Pressurized water reactor nuclear energy steam supply system and method

Abstract

A pressurized water reactor nuclear energy steam supply system (101), comprising: a pressurized water reactor nuclear power unit (11), which is connected to a first thermal treatment loop and outputs, to the first thermal treatment loop, steam to be treated; a steam generator (12), comprising a pipe side path and a shell side path that are isolated from each other, wherein the pipe side path is located in the first thermal treatment loop, the shell side path is located in a second thermal treatment loop independent of the first thermal treatment loop, and the steam generator (12) can use said steam the first thermal treatment loop as a heat source input to generate clean steam in the second thermal treatment loop; and a steam compressor device (13), which is connected to the second thermal treatment loop, wherein the steam compressor device (13) can treat the clean steam in the second thermal treatment loop to increase the pressure and / or temperature of the clean steam. The solution can produce medium-pressure superheated steam or high-pressure superheated steam, so as to satisfy the industrial steam parameter requirements of a chemical industry base, thereby achieving the purpose of comprehensively utilizing the nuclear energy of the pressurized water reactor nuclear power unit.

Description

Pressurized water reactor nuclear energy steam supply system and method

[0001] CROSS-REFERENCE TO RELATED APPLICATIONS

[0002] This application claims priority to Chinese patent application 202311775088.X, filed on December 21, 2023, entitled “Pressurized Water Reactor Nuclear Energy Steam Supply System and Method,” the entire contents of which are incorporated herein by reference. Technical Field

[0003] The present invention relates to the field of pressurized water reactor nuclear energy applications, and in particular to a pressurized water reactor nuclear energy steam supply system and method. Background Art

[0004] Currently, nuclear energy industrial steam supply projects are in the preliminary research or early implementation stages, primarily encompassing two technical approaches. The first utilizes PWR main steam as a heat source to heat the demineralized water in the PWR's tertiary circuit to produce low-pressure superheated steam. PWR-provided industrial steam parameters, for example, are 2.5 MPa at 250°C or 1.8 MPa at 248°C. However, this approach, due to the steam parameter limitations of PWR nuclear power plants, can only provide low-pressure superheated steam.

[0005] On the other hand, existing technologies have proposed combining pressurized water reactors (PWRs) and high-temperature gas-cooled reactors (HTGRs), leveraging the economic benefits of PWRs and the high steam temperatures of HTGRs to provide medium- and low-pressure superheated steam. However, this approach has drawbacks: the technical maturity of HTGRs still requires further improvement and verification, the construction cost is high, the economics are poor, the system is complex, and the ability to supply high-temperature, high-pressure industrial steam from nuclear power plants has yet to be realized.

[0006] Therefore, the field urgently needs a nuclear energy steam supply solution that takes into account both economy and system complexity and can effectively break through the steam parameter limitations of pressurized water reactor nuclear power units.

[0007] Summary of the Invention

[0008] In response to the deficiencies in the prior art, the present invention provides a pressurized water reactor nuclear energy steam supply system and method, which can utilize the thermal energy of a pressurized water reactor nuclear power unit to generate medium-pressure or high-pressure superheated steam to meet user needs.

[0009] A first aspect of the present invention provides a pressurized water reactor nuclear energy steam supply system, characterized by comprising:

[0010] a pressurized water reactor nuclear power unit connected to the first heat treatment loop and outputting steam to be treated into the first heat treatment loop;

[0011] A steam generator comprising a tube-side path and a shell-side path isolated from each other, wherein the tube-side path is located in a first heat treatment circuit and the shell-side path is located in a second heat treatment circuit independent of the first heat treatment circuit, and the steam generator is capable of inputting the steam to be treated in the first heat treatment circuit as a heat source to generate clean steam in the second heat treatment circuit;

[0012] The steam compressor device is connected to the second heat treatment circuit, and the steam compressor device can process the clean steam in the second heat treatment circuit to increase the pressure and / or temperature of the clean steam.

[0013] Preferably, the pressurized water reactor nuclear energy steam supply system also includes a feed water heater and a feed water preheater, both of which are connected to the first heat treatment loop and the second heat treatment loop to perform heat exchange between the fluid in the first heat treatment loop and the fluid in the second heat treatment loop.

[0014] Preferably, in the first heat treatment loop, the feed water heater and the feed water preheater are located between the steam generator and the pressurized water reactor nuclear power unit in sequence.

[0015] Preferably, the pressurized water reactor nuclear energy steam supply system also includes a deaerator and a desalted water tank, which are connected to the second heat treatment loop. In the second heat treatment loop, the desalted water tank is located at the starting end of the second heat treatment loop and is connected to the feed water preheater.

[0016] Preferably, the pressurized water reactor nuclear energy steam supply system further comprises at least one booster pump located in the second heat treatment loop, the booster pump being located between the desalted water tank and the feedwater preheater and / or between the deaerator and the feedwater heater.

[0017] Preferably, the pressurized water reactor nuclear energy steam supply system further comprises a superheat heater and / or an electric heater, and the superheat heater and / or the electric heater are arranged before the steam compressor device in the second heat treatment loop.

[0018] Preferably, the steam compressor device includes an electric steam compressor or a steam-driven steam compressor, and includes a single-cylinder structure or a double-cylinder structure.

[0019] Preferably, when the pressure ratio of the steam compressor device belongs to the first pressure ratio range, the steam compressor device is a double-cylinder structure, and when the pressure ratio of the steam compressor device belongs to the second pressure ratio range, the steam compressor device is a single-cylinder structure, wherein the pressure ratio is the steam ratio between the outlet end and the inlet end of the steam compressor device, and the upper limit value of the first pressure ratio range is greater than the upper limit value of the second pressure ratio range.

[0020] Preferably, when the shaft power of the steam compressor device belongs to the first shaft power range, the steam compressor device is a steam-driven steam compressor, and when the shaft power of the steam compressor device belongs to the second shaft power range, the steam compressor device is an electric steam compressor, wherein the shaft power is the power required for the steam compressor device to compress clean steam, and the upper limit value of the first shaft power range is greater than the upper limit value of the second shaft power range.

[0021] A second aspect of the present invention provides a pressurized water reactor nuclear energy steam supply method, which uses the pressurized water reactor nuclear energy steam supply system provided by the first aspect of the present invention to deliver clean steam that meets steam parameter requirements to the user side, characterized in that the method includes the following steps:

[0022] operating pressurized water reactor nuclear power plants to export steam for processing;

[0023] Passing the steam to be treated through a first heat treatment circuit, while generating clean steam in a second heat treatment circuit independent of the first heat treatment circuit by a steam generator;

[0024] The clean steam is passed through a steam compressor device to increase the pressure and / or temperature of the clean steam according to steam parameter requirements.

[0025] This application uses pressurized water reactor nuclear energy heat to produce medium-pressure superheated steam or high-temperature and high-pressure superheated steam through a special first heat treatment loop, a second heat treatment loop and a steam compressor connection method to meet the industrial steam parameter requirements of the chemical base, thereby achieving the purpose of comprehensive utilization of nuclear energy of the pressurized water reactor nuclear power unit. BRIEF DESCRIPTION OF THE DRAWINGS

[0026] In order to more clearly illustrate the technical solutions of the embodiments of the present application, the following briefly introduces the drawings required for use in the embodiments of the present application.

[0027] FIG1 is a system block diagram of a pressurized water reactor nuclear energy steam supply system according to a specific embodiment of the present invention;

[0028] 2 to 10 are system block diagrams of pressurized water reactor nuclear energy steam supply systems according to some variant embodiments of the present invention based on the embodiment shown in FIG. 1 ;

[0029] FIG11 is a schematic flow chart of a method for supplying steam to a pressurized water reactor nuclear energy according to a specific embodiment of the present invention.

[0030] Reference numerals:

[0031] 101, 102, 103, 104, 105, 106, 107, 108, 109, 110 - pressurized water reactor nuclear power steam supply system;

[0032] 11- Pressurized water reactor nuclear power unit;

[0033] 12-steam generator;

[0034] 13-steam compressor device; 131-steam compressor; 132-electric compressor;

[0035] 14-user side; 141-medium voltage user; 142-high voltage user;

[0036] 15-feedwater heater;

[0037] 16-deaerator;

[0038] 17-feedwater preheater;

[0039] 18-Desalted water tank;

[0040] 19-boost pump;

[0041] 21-overheat heater;

[0042] 22-Electric heater. DETAILED DESCRIPTION

[0043] In order to better understand the technical solution of the present application, the embodiments of the present application are described in detail below with reference to the accompanying drawings.

[0044] The steam parameters of pressurized water reactor (PWR) nuclear power plants are limited, allowing them to only provide low-pressure superheated steam. The combined configuration of PWRs and high-temperature gas-cooled reactors (HTGRs) is costly, uneconomical, and complex. Furthermore, the ability to supply high-temperature, high-pressure industrial steam from nuclear power plants remains a challenge. This invention aims to utilize the heat from PWR nuclear energy to produce medium-pressure or high-temperature, high-pressure superheated steam using a novel thermal system. This approach meets the industrial steam parameter requirements of chemical bases, thereby achieving the comprehensive utilization of nuclear energy from PWR nuclear power plants.

[0045] FIG1 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 101 according to a specific embodiment of the present invention.

[0046] As shown in Figure 1 , the pressurized water reactor nuclear energy steam supply system 101 of this embodiment includes a pressurized water reactor nuclear power unit 11, a steam generator 12, a steam compressor assembly 13, and a user-side 14 at the final output terminal. To coordinate with the pressurized water reactor nuclear power unit 11 and achieve nuclear energy steam supply, the pressurized water reactor nuclear energy steam supply system 101 in the embodiment of Figure 1 also includes a feedwater heater 15, a deaerator 16, a feedwater preheater 17, and a desalted water tank 18. In this embodiment, the steam compressor assembly 13 is a steam-driven steam compressor 131, and the user-side 14 is a medium-pressure user 141.

[0047] The pressurized water reactor (PWR) nuclear power steam supply system 101 includes a first heat treatment loop and a second heat treatment loop, each of which is independent. The PWR nuclear power unit 11, steam generator 12, feedwater heater 15, and feedwater preheater 17 are connected by pipelines to form the first heat treatment loop. The feedwater preheater 17, deaerator 16, feedwater heater 15, steam generator 12, and steam compressor 13 are connected by pipelines to form the second heat treatment loop. The fluid in the first heat treatment loop exchanges heat with the fluid in the second heat treatment loop in the feedwater preheater 17, feedwater heater 15, and steam generator 12, respectively, ultimately forming medium-pressure or high-pressure superheated steam in the second heat treatment loop to meet user needs.

[0048] The pressurized water reactor nuclear power unit 11 is connected to the first heat treatment loop and outputs untreated steam to the first heat treatment loop. The pressurized water reactor nuclear power unit 11 outputs untreated steam after a nuclear reaction occurs, and the untreated steam only flows through the first heat treatment loop in the pressurized water reactor nuclear energy steam supply system 101. In this embodiment, the first heat treatment loop is the path through which the untreated steam generated after the reaction of the pressurized water reactor nuclear power unit 11 flows after it is output. It should be noted that the untreated steam can be understood as the main steam of the pressurized water reactor nuclear power unit 11, or in the form of steam such as high-pressure cylinder exhaust, high-temperature reheat steam, etc. The gas source form of the first heat treatment loop is not limited in this application. The first heat treatment loop includes a steam generator 12, a feedwater heater 15, and a feedwater preheater 17 in sequence.

[0049] The steam generator 12 in this embodiment specifically includes a tube-side path and a shell-side path, which are isolated from each other. The tube-side path is located within a first heat treatment loop that flows through the pressurized water reactor (PWR) nuclear power unit 11, the steam generator 12, the feedwater heater 15, and the feedwater preheater 17, ultimately returning to the PWR nuclear power unit 11. Correspondingly, the shell-side path is located within a second heat treatment loop that is independent of the first heat treatment loop. In this embodiment, the steam generator 12 is configured to use the medium flowing through the first heat treatment loop as a heat source input to generate clean steam in the second heat treatment loop.

[0050] In the first thermal treatment loop, the feedwater heater 15 and feedwater preheater 17 are sequentially located between the steam generator 12 and the pressurized water reactor 11. The steam to be treated output by the pressurized water reactor 11 passes through the steam generator 12, the feedwater heater 15, and the feedwater preheater 17, and then exchanges heat with the feedwater in the second thermal treatment loop to form drain water, which is then returned to the pressurized water reactor 11.

[0051] The pressurized water reactor nuclear energy steam supply system 101 also includes a deaerator 16 and a desalted water tank 18, which are connected to a second heat treatment loop. In the second heat treatment loop, the desalted water tank 16 is located at the starting end of the second heat treatment loop and is connected to a feed water preheater 17 to supply water to the feed water preheater 17.

[0052] In the second thermal treatment loop, the feedwater preheater 17 is connected between the desalted water tank 18 and the deaerator 16, and the feedwater heater 15 is connected between the deaerator 16 and the steam generator 12. The feedwater in the desalted water tank 18 passes through the feedwater preheater 17, the deaerator 16, the feedwater heater 15, and the steam generator 12 in sequence, where it is heat-exchanged with the steam to be treated in the first thermal treatment loop and converted into low-pressure or medium-pressure saturated steam. The fluid is ultimately delivered to the steam compressor 13.

[0053] Since the first heat treatment loop and the second heat treatment loop are independent of each other, the radioactivity of the pressurized water reactor main steam is prevented from entering the steam output by the second heat treatment loop. Therefore, the low-pressure or medium-pressure saturated steam generated by the second heat treatment loop is clean steam.

[0054] The pressurized water reactor nuclear energy steam supply system 101 in this embodiment also includes two booster pumps 19 located in the second heat treatment loop, respectively located between the desalted water tank 18 and the feed water preheater 17, and between the deaerator 16 and the feed water heater 15, but the present application is not limited to this. In other embodiments of the present application, the booster pump 19 can be arranged at one location according to actual conditions.

[0055] The steam compressor device 13 is connected to the second heat treatment circuit, and the steam compressor device 13 can process the clean steam in the second heat treatment circuit to increase the pressure and / or temperature of the clean steam.

[0056] In this embodiment, the steam compressor device 13 has an inlet end a and an outlet end b that are opposite each other. As shown in Figure 1 , the inlet end a is connected to the second heat treatment circuit, and the outlet end b is connected to the user 14 (in this embodiment, the medium-pressure user 141). Specifically, the steam compressor device 13 is configured to process clean steam flowing from the second heat treatment circuit into the inlet end a, thereby increasing the pressure and / or temperature of the clean steam according to the steam parameter requirements of the user 14.

[0057] The pressurized water reactor nuclear energy steam supply system 101 of this embodiment, through the combination of a pressurized water reactor nuclear power unit 11 and a steam compressor device 13, uses the heat of the pressurized water reactor to generate low-pressure or medium-pressure saturated steam, and then uses the steam compressor device 13 to heat the steam into medium-pressure or high-pressure superheated steam, thereby increasing the steam parameters supplied by the pressurized water reactor nuclear power plant from low-pressure superheat parameters to medium-pressure or high-pressure superheat parameters, meeting the industrial steam parameter requirements of the chemical base and realizing the comprehensive utilization of nuclear energy of the pressurized water reactor nuclear power unit 11.

[0058] The steam compressor device 13 includes an electric steam compressor 132 (see FIG. 2 ) or a steam-driven steam compressor 131 , and both the electric steam compressor 132 and the steam-driven steam compressor 131 include a single-cylinder structure or a double-cylinder structure.

[0059] In the embodiment of FIG1 , the steam compressor device 13 is a steam-driven steam compressor 131. The steam-driven steam compressor 131 compresses the steam delivered by the second heat treatment circuit, raising the pressure and / or temperature of the clean steam to generate medium-pressure or high-pressure superheated steam, which is then delivered to the user side 14. In this embodiment, the user side 14 is a medium-pressure user 141. The steam-driven steam compressor 131 outputs medium-pressure superheated steam, which is delivered to the medium-pressure user 141 to meet the steam demand of the medium-pressure user 141.

[0060] FIG2 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 102 according to another specific embodiment of the present invention.

[0061] As shown in Figure 2 , the steam compressor device 13 is an electric steam compressor 132. The electric steam compressor 132 compresses the steam delivered by the second heat treatment loop, raising the pressure and / or temperature of the clean steam to generate medium-pressure or high-pressure superheated steam, which is then delivered to the user side 14. In this embodiment, the user side 14 is a medium-pressure user 141. The electric steam compressor 132 outputs medium-pressure superheated steam and delivers it to the medium-pressure user 141 to meet the steam demand of the medium-pressure user 141.

[0062] . In the present application, preferably, the following strategy can be adopted to determine the adjustment method for the system structure. First, when the shaft power of the steam compressor device 13 belongs to the first shaft power range, the steam compressor device 13 can be selected to be implemented as a steam compressor 131, and when the shaft power of the steam compressor device 13 belongs to the second shaft power range, the steam compressor device 13 is selected to be implemented as an electric steam compressor 132. Among them, the shaft power is the power required for the steam compressor device 13 to compress clean steam, and the upper limit value of the first shaft power range is greater than the upper limit value of the second shaft power range. This means that for the case of a larger pressure ratio, a steam compressor is selected, and for the case of a smaller pressure ratio, an electric steam compressor is selected, wherein the pressure ratio is the steam ratio between the outlet and inlet ends of the steam compressor device.

[0063] Further preferably, the steam compressor device 13 in the above-mentioned multiple embodiments can also be specifically implemented as a single-cylinder structure or a double-cylinder structure. Preferably, when the pressure ratio of the steam compressor device 13 belongs to the first pressure ratio range, the steam compressor device 13 is a double-cylinder structure, and when the pressure ratio of the steam compressor device 13 belongs to the second pressure ratio range, the steam compressor device is a single-cylinder structure, wherein the upper limit value of the first pressure ratio range is greater than the upper limit value of the second pressure ratio range. This means that when the pressure ratio is large, the double-cylinder structure is selected; and when the pressure ratio is small, the single-cylinder structure is selected. Thereby, the clean steam processed by the steam compressor device 13 can be better adapted to the steam parameter requirements of the user side 14. In addition, the driving device of the steam compressor device 13 can be set according to the steam supply load and the steam parameter requirements of the user side, and this application does not impose any restrictions on this.

[0064] In other embodiments, the pressurized water reactor nuclear energy steam supply system also includes a superheater 21 (see Figure 3) and / or an electric heater 22 (see Figure 6), and the superheater 21 and / or the electric heater 22 are arranged before the steam compressor device 13 in the second heat treatment loop.

[0065] Generally speaking, whether to add these devices is determined by considering the superheat requirements of the overall system. When the superheat at the outlet end b of the steam compressor device 13 is relatively high, the superheat requirement at the inlet end of the steam compressor device 13 is relatively high. In this case, it is preferred to add a superheat heater 21 or an electric heater 22 to the corresponding overall system. In this application, based on the steam parameter requirements of different user-side 14, while considering both system complexity and economy, it is preferred to add a superheat heater 21 and / or an electric heater 22 to the existing basic embodiment in different situations to better meet the steam parameter requirements under different operating conditions.

[0066] FIG3 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 103 according to another specific embodiment of the present invention.

[0067] As shown in Figure 3, the pressurized water reactor nuclear energy steam supply system 103 also includes a superheater 21, which is connected to both the first and second heat treatment loops. In the first heat treatment loop, the superheater 21 is located between the pressurized water reactor nuclear power unit 11 and the steam generator 12. In the second heat treatment loop, the superheater 21 is located between the steam generator 12 and the steam compressor 13. After the fluids from the first and second heat treatment loops exchange heat in the steam generator 12, clean steam is generated in the second heat treatment loop. The clean steam then further exchanges heat with the fluid from the first heat treatment loop in the superheater 21 to form superheated steam, which is then transmitted to the steam compressor 13. In this embodiment, the steam compressor 13 is a steam-driven compressor 131, which compresses the superheated steam transported from the second heat treatment loop, raising the pressure and / or temperature of the clean steam to generate medium-pressure or high-pressure superheated steam, which is then transmitted to the user side 14. The user side 14 of this embodiment is a medium-pressure user 141 , and the steam compressor 131 outputs medium-pressure superheated steam, which is transported to the medium-pressure user 141 to meet the steam demand of the medium-pressure user 141 .

[0068] FIG4 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 104 according to another specific embodiment of the present invention.

[0069] The difference between the pressurized water reactor nuclear energy steam supply system 104 shown in FIG4 and the pressurized water reactor nuclear energy steam supply system 103 shown in FIG3 is that the steam compressor 13 is an electric steam compressor 132. The clean steam in the second heat treatment loop is superheated in the superheat heater 21 and then transferred to the electric steam compressor 132.

[0070] The various variant embodiments of the present invention specifically provide for preferred adjustments to the system configuration according to the steam parameter requirements corresponding to different user sides 14 .

[0071] FIG5 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 105 according to another specific embodiment of the present invention.

[0072] The difference between the pressurized water reactor nuclear energy steam supply system 105 shown in FIG5 and the pressurized water reactor nuclear energy steam supply system 101 shown in FIG1 is that the user side 14 is a high-pressure user 142. The steam compressor 131 outputs high-pressure superheated steam and transmits it to the high-pressure user 142 to meet the steam demand of the high-pressure user 142.

[0073] FIG6 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 106 according to another specific embodiment of the present invention.

[0074] As shown in Figure 6, the pressurized water reactor nuclear energy steam supply system 106 also includes an electric heater 22 connected to the second heat treatment loop. In the second heat treatment loop, the electric heater 22 is positioned between the steam generator 12 and the steam compressor 13. After the fluids from the first heat treatment loop and the second heat treatment loop exchange heat within the steam generator 12, clean steam is formed in the second heat treatment loop. The clean steam is then further heated by the electric heater 22 to form superheated steam, which is then transmitted to the steam compressor 13. In this embodiment, the steam compressor 13 is a steam compressor 131. The steam compressor 131 compresses the superheated steam transported from the second heat treatment loop, raising the pressure and / or temperature of the clean steam to generate high-pressure superheated steam, which is then transmitted to the user 14. In this embodiment, the user 14 is a high-pressure user 142. The steam compressor 131 outputs high-pressure superheated steam and transmits it to the high-pressure user 142 to meet its steam needs.

[0075] FIG7 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 107 according to another specific embodiment of the present invention.

[0076] As shown in Figure 7, the pressurized water reactor nuclear power steam supply system 107 also includes a superheater 21 and an electric heater 22. The superheater 21 is connected to both the first and second heat treatment loops. In the first heat treatment loop, the superheater 21 is positioned between the pressurized water reactor nuclear power unit 11 and the steam generator 12. In the second heat treatment loop, the superheater 21 is positioned between the steam generator 12 and the steam compressor 13. The electric heater 22 is connected to the second heat treatment loop. In the second heat treatment loop, the electric heater 22 is positioned between the superheater 21 and the steam compressor 13.

[0077] After the fluid of the first heat treatment circuit and the fluid of the second heat treatment circuit exchange heat in the steam generator 12, clean steam is formed in the second heat treatment circuit. The clean steam then continues to exchange heat with the fluid of the first heat treatment circuit in the superheat heater 21, and is then transferred to the electric heater 22 for further heating to form superheated steam, which is then transferred to the steam compressor device 13. The steam compressor device 13 in this embodiment is a steam compressor 131. The steam compressor 131 compresses the superheated steam delivered by the second heat treatment circuit, increases the pressure and / or temperature of the clean steam, generates high-pressure superheated steam, and then delivers it to the user side 14. The user side 14 in this embodiment is a high-pressure user 142. The steam compressor 131 outputs high-pressure superheated steam and delivers it to the high-pressure user 142 to meet the steam demand of the high-pressure user 142.

[0078] FIG8 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 108 according to another specific embodiment of the present invention.

[0079] The pressurized water reactor nuclear energy steam supply system 108 shown in FIG8 differs from the pressurized water reactor nuclear energy steam supply system 101 shown in FIG1 in that the user side 14 is a high-pressure user 142, and the steam compressor device 13 is an electric steam compressor 132. The electric steam compressor 132 outputs high-pressure superheated steam, which is delivered to the high-pressure user 142 to meet the steam demand of the high-pressure user 142.

[0080] FIG9 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 109 according to another specific embodiment of the present invention.

[0081] The difference between the pressurized water reactor nuclear energy steam supply system 109 shown in FIG9 and the pressurized water reactor nuclear energy steam supply system 106 shown in FIG6 is that the steam compressor unit 13 is an electric steam compressor 132. The electric heater 22 outputs superheated steam and transmits it to the electric steam compressor 132. The electric steam compressor 132 outputs high-pressure superheated steam and transmits it to the high-pressure user 142 to meet the steam demand of the high-pressure user 142.

[0082] FIG10 is a system block diagram of a pressurized water reactor nuclear energy steam supply system 110 according to another specific embodiment of the present invention.

[0083] The difference between the pressurized water reactor nuclear energy steam supply system 110 shown in Figure 10 and the pressurized water reactor nuclear energy steam supply system 107 shown in Figure 7 is that the steam compressor unit 13 is an electric steam compressor 132. The electric heater 22 outputs superheated steam and transmits it to the electric steam compressor 132. The electric steam compressor 132 outputs high-pressure superheated steam and transmits it to the high-pressure user 142 to meet the steam demand of the high-pressure user 142.

[0084] The user side 14 shown in Figures 1 to 4 is an embodiment of a medium-pressure user 141. It utilizes the unprocessed steam output from the pressurized water reactor (PWR) nuclear power unit 11 as a heat source. The unprocessed steam preheats, deoxidizes, and heats the desalted water in the second heat treatment loop. Low-pressure superheated steam is generated by the steam generator 12. A superheater 21 is added as necessary based on the steam parameter requirements of the user side 14. The low-pressure superheated steam is then converted to medium-pressure superheated steam via a steam compressor 131 or an electric steam compressor 132. In this cycle, the main media flow is the unprocessed steam entering the steam generator 12, the feedwater heater 15, and the feedwater preheater 17. After cooling and condensation, it becomes drain steam and is returned to the conventional island heat recovery system located in the PWR nuclear power unit 11. The deaerator 16 uses fresh steam (clean steam) generated by the steam generator 12 to ensure that the industrial steam ultimately delivered to the user side 14 is completely isolated from the first heat treatment loop and contains no radioactive materials. The water medium in the second heat treatment loop is pressurized from the desalted water tank 18 by the booster pump 19, enters the feed water preheater 17 for heating, and is deoxygenated in the deaerator 16. It is then pressurized by the booster pump 19 and enters the feed water heater 15 for heat exchange and heating, and then enters the steam generator 12 to generate low-pressure superheated steam. If necessary, a superheated heater can be set to generate low-pressure superheated steam, and then the low-pressure superheated steam is compressed into medium-pressure superheated steam by the steam compressor 131 or the electric steam compressor 132, ultimately realizing the demand of the pressurized water reactor nuclear power unit 11 for supplying medium-pressure steam to the user side 14 such as the chemical base.

[0085] The user side 14 shown in Figures 5-10 is an embodiment of a high-pressure user 142. It utilizes the unprocessed steam generated by the pressurized water reactor (PWR) nuclear power unit 11 as a heat source. The unprocessed steam preheats, deoxidizes, and heats the desalted water in the second heat treatment loop. This steam then passes through the steam generator 12 to generate low- or medium-pressure superheated steam. This low- or medium-pressure superheated steam is then converted to high-pressure superheated steam via the steam compressor 131 or the electric steam compressor 132. The main flow of media is as follows: the unprocessed steam enters the superheat heater 21 (if present), the steam generator 12, the feedwater heater 15, and the feedwater preheater 17. After cooling and condensation, it becomes drain water and is returned to the conventional island reheat system within the PWR nuclear power unit 11. The steam used by the deaerator 16 is derived from fresh steam generated by the evaporator 12, ensuring that the industrial steam ultimately delivered to the user side 14 is completely isolated from the first heat treatment loop and contains no radioactive materials. In the second heat treatment loop, the make-up water output from the desalted water tank is pressurized by a booster pump 19, enters the feedwater preheater 17 for heating, and is deoxygenated by the deaerator 16. After being pressurized by the booster pump 19, it enters the feedwater heater 15 for heat exchange and temperature increase, and then enters the steam generator 12 to generate low-pressure or medium-pressure superheated steam. If necessary, a superheater 21 and / or an electric heater 22 can be provided. The low-pressure or medium-pressure superheated steam is then compressed into high-pressure superheated steam by a steam compressor 131 or an electric steam compressor 132, ultimately meeting the high-pressure steam requirements of the pressurized water reactor nuclear power unit 11 for users 14, such as chemical bases. Overall, considering the different steam parameter requirements for medium-pressure and high-pressure steam supply, the preferred steam parameter ranges for the output of the aforementioned preferred embodiments of this application are: 1-18 MPa, 200-600°C.

[0086] FIG11 is a flow chart of a pressurized water reactor nuclear energy steam supply method 201 according to a specific embodiment of the present invention.

[0087] As shown in FIG11 , a pressurized water reactor nuclear energy steam supply method 201 is suitable for a pressurized water reactor nuclear energy steam supply system according to any embodiment of the present invention to deliver clean steam that meets steam parameter requirements to the user side.

[0088] Specifically, the pressurized water reactor nuclear energy steam supply method 201 includes the following steps: Step S1: operating a pressurized water reactor nuclear power unit to output steam to be processed; Step S2: passing the steam to be processed through a first heat treatment circuit, while simultaneously generating clean steam in a second heat treatment circuit independent of the first heat treatment circuit using a steam generator; Step S3: passing the clean steam through a steam compressor to increase the pressure and / or temperature of the clean steam according to steam parameter requirements. The pressurized water reactor nuclear energy steam supply method 201 can be adapted to the pressurized water reactor nuclear energy steam supply system of any of the aforementioned embodiments.

[0089] The above description is merely a preferred embodiment of the present application and is not intended to limit the present application. Various modifications and variations are possible for those skilled in the art. Any modifications, equivalent substitutions, or improvements made within the spirit and principles of the present application shall be included within the scope of protection of the present application.

Claims

1. A pressurized water reactor nuclear energy steam supply system, characterized in that: include: A pressurized water reactor nuclear power unit connected to a first heat treatment loop and outputting steam to be treated into the first heat treatment loop; A steam generator, comprising a tube-side path and a shell-side path isolated from each other, wherein the tube-side path is located in the first heat treatment circuit, and the shell-side path is located in a second heat treatment circuit independent of the first heat treatment circuit, and the steam generator is capable of inputting the steam to be treated in the first heat treatment circuit as a heat source to generate clean steam in the second heat treatment circuit; A steam compressor device is connected to the second heat treatment loop, and the steam compressor device can process the clean steam in the second heat treatment loop to increase the pressure and / or temperature of the clean steam.

2. The pressurized water reactor nuclear energy steam supply system according to claim 1, characterized in that: It also includes a feedwater heater and a feedwater preheater, both of which are connected to the first heat treatment circuit and the second heat treatment circuit to perform heat exchange between the fluid in the first heat treatment circuit and the fluid in the second heat treatment circuit.

3. The pressurized water reactor nuclear energy steam supply system according to claim 2, characterized in that: In the first heat treatment loop, the feed water heater and the feed water preheater are located between the steam generator and the pressurized water reactor nuclear power unit in sequence.

4. The pressurized water reactor nuclear energy steam supply system according to claim 3, characterized in that: It also includes a deaerator and a desalted water tank, which are connected to a second heat treatment loop. In the second heat treatment loop, the desalted water tank is located at the starting end of the second heat treatment loop and is connected to the feed water preheater. The feed water in the desalted water tank passes through the feed water preheater, the deaerator, the feed water heater and the steam generator in sequence.

5. The pressurized water reactor nuclear energy steam supply system according to claim 4, characterized in that: It also includes at least one booster pump located in the second heat treatment loop, wherein the booster pump is located between the desalted water tank and the feed water preheater and / or between the deaerator and the feed water heater.

6. The pressurized water reactor nuclear energy steam supply system according to claim 1, characterized in that: It also includes a superheating heater and / or an electric heater, and the superheating heater and / or the electric heater are arranged before the steam compressor device in the second heat treatment circuit.

7. The pressurized water reactor nuclear energy steam supply system according to claim 1, characterized in that: The steam compressor device includes an electric steam compressor or a steam-driven steam compressor, and includes a single-cylinder structure or a double-cylinder structure.

8. The pressurized water reactor nuclear energy steam supply system according to claim 7, characterized in that: When the pressure ratio of the steam compressor device belongs to the first pressure ratio range, the steam compressor device is of the double-cylinder structure, and when the pressure ratio of the steam compressor device belongs to the second pressure ratio range, the steam compressor device is of the single-cylinder structure, wherein the pressure ratio is the steam ratio between the outlet end and the inlet end of the steam compressor device, and the upper limit value of the first pressure ratio range is greater than the upper limit value of the second pressure ratio range.

9. The pressurized water reactor nuclear energy steam supply system according to claim 7, characterized in that: When the shaft power of the steam compressor device belongs to the first shaft power range, the steam compressor device is the steam-driven steam compressor, and when the shaft power of the steam compressor device belongs to the second shaft power range, the steam compressor device is the electric steam compressor, wherein the shaft power is the power required for the steam compressor device to compress the clean steam, and the upper limit value of the first shaft power range is greater than the upper limit value of the second shaft power range.

10. A method for supplying steam to a pressurized water reactor nuclear energy, using the pressurized water reactor nuclear energy steam supply system as claimed in any one of claims 1 to 9 to deliver clean steam that meets steam parameter requirements to a user side, characterized in that: The method comprises the following steps: Operating pressurized water reactor nuclear power plants to export steam for processing; Passing the steam to be treated through a first heat treatment circuit, while generating clean steam in a second heat treatment circuit independent of the first heat treatment circuit by a steam generator; The clean steam is passed through a steam compressor device to increase the pressure and / or temperature of the clean steam according to the steam parameter requirements.

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