can be equal to or higher than the conventional 10% increase, and the output exceeds 10% without replacing the moisture separator. Is possible.

  The present invention is not limited to the above-described embodiments, and various modifications can be made. For example, in a boiling water type light water reactor system, a moisture separation superheater or a moisture separation reheater 11 shown in FIG. 7 may be used instead of the moisture separator 4. In this case, the steam extraction line 31 and the drain line 32 are added, but there are no major changes in the main parameters such as the feed water temperature and the main steam flow rate at the time of increased output, and the operation method of the present invention is the same as in the above embodiment. Can be applied, and in that case, the same effect can be obtained.

  Moreover, although the said embodiment is a thing at the time of applying this invention to a boiling water type light water reactor system, this invention is applicable also to a pressurized water type light water reactor system.

BRIEF DESCRIPTION OF THE DRAWINGS It is the whole schematic which shows the whole structure of the boiling water type light water reactor system concerning one Example of this invention with the heat balance of the operating method by one Example of this invention. It is the whole schematic figure which shows the heat balance of the boiling water type light water reactor system before increase output. It is the whole schematic diagram which shows the heat balance of the boiling water type light water reactor system at the time of the conventional increase output method application. It is a schematic diagram which shows the relationship between an operation cycle, a reactor heat output, a main steam flow rate, and an extraction flow rate. It is a schematic diagram which shows the relationship between an operation cycle, a reactor heat output, a main steam flow rate, and an extraction flow rate. It is a schematic diagram which shows the relationship between an operation cycle, a reactor heat output, a main steam flow rate, and an extraction flow rate. It is a whole schematic diagram showing the whole composition of a boiling water type light water reactor system provided with a moisture separation superheater or a moisture separation reheater.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 ... Reactor pressure vessel 2 ... Main steam pipe 3 ... High pressure turbine 4 ... Moisture separator 5 ... Low pressure turbine 6 ... Condenser 7 ... Low pressure feed water heater 8 ... Main feed water pump 9 ... High pressure feed water heater 10 ... Extraction Tracheal flow control valve 11 ... moisture separation superheater or moisture separation reheater 21 ... reactor 22 ... steam system (22A ... high-pressure steam system, 22B ... low-pressure steam system)
23 ... Water supply system 24 ... Water supply pipe 25-28 ... Extraction pipe 31 ... Steam extraction line 32 ... Drain line

Claims (13)

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 of operating a nuclear power plant comprising a feed water heater that heats feed water supplied from the condenser, and a feed water system that guides feed water discharged from the feed water heater toward the nuclear reactor,
Increasing the second reactor heat output in the second operating cycle of the reactor above the first reactor heat output in the first operating cycle prior to the second operating cycle;
The ratio of the extracted steam mass flow rate extracted from the high pressure steam system in the first operating cycle to the feed water heater with respect to the mass flow rate of the main steam at the reactor outlet is compared with the high pressure steam in the second operating cycle. The ratio of the mass flow rate of the bleed steam from the system to the mass flow rate of the main steam at the reactor outlet is decreased, and the temperature of the feed water discharged from the feed water heater is set to 1 in the second operation cycle from the first operation cycle. A method for operating a nuclear power plant, characterized in that the temperature is lowered by 40 ° C. or more. 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 of operating a nuclear power plant comprising a feed water heater that heats feed water supplied from the condenser, and a feed water system that guides feed water discharged from the feed water heater toward the nuclear reactor,
Increasing the second reactor heat output in the second operating cycle of the reactor above the first reactor heat output in the first operating cycle prior to the second operating cycle;
The ratio of the mass flow rate of the extracted steam extracted from the high-pressure steam system and the low-pressure steam system to the feed water heater in the first operation cycle with respect to the mass flow rate of the main steam at the reactor outlet, In the operation cycle, the ratio of the mass flow rate of the extracted steam from the high-pressure steam system and the low-pressure steam system to the mass flow rate of the main steam at the reactor outlet is decreased, and the ratio from the high-pressure steam system in the second operation cycle is reduced. The ratio of the ratio of the mass flow rate of the extracted steam to the mass flow rate of the main steam at the reactor outlet with respect to the first operating cycle is defined as the ratio of the mass flow rate of the extracted steam from the low-pressure steam system in the second operating cycle. The ratio with respect to the mass flow rate of the main steam at the outlet is made larger than the decreasing ratio with respect to the first operation cycle, and the temperature of the feed water discharged from the feed water heater is set to the first operation. Cycle 1 ℃ or more than the second operating cycle, the method of operating a nuclear plant, characterized in that lower 40 ° C. or less. 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 of operating a nuclear power plant comprising a feed water heater that heats feed water supplied from the condenser, and a feed water system that guides feed water discharged from the feed water heater toward the nuclear reactor,
Increasing the second reactor heat output in the second operating cycle of the reactor above the first reactor heat output in the first operating cycle prior to the second operating cycle;
In contrast to the mass flow rate of the extraction steam extracted from the high-pressure steam system in the first operation cycle and led to the feed water heater, the mass flow rate of the extraction steam from the high-pressure steam system is decreased in the second operation cycle, A method for operating a nuclear power plant, characterized in that the temperature of feed water discharged from the feed water heater is lower by 1 ° C. or more and 40 ° C. or less in the second operation cycle than in the first 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 of operating a nuclear power plant comprising a feed water heater that heats feed water supplied from the condenser, and a feed water system that guides feed water discharged from the feed water heater toward the nuclear reactor,
Increasing the second reactor heat output in the second operating cycle of the reactor above the first reactor heat output in the first operating cycle prior to the second operating cycle;
In contrast to the mass flow rate of extraction steam extracted from the high-pressure steam system and the low-pressure steam system in the first operation cycle and guided to the feed water heater, in the second operation cycle, from the high-pressure steam system and the low-pressure steam system And reducing the mass flow rate of the extracted steam from the high-pressure steam system in the second operating cycle with respect to the first operating cycle in the second operating cycle. The mass flow rate of the extracted steam from the system is greater than the rate of decrease with respect to the first operating cycle, and the temperature of the feed water discharged from the feed water heater is 1 ° C. or more in the second operating cycle from the first operating cycle, A method for operating a nuclear power plant, wherein the temperature is lowered by 40 ° C. or less. 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 steam discharged from the low-pressure turbine;
A low-pressure feed water heater installed on the downstream side of the condenser and upstream of the main feed water pump for heating the feed water supplied from the condenser; and on the downstream side of the main feed water pump and the reactor In a method for operating a nuclear power plant comprising a high-pressure feed water heater installed on the upstream side, and a feed water system for guiding feed water discharged from the high-pressure feed water heater toward the nuclear reactor,
Increasing the second reactor heat output in the second operating cycle of the reactor above the first reactor heat output in the first operating cycle prior to the second operating cycle;
Relative to the mass flow rate of extracted steam which bled before Ki蒸 exhaust system leading to the high-pressure feed water heater in the first operating cycle, the mass flow rate of extracted steam from Ki蒸 exhaust system before the second operation cycle By reducing the temperature increase amount in the high-pressure feed water heater in the first operation cycle, the temperature increase amount in the high-pressure feed water heater is reduced in the second operation cycle, thereby reducing the feed water heating. A method for operating a nuclear plant, characterized in that the temperature of the feed water discharged from the vessel is lower by 1 ° C. or more and 40 ° C. or less in the second operation cycle than in the first operation cycle. A nuclear reactor,
One of a moisture separator, a moisture separation superheater, and a moisture separation reheater is provided between the high pressure turbine and the low pressure turbine, the high pressure turbine being supplied with steam generated in the nuclear reactor, and the low pressure turbine. A steam system having one;
A condenser for condensing steam discharged from the low-pressure turbine;
A feed water heater that heats feed water supplied from the condenser, and that feeds the feed water discharged from the feed water heater toward the reactor;
Steam is extracted from the middle of the high-pressure turbine, downstream from the high-pressure turbine outlet and upstream from any one of the moisture separator, the moisture separation superheater, or the moisture separation reheater, In a method for operating a nuclear power plant comprising at least one high-pressure extraction pipe leading to the feed water heater,
Increasing the second reactor heat output in the second operating cycle of the reactor above the first reactor heat output in the first operating cycle prior to the second operating cycle;
In at least one system of the high-pressure extraction pipe in which the extraction steam has flowed in the first operation cycle, the extraction steam is stopped in the second operation cycle or the extraction steam amount adjusting means is provided to adjust the extraction steam amount. Thereby, the temperature of the feed water discharged from the feed water heater is lowered by 1 ° C. or more and 40 ° C. or less in the second operation cycle from the first operation cycle. In the operating method of the nuclear power plant of any one of Claims 1-4 and 6 ,
The second reactor thermal output in the second operating cycle of the reactor is Q2 (kW), the first reactor thermal output in the first operating cycle prior to the second operating cycle is Q1 (kW), and the feed water heater The temperature of the first operating cycle of the feed water discharged from the reactor is T1 (° C), the temperature in the second operating cycle is T2 (° C), and the core flow rate flowing into the reactor in the second operating cycle is W ( kg / s) and the increased output width is A (%)
0 <A ≦ 5,
T2 ≦ T1-7.7 × (Q2-Q1) / (4.5 × W),
A method for operating a nuclear power plant that satisfies the following relationship. In the operating method of the nuclear power plant of any one of Claims 1-4 and 6 ,
The second reactor thermal output in the second operating cycle of the reactor is Q2 (kW), the first reactor thermal output in the first operating cycle prior to the second operating cycle is Q1 (kW), and the feed water heater The temperature of the first operating cycle of the feed water discharged from the reactor is T1 (° C), the temperature in the second operating cycle is T2 (° C), and the core flow rate flowing into the reactor in the second operating cycle is W ( kg / s) and the increased output width is A (%)
5 <A ≦ 10,
T2 ≦ T1-7.7 × (Q2 × (A + 95) / 100-Q1) / (4.5 × W)
A method for operating a nuclear power plant that satisfies the following relationship. In the operating method of the nuclear power plant of any one of Claims 1-4 and 6 ,
The second reactor thermal output in the second operating cycle of the reactor is Q2 (kW), the first reactor thermal output in the first operating cycle prior to the second operating cycle is Q1 (kW), and the feed water heater The temperature of the first operating cycle of the feed water discharged from the reactor is T1 (° C), the temperature in the second operating cycle is T2 (° C), and the core flow rate flowing into the reactor in the second operating cycle is W ( kg / s) and the increased output width is A (%)
10 <A <30,
T2 ≦ T1-7.7 × (Q2 × (A + 90) / 100-Q1) / (4.5 × W)
A method for operating a nuclear power plant that satisfies the following relationship.   In the operation method of the nuclear power plant of Claim 5,
  The second reactor heat output in the second operation cycle of the reactor is Q2 (kW), the first reactor heat output in the first operation cycle before the second operation cycle is Q1 (kW), and the high-pressure feed water heating The temperature of the first operating cycle of the feed water discharged from the reactor is T1 (° C), the temperature in the second operating cycle is T2 (° C), and the core flow rate flowing into the reactor in the second operating cycle is W (Kg / s) and the increased output width is A (%)
    0 <A ≦ 5,
    T2 ≦ T1-7.7 × (Q2-Q1) / (4.5 × W),
A method for operating a nuclear power plant that satisfies the following relationship.   In the operation method of the nuclear power plant of Claim 5,
  The second reactor heat output in the second operation cycle of the reactor is Q2 (kW), the first reactor heat output in the first operation cycle before the second operation cycle is Q1 (kW), and the high-pressure feed water heating The temperature of the first operating cycle of the feed water discharged from the reactor is T1 (° C), the temperature in the second operating cycle is T2 (° C), and the core flow rate flowing into the reactor in the second operating cycle is W (Kg / s) and the increased output width is A (%)
    5 <A ≦ 10,
    T2 ≦ T1-7.7 × (Q2 × (A + 95) / 100-Q1) / (4.5 × W)
A method for operating a nuclear power plant that satisfies the following relationship.   In the operation method of the nuclear power plant of Claim 5,
  The second reactor heat output in the second operation cycle of the reactor is Q2 (kW), the first reactor heat output in the first operation cycle before the second operation cycle is Q1 (kW), and the high-pressure feed water heating The temperature of the first operating cycle of the feed water discharged from the reactor is T1 (° C), the temperature in the second operating cycle is T2 (° C), and the core flow rate flowing into the reactor in the second operating cycle is W (Kg / s) and the increased output width is A (%)
    10 <A <30,
    T2 ≦ T1-7.7 × (Q2 × (A + 90) / 100-Q1) / (4.5 × W)
A method for operating a nuclear power plant that satisfies the following relationship. The operation method of an atomic weight plant according to any one of claims 1 to 12 , wherein the first operation cycle is an initial operation cycle after the installation of the nuclear power plant, and the second operation cycle is after the first operation cycle. The operation method of the nuclear power plant, wherein at least one operation cycle is interposed between the first operation cycle and the second operation cycle.

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