Integral Pressurized Water Reactor

Abstract

The present invention relates to a monolithic nuclear fuel assembly and an integral pressurized water reactor including the same. The reactor includes a steam generator, a cooling water pump, and a pressurizer inside the reactor, and is designed to occupy a minimum amount of space to enable easy relocation and installation, and is a reactor that minimizes neutron irradiation degradation.

Description

Technology Field

[0001] The present invention relates to a monolithic nuclear fuel assembly and an integral pressurized water reactor including the same. Background Technology

[0003] In the case of commercial Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs), multiple nuclear fuels are installed in the core. Control rods are inserted into or withdrawn from the reactor to control it, and boric acid is added to or removed from the coolant to regulate the average nuclear reaction of the reactor. Furthermore, the control rods are not controlled individually, but rather multiple control rods (4 or 12) are connected and bundled together to control the nuclear reactivity by inserting or withdrawing them from the reactor core.

[0004] Integrated reactors, which have a smaller power generation capacity compared to medium-to-large reactors, are designed with a smaller external size to enable easy relocation and installation. They are suitable for supplying power to remote areas without power lines, islands and ships, independent power supply facilities not connected to the power grid, and distributed power systems in data centers and large-scale industrial facilities. Because the core size of an integrated reactor is smaller than that of a medium-to-large nuclear power plant, major components constituting the primary system, such as steam generators, cooling water pumps, and control devices, are integrated and installed inside the reactor.

[0005] FIG. 1 is a conceptual diagram illustrating the main equipment and layout of the primary system of a pressurized water reactor as an example of a conventional large reactor, FIG. 2 is a perspective view and cross-sectional view of a conventional pressurized water reactor, FIG. 3 is a perspective view and cross-sectional view of a conventional nuclear fuel assembly, FIG. 4 is a horizontal cross-sectional view and partial enlarged view of a conventional reactor core, and FIG. 5 is a perspective view of a conventional control rod assembly.

[0006] First, as shown in Fig. 1, the reactor, steam generator, pressurizer, and coolant pump constitute the main equipment of the primary system and are connected by large pipes.

[0007] FIG. 2 shows a vertical cross-section of a pressurized light water reactor as an example of a large reactor. It shows the reactor core (10), the upper guide structure (20) at the top, and the reactor upper structure (50) protecting the control rod drive device (40) above the reactor cover (30).

[0008] Figure 3 shows an example of a nuclear fuel assembly, a cross-sectional view of a nuclear fuel assembly used in a pressurized light water reactor. Looking at the cross-sectional view, it can be seen that a control rod guide tube (70) is installed in place of the nuclear fuel rod (60).

[0009] Figure 4 shows a horizontal cross-section of the reactor core shown in Figure 2. The part (A) marked with a white dot in the partial enlargement view is the location where the control rods are inserted. The control rods are not controlled individually, but rather the reactor is controlled by inserting or withdrawing them from the core in a bundle.

[0010] FIG. 5 shows examples of a 4-element control rod assembly (80) with 4 control rods bundled together and a 12-element control rod assembly (90) with 12 control rods bundled together. As a result of connecting and bundling multiple control rods in this way, the weight of the control rod assembly increases, and the length of the entire control system, including the control rod drive unit, becomes more than three times the length of the control rods actually inserted into the core. Consequently, this results in an increase in the height of the reactor. Furthermore, in the case of a small reactor core, if the fuel assembly of FIG. 3 is used, the control rod guide tubes occupy the fuel rod positions, which has the effect of lowering the maximum core output, and also has the effect of excessively affecting the core nuclear reactivity of the control rod assembly.

[0011] In other words, even if a small reactor is constructed by simply miniaturizing the main equipment and layout of the pressurized water reactor in Fig. 1, it is difficult to lower the height of the entire reactor system. Furthermore, when using a fuel assembly with a reduced length as shown in Fig. 3, the control rods occupy the position of the fuel rods, which lowers the fuel combustion efficiency. Additionally, there is a disadvantage in controlling excessive local nuclear reactivity, and since the steam generator and cooling water pump cannot be integrated with the reactor, the cost of power generation increases during the manufacturing, installation, and operation of the small reactor, resulting in a problem of lower economic feasibility compared to large reactors.

[0012] In addition, conventional large-scale nuclear power plants face the problem that the reactor vessel materials deteriorate due to neutron irradiation as the plant's operating life increases, which limits the ability to extend the plant's lifespan. Prior art literature

[0014] Republic of Korea, Registered Patent Publication, No. 10-2187262 (2020.11.30) The problem to be solved

[0015] The present invention was devised to solve the aforementioned problems and aims to minimize the overall height of the reactor by changing the structure and arrangement method while maintaining the functions of the main equipment of the pressurized water reactor and lowering the height of the control rods.

[0016] In addition, by installing control rod guide tubes between the fuel rods, more fuel rods can be installed, thereby increasing the reactor power and providing the effect of improving the burnup of the fuel.

[0017] In addition, by placing the steam generator at the outer edge of the core and installing the cooling water pump at the top of the core, the main equipment of the primary system, which was previously separated, is integrated inside the reactor into a single structure, thereby providing the effect of enabling effective manufacturing, assembly, installation, and operation of the reactor.

[0018] In addition, it aims to provide the effect of significantly extending the lifespan of the reactor by minimizing degradation of the reactor vessel caused by neutron irradiation. means of solving the problem

[0020] An integrated pressurized light water reactor according to the present invention comprises: a reactor vessel having a flange portion formed on its upper side; a steam generator installed inside the reactor vessel and having a first flange portion formed to be coupled to the upper side of the flange portion; a core shroud coupled to the lower side inside the steam generator; a core support barrel coupled to the upper side of the core shroud; an integrated fuel assembly provided inside the core shroud and having control rod guide tubes installed between a plurality of fuel rods; a control rod cylinder housing coupled to the upper side of the integrated fuel assembly to be received inside the core support barrel and having a control rod cylinder coupled to the upper side of the control rod guide tube; a control circuit support structure having a barrel portion coupled to the upper side of the control rod cylinder housing to be received inside the core support barrel and a second flange portion formed on the upper side of the barrel portion and coupled to the upper side of the first flange portion; a reactor cover coupled to the upper side of the second flange portion and a cooling water coupled to the center of the reactor cover. It is characterized by including a pump.

[0021] Additionally, preferably, the integrated nuclear fuel assembly has a cylindrical or hexagonal prism shape and is characterized by including a plurality of support grids that support the nuclear fuel rods and control rod guide tubes, a nuclear fuel upper support plate into which a control rod guide tube is inserted and installed protruding upward from the uppermost support grid among the plurality of support grids, a control rod guide tube upper nut fixed to the upper end of the control rod guide tube protruding upward from the nuclear fuel upper support plate, a nuclear fuel lower support plate into which a control rod guide tube is inserted and installed protruding downward from the lowermost support grid among the plurality of support grids, and a control rod guide tube lower cap nut fixed to the lower end of the control rod guide tube protruding downward from the nuclear fuel lower support plate.

[0022] In addition, preferably, the upper support plate of the nuclear fuel is characterized by being supported on the upper surface of an upper flange portion formed on the upper side of the core shroud.

[0023] In addition, preferably, the support grid is characterized in that a control rod or an in-core instrument can be inserted into the control rod guide tube.

[0024] Additionally, preferably, the steam generator comprises: an upper header formed by partitions by a bulkhead, wherein a steam discharge nozzle spaced apart along the outer surface of the first flange portion; a cooling water supply nozzle spaced apart along the outer surface of the first flange portion and formed alternately with the steam discharge nozzle; a high temperature header communicating with the steam discharge nozzle; and a low temperature header communicating with the cooling water supply nozzle; a heat transfer tube formed on the lower side of the upper header, wherein a high temperature side heat transfer tube communicating in a straight line with the high temperature header and a low temperature side heat transfer tube communicating in a straight line with the low temperature header are provided; and a lower header formed on the lower side of the heat transfer tube, wherein a connecting header connecting the low temperature side heat transfer tube and the high temperature side heat transfer tube, arranged adjacently so that cooling water discharged from the low temperature side heat transfer tube flows into the high temperature side heat transfer tube, is formed by partitions.

[0025] In addition, preferably, the inner and outer diameters of the upper header and the lower header are spaced apart so that a sufficient distance is formed between the innermost heat transfer tube and the outermost heat transfer tube, thereby shielding neutron rays emitted from the core and minimizing the effect of neutron irradiation on the reactor vessel.

[0026] Additionally, preferably, the core shroud is characterized by having an outer diameter smaller than the inner diameter of the steam generator and having a space formed therein with an inner surface having a shape corresponding to the outer surface of the integrated nuclear fuel assembly.

[0027] Additionally, preferably, the core support barrel is characterized by having an outer diameter equal to the outer diameter of the core shroud, a cooling water discharge hole formed radially on the upper outer circumference, and a third flange portion supported on the upper surface of the first flange portion.

[0028] Additionally, preferably, the control circuit support structure further comprises a cooling water discharge port formed radially on the outer surface of one side of the barrel portion so as to be positioned adjacent to the cooling water discharge port, a cooling water flow guide portion located below the cooling water discharge port and formed to be inclined upward on the inner surface of one side of the barrel portion, a cooling water intake port formed at the end of the cooling water flow guide portion, an elastic plate formed on the inner bottom surface of the second flange portion, and a nozzle formed spaced apart along the outer surface of the second flange portion, so that the flow of cooling water rising from the core is guided to the cooling water intake port along the cooling water flow guide portion.

[0029] In addition, preferably, the nozzle is connected to a plurality of circuits connecting the inside and outside of the reactor, such as a control rod control system, a pressurization system, a shutdown cooling system, a safety injection system, and a reactor coolant system, respectively, and is characterized in that, if necessary, the plurality of circuits can be commonly connected to a single nozzle.

[0030] Additionally, preferably, when the nozzle is connected to the control rod control system, the inner end is coupled to an internal pipe that guides the operating fluid to the control rod cylinder, and the outer end is coupled to an external pipe that guides the inflow of the operating fluid into the internal pipe, and a gas collection device is formed at the uppermost end of the external pipe to collect the operating fluid in a vapor state generated in the internal pipe.

[0031] Additionally, preferably, the cooling water pump comprises a sealed main motor coupled to the upper center of the reactor cover, a drive shaft installed through the center of the reactor cover and rotated by the main motor, a lower housing coupled to the lower part of the drive shaft having a cooling water intake port formed along the axial direction, and a pump housing including an upper housing formed at a certain distance above the lower housing, and an impeller pump housed inside the pump housing and radially discharging the cooling water sucked in through the cooling water intake port so that it flows into the upper interior of the steam generator through the cooling water discharge port. Effects of the invention

[0033] The integrated pressurized water reactor according to the present invention integrates the steam generator, cooling water pump, and reactor control system, which are major components constituting the primary system of the pressurized water reactor, into a single reactor, adopts a hydraulic control system for control rod driving to lower the height of the control system, thereby minimizing the overall height of the reactor, and arranges control rod guide tubes between the fuel rods to increase core output and burnup, thereby minimizing the size of the reactor system and enabling efficient manufacturing, assembly, installation, and operation, and provides the effect of minimizing the degradation of reactor materials by maximizing neutron shielding. Brief explanation of the drawing

[0035] Figure 1 is a conceptual diagram illustrating the main equipment and layout of the primary system of a pressurized water reactor as an example of a conventional large reactor. FIG. 2 is a perspective view and a cross-sectional view of a conventional pressurized water reactor. FIG. 3 is a perspective view and a cross-sectional view of a conventional nuclear fuel assembly. Figure 4 is a horizontal cross-sectional view and a partial enlarged view of a conventional reactor core. FIG. 5 is a perspective view of a conventional control rod assembly. FIG. 6 is a perspective view of an integrated pressurized water reactor according to the present invention. FIG. 7 is an exploded perspective view of an integrated pressurized water reactor according to the present invention. FIG. 8 is a cross-sectional view and a partial enlarged view of an integrated pressurized water reactor according to the present invention. FIG. 9 is a diagram illustrating the flow of primary system cooling water according to the present invention. FIGS. 10 and 11 are drawings illustrating a steam generator according to the present invention and the flow of secondary system cooling water through the steam generator. FIG. 12 is a conceptual diagram illustrating the combination of a core shroud, a core support barrel, and an integrated nuclear fuel assembly according to the present invention. FIG. 13 is a perspective view of an integrated nuclear fuel assembly according to the present invention. Figure 14 is a partial enlarged view and cross-sectional view of (A) of Figure 13. Figure 15 is a partial enlarged view and cross-sectional view of (B) of Figure 13. FIG. 16 is a drawing illustrating the arrangement of nuclear fuel rods and control rod guide tubes inside a support grid according to the present invention. FIG. 17 is a conceptual diagram illustrating the combination of a control rod cylinder housing, a control circuit support structure, and a cooling water pump according to the present invention. FIG. 18 is a cross-sectional view illustrating the combination of a control rod cylinder housing, a control circuit support structure, and a cooling water pump according to the present invention. FIG. 19 is a conceptual diagram illustrating the combination of a reactor cover and a cooling water pump according to the present invention. FIG. 20 is a cross-sectional view illustrating the combination of a reactor cover and a cooling water pump according to the present invention. Specific details for implementing the invention

[0036] Hereinafter, the technical concept of the present invention will be explained in more detail using the attached drawings.

[0037] The attached drawings are merely examples illustrated to further explain the technical concept of the present invention, and therefore the technical concept of the present invention is not limited to the form of the attached drawings.

[0039] The present invention relates to a monolithic nuclear fuel assembly and an integral pressurized water reactor including the same. The present invention will be described in detail below with reference to the attached drawings.

[0041] FIG. 6 is a perspective view of an integrated pressurized water reactor according to the present invention, FIG. 7 is an exploded perspective view of an integrated pressurized water reactor according to the present invention, and FIG. 8 is a cross-sectional view and a partial enlarged view of an integrated pressurized water reactor according to the present invention.

[0042] As illustrated in FIGS. 6 and 7, the integrated pressurized water reactor according to the present invention comprises a reactor vessel (100), a steam generator (200), a core shroud (300), a core support barrel (400), an integrated nuclear fuel assembly (500), a control rod cylinder housing (600), a control circuit support structure (700), a reactor cover (800), and a cooling water pump (900).

[0043] That is, the main equipment constituting the primary system of the pressurized water reactor shown in FIG. 1 is all installed in the reactor vessel (100) of FIG. 6.

[0044] Additionally, when assembling the reactor, a combination of a steam generator (200), a core shroud (300), and a core support barrel (400) is installed in the reactor vessel (100), an integrated nuclear fuel assembly (500), a combination of a control rod cylinder housing (600) and a control circuit support structure (700), a reactor cover (800), and a cooling water pump (900) are installed sequentially.

[0045] Subsequently, as shown in FIG. 8, the reactor assembly is completed by fastening the reactor vessel (100), steam generator (200), control circuit support structure (700), and reactor cover (800) to each other through a stud bolt (B) and a nut (N) that penetrate the flange portions formed in each of them, and details regarding this will be described later.

[0047] Below, the configuration of the integrated pressurized light water reactor according to the present invention will be described in detail.

[0048] First, a flange portion (110) is formed on the upper side of the reactor vessel (100), and bolt fastening holes (H) are formed radially spaced apart in the flange portion (110).

[0049] As shown in FIG. 8, the lower structure of the reactor vessel (100) is formed in an ellipdoidal or torospherical shape, unlike a conventional spherical cross-section, thereby achieving a lower height compared to a hemispherical shape and improving space efficiency.

[0050] Likewise, the reactor cover (800) is also formed as an ellipsoid or a ring instead of a hemisphere.

[0051] However, since the aforementioned non-spherical cross-sectional structure increases stress concentration compared to a spherical cross-sectional structure, it is desirable to form the wall thickness of the relevant area thicker to ensure structural safety.

[0052] Next, the steam generator (200) is installed inside the reactor vessel (100), and a first flange portion (210) is formed that is coupled to the upper part of the flange portion (110), and the flange surfaces where the reactor flange portion (110) and the first flange portion (210) meet are fastened together by a stud bolt (B) and a nut (N).

[0053] To this end, the first flange portion (210) is configured such that a bolt fastening hole (H) is formed at the same position as the bolt fastening hole (H), so that they can be accurately aligned and fastened to each other.

[0054] Below, the detailed configuration of the steam generator (200) will be described in detail, with reference to FIG. 10 and FIG. 11.

[0055] FIGS. 10 and 11 are drawings illustrating the flow of secondary system cooling water through a steam generator (200) and a steam generator (200) according to the present invention.

[0056] As illustrated in FIGS. 10 and 11, the steam generator (200) is configured to include a steam discharge nozzle (220), a cooling water supply nozzle (230), an upper header (240), a heat transfer tube (250), and a lower header (260).

[0057] The above steam generator (200) has a cavity formed inside, which accommodates the lower end of the core support barrel (400) and control circuit support structure (700) to be described later.

[0058] First, a plurality of steam discharge nozzles (220) are spaced apart along the outer surface of the first flange portion (210), and a plurality of cooling water supply nozzles (230) are spaced apart along the outer surface of the first flange portion (210), and are formed alternately with the steam discharge nozzles (220).

[0059] The upper header (240) is divided by a partition wall (245) into a plurality of high-temperature headers (241) communicating with the steam discharge nozzle (220) and a plurality of low-temperature headers (243) communicating with the cooling water supply nozzle (230).

[0060] In detail, the upper header (240) is preferably arranged such that two high-temperature headers (241) and two low-temperature headers (243) are symmetrically positioned diagonally with respect to the cavity by means of a partition wall (245), and accordingly, the steam discharge nozzle (220) and the cooling water supply nozzle (230) are also preferably arranged such that they are symmetrically positioned diagonally with respect to the cavity.

[0061] The 2+2 steam nozzle and feedwater nozzle arrangement exemplified in FIG. 10 can be reduced to 1+1 or increased to 3+3 or 4+4 depending on the steam generated in the steam generator and the amount of feedwater required, FIG. 10 is one example.

[0062] The heat transfer tube (250) is formed on the lower side of the upper header (240) and is coupled to the upper header tube plate (247). It is provided with a high-temperature side heat transfer tube communicating with the high-temperature header (241) and a low-temperature side heat transfer tube communicating with the low-temperature header (243). It is preferably formed in a straight shape to enable periodic inspection and maintenance, reduce cooling water flow resistance, and increase structural rigidity.

[0064] The lower header (260) is formed on the lower side of the heat transfer tube (250), and the heat transfer tube (250) and the lower header (260) are joined by a lower header tube plate (263). A connecting header (261) connecting the low-temperature heat transfer tube and the high-temperature heat transfer tube, which are arranged adjacently so that cooling water discharged from the low-temperature heat transfer tube flows into the high-temperature heat transfer tube, is partitioned by a partition wall (265).

[0065] It is preferable that the lower header (260) has a lower header plate (263) formed to form a flow path so that the cooling water descending from the low-temperature header (243) collects and moves back to the high-temperature header (241).

[0066] The lower header (260) is divided into two zones, and the cooling water discharged from the heat transfer tube (250) on the low-temperature header (243) side and collected therefrom flows into the heat transfer tube (250) on the high-temperature header (241) side.

[0067] In detail, the secondary system cooling water is supplied through the cooling water supply nozzle (230), flows into the low-temperature header (243), descends along the inside of the low-temperature heat transfer tube, receives heat from the primary system cooling water, and collects in the lower header (260).

[0068] Subsequently, it rises along the inner side of the high-temperature heat transfer tube connected to the lower header (260), receives heat from the primary system cooling water, boils, converts into a steam state, collects in the high-temperature header (241), and is configured to be discharged to the outside through the steam discharge nozzle (220). The sludge generated during this process accumulates inside the lower header (260), and is configured to be removed during regular inspection.

[0070] Additionally, preferably, the outer and inner diameters of the upper header (240) and lower header (260) of the steam generator (200) are spaced apart so that a sufficient distance is formed between the innermost heat transfer tube (250) and the outermost heat transfer tube (250), thereby shielding neutrons emitted from the core and minimizing the neutron irradiation effect on the reactor vessel (100).

[0072] Below, with reference to FIG. 12, the configuration of the core shroud (300), core support barrel (400), and integrated nuclear fuel assembly (500) will be described in detail.

[0073] FIG. 12 is a conceptual diagram illustrating the combination of a core shroud (300), a core support barrel (400), and an integrated nuclear fuel assembly (500) according to the present invention.

[0074] As shown in FIG. 12, the core shroud (300) is coupled to the lower side inside the steam generator (200), and a space (370) with a horizontal cross-section that is circular or hexagonal is formed inside so that an integrated nuclear fuel assembly (500) to be described later can be accommodated.

[0075] Specifically, it is preferable that the space portion (370) has an inner surface formed with a shape corresponding to the outer surface of the integrated nuclear fuel assembly (500).

[0076] Additionally, it is preferable that the core shroud (300) be formed to have an outer diameter smaller than the inner diameter of the steam generator (200) by an installation tolerance.

[0077] In addition, it is preferable that a neutron reflector (310) be formed to surround the outer surface of the core shroud (300).

[0079] The core support barrel (400) is coupled to the upper side of the core shroud (300) and is preferably formed to have an outer diameter equal to the outer diameter of the core shroud (300).

[0080] In detail, it is preferable that the core support barrel (400) be joined to an upper flange portion (330) formed on the upper side of the core shroud (300) by means such as welding. Accordingly, when assembling the reactor, the combined assembly of the core shroud (300) and the core support barrel (400) is installed inside the steam generator (200), and then the integrated nuclear fuel assembly (500), which will be described later, is installed inside the core shroud (300).

[0081] Additionally, the core support barrel (400) is composed of a cooling water discharge hole (410) formed radially on the upper outer surface and a third flange portion (430) supported on the upper surface of the first flange portion (210), as shown in FIGS. 8, 9 and 12.

[0082] In detail, the core support barrel (400) has a third flange portion (430) formed on the inner upper side of the first flange portion (210), and an inlet groove (211) is formed on the upper side of the inner surface of the first flange portion (210) to be inserted inward so that the third flange portion (430) is seated thereon, and the core support barrel (400) and the steam generator (200) can be stably connected by the above-described structure.

[0084] Next, the integrated nuclear fuel assembly (500) is provided inside the core shroud (300), and a control rod guide tube (520) is inserted and coupled between a plurality of nuclear fuel rods (510), as shown in FIGS. 13 to 16.

[0085] FIG. 13 is a perspective view of an integrated nuclear fuel assembly (500) according to the present invention, where FIG. 13 (A) illustrates a circular cross-section core and FIG. 13 (B) illustrates a hexagonal cross-section core. FIG. 14 is a partial enlarged view and cross-sectional view of FIG. 13 (A), FIG. 15 is a partial enlarged view and cross-sectional view of FIG. 13 (B), and FIG. 16 is a drawing illustrating the arrangement of a support grid (530), a nuclear fuel rod (510), and a control rod guide tube (520) according to the present invention.

[0086] The reactor according to the present invention uses an integrated nuclear fuel assembly (500) to simplify the manufacturing and replacement of nuclear fuel and provides the effect of preventing buckling by ensuring that tensile stress is applied to the control rod guide tube (520), which is a major structural component. In the case of large nuclear power plants, the nuclear fuel is designed to be subjected to compressive stress, so if the temperature of the core increases in an emergency situation, a phenomenon in which buckling of the nuclear fuel occurs may occur.

[0087] First, as illustrated in FIG. 13, the integrated nuclear fuel assembly (500) includes a support grid (530), a nuclear fuel upper support plate (540), a control rod guide tube upper nut (550), a nuclear fuel lower support plate (560), and a control rod guide tube lower cap nut (570).

[0088] The horizontal cross-section can be designed in a circular or hexagonal shape, a decision made considering factors such as core thermal output, nuclear fuel burnup, and the economic feasibility of the manufacturing process. Generally, from the perspective of neutron economy, neutron leakage in a circular cross-section core is lower than in a hexagonal cross-section core.

[0089] The above support grid (530) is composed of a top support grid (530A, Top-spacer grid), a middle support grid (530B, Mid-spacer grid), and a bottom support grid (530C, Bottom-spacer grid). The middle support grid (530B) is intended to support the fuel rod (510) to maintain a distance from adjacent fuel rods (510), suppress excessive displacement caused by vibration, and maintain the structural integrity of the integrated fuel assembly (500). If the length of the fuel rod (510) is long, it may be added, and if it is short, the number may be reduced or omitted.

[0091] Specifically, the support grid (530) is formed by overlapping a horizontal grid plate (531), a left-facing grid plate (532) that is inclined downward at a certain angle to the left, and a right-facing grid plate (533) that is inclined upward at a certain angle to the right to form a grid shape, as shown on the left side of FIG. 16.

[0092] Specifically, the left-facing grid plate (532) is tilted at an angle of 30 degrees to the left, and the right-facing grid plate (533) is tilted at an angle of 30 degrees to the right, and the length of each plate is preferably adjusted to match the shape and size of the core.

[0093] The horizontal grid plate (531), the left-facing grid plate (532), and the right-facing grid plate (533) are joined together so as to support the nuclear fuel rod (510) and the control rod guide tube (520), and this joining forms a pattern in which a hexagonal first grid (534), a triangular second grid (535) formed outside the long side of the first grid (534), and a triangular third grid (536) formed outside the short side of the first grid (534) are repeatedly arranged.

[0094] Specifically, the first grid (534) is in the shape of an irregular hexagon, and a second grid (535) and a third grid (536) are formed alternately outside the sides of the first grid (534), and it is preferable that the third grid (536) be formed in a smaller size than the second grid (535).

[0095] A nuclear fuel rod (510) is inserted and supported in the inner space of the first grid (534), and a control rod guide tube (520) is inserted and supported in the inner space of the second grid (535).

[0096] That is, the fuel rod (510) and the control rod guide tube (520) are each stably supported by contacting the three sides of the first grid (534) and the second grid (535).

[0097] In addition, if the control rod guide tube (520) is installed only in some locations of the second grid (535), the flow of cooling water in those parts may be obstructed, causing a local flow imbalance, which may result in local overheating.

[0098] Accordingly, it is desirable to configure the control rod guide tubes (520) to be installed at all grid positions of the second grid (535) so that the flow of the cooling water is maintained uniformly throughout.

[0099] In addition, the first grid (534) is formed to be relatively larger than the second grid (535), so that the diameter of the control rod inserted into the control rod guide tube (520) is formed to be relatively smaller than that of the nuclear fuel rod (510), which provides the effect of reducing the nuclear reactivity influence of individual control rods compared to the conventional method when inserting or withdrawing control rods from the core. Therefore, compared to conventional control rods, a number of control rods are installed, and the control rods are distributed more evenly in the core, allowing for precise nuclear reactivity control.

[0101] Additionally, preferably, a control rod or an in-core instrument can be optionally inserted into the control rod guide tube (520) inserted into the inner space of the second grid (535), and the in-core instrument measures the core neutron flux density and coolant temperature, etc. Meanwhile, preferably, the in-core instrument should be manufactured so as to be insertable into the inner diameter of the control rod guide tube.

[0103] That is, the support grid (510) is configured to fix the nuclear fuel rod (510) and the control rod guide tube (520), and the support grid plate, composed of a horizontal grid plate (531), a left-facing grid plate (532), and a right-facing grid plate (533), serves to fix the nuclear fuel rod (510) and the control rod guide tube (520).

[0104] The conventional nuclear fuel assembly exemplified in Fig. 3 has a square cross-section, in which case the control rod guide tube occupies the position of the nuclear fuel rod.

[0105] In contrast, the present invention illustrated in FIG. 16 places a nuclear fuel rod (510) at the center of a hexagon, which is a first grid (534), and supports the nuclear fuel rod by having three sides of the supports intersect at 60° to each other.

[0106] In addition, a control rod guide tube (520) is installed at the center of the triangle, which is the second grid (535), so that the support grid (530) supports and fixes the nuclear fuel rod (510) and the control rod guide tube (520).

[0107] In addition, as illustrated in FIG. 13, the cross-section of the integrated nuclear fuel assembly (500) can be implemented as a circle or a hexagon, and the support grid (530) described above enables the integrated nuclear fuel assembly (500) to maintain such a shape.

[0108] That is, the above hexagonal support grid arrangement has superior rigidity against deformation compared to a square grid, thereby providing the effect of stably maintaining the arrangement of the nuclear fuel rods (510) even under various operating conditions.

[0110] Next, as illustrated in FIG. 14, the upper support plate (540) of the nuclear fuel has a through hole formed therein so that a control rod guide tube (520) protruding above the upper support grid (530A) is inserted and installed, and a cooling water passage (541) is formed adjacent to the through hole so that the support plate does not obstruct the flow of cooling water.

[0111] The upper nut (550) of the control rod guide tube is fixed to the upper end of the control rod guide tube (520) that protrudes above the upper support plate (540) of the nuclear fuel.

[0112] Next, as illustrated in FIG. 15, the nuclear fuel bottom support plate (560) has a through hole formed so that a control rod guide tube (520) protruding downward from the bottom support grid (530C) is inserted and coupled, and a cooling water passage (561) is formed adjacent to the through hole so that the support plate does not obstruct the flow of cooling water.

[0113] The lower cap nut (570) of the control rod guide tube is fixed to the lower end of the control rod guide tube (520) that protrudes downward from the lower nuclear fuel support plate (560). Unlike the upper nut (550) of the control rod guide tube, this is formed as a cap nut with a sealed structure that does not have a through hole formed in the center, thereby preventing interference with the insertion of the control rod due to the inflow of cooling water and preventing a cooling water bypass path.

[0115] Through the structure of the integrated fuel assembly (500) described above, if damage to a fuel rod (510) occurs during reactor operation, access to the individual fuel rod (510) is possible by withdrawing the integrated fuel assembly (500) from the reactor to replace the damaged fuel rod (510), and then separating the upper support plate (540) of the control rod fuel by releasing the upper nut (550) of the control rod guide tube.

[0116] Afterward, the damage to the fuel rods can be inspected, and the damaged fuel rods (510) can be selectively removed. The integrated fuel assembly (500) with a new fuel rod (510) inserted at the removed location can then be reinstalled in the reactor for use.

[0118] In addition, as shown in FIG. 3, the conventional nuclear fuel assembly is a compression load-based design, in which the nuclear fuel is compressed and supported by a lower support system and an upper guide system, and a nuclear fuel pressing spring (65) is present at the top.

[0119] On the other hand, as shown in FIG. 12, the present invention omits the conventional nuclear fuel pressing spring and the nuclear fuel upper support plate (540) is seated and supported on the upper flange portion (330) of the core shroud (300).

[0120] At this time, the nuclear fuel rod (510) and the control rod guide tube (520) are inserted and fixed in the support grid (530), and the control rod guide tube (520) receives the weight of the nuclear fuel rod (510) as a tensile load.

[0121] Meanwhile, preferably, the length of the control rod guide tube (520) is formed to be longer than that of the nuclear fuel rod (510), so that the upper and lower ends of the nuclear fuel rod (510) are spaced apart from the nuclear fuel upper support plate (540) and the nuclear fuel lower support plate (560), thereby preventing damage caused by thermal expansion and vibration.

[0122] Due to structural differences from conventional nuclear fuel assemblies, the present invention resolves structural problems such as buckling deformation that may occur in existing compression load-based designs, thereby improving the structural rigidity of the nuclear fuel assembly (500), simplifying the shape to reduce manufacturing costs, and providing ease of assembly and maintenance.

[0124] Below, the configuration and connection relationship of the control rod cylinder housing (600), the control circuit support structure (700), and the cooling water pump (900) will be described in detail, with reference to FIG. 17 and FIG. 18.

[0125] FIG. 17 is a conceptual diagram illustrating the combination of a control rod cylinder housing (600), a control circuit support structure (700), and a cooling water pump (900) according to the present invention, and FIG. 18 is a cross-sectional view of FIG. 17.

[0126] First, the control rod cylinder housing (600) is coupled to the upper side of the integrated nuclear fuel assembly (500) so as to be received inside the core support barrel (400), and the control rod cylinder (610) coupled to the upper side of the control rod guide tube (520) is received.

[0127] That is, the control rod cylinder housing (600) fixes and supports the control rod cylinder (610), and aligns and connects the control rod cylinder (610) to the upper part of the control rod guide tube (520).

[0128] Additionally, the control rod cylinder housing (600) has an upper flange portion (630) formed on its upper side, and it is preferable that the barrel portion (710) of the control circuit support structure (700), which will be described later, is coupled to the upper part of the upper flange portion (630).

[0130] Next, the control circuit support structure (700) is provided with a barrel portion (710) that is coupled to an upper flange portion (630) on the upper side of the control rod cylinder housing (600) so as to be received inside the core support barrel (400), and a second flange portion (720) formed on the upper side of the barrel portion (710) and coupled to the upper side of the first flange portion (210), as shown in FIG. 8.

[0131] As shown in FIG. 8, it is preferable that the flange surfaces where the first flange portion (210) and the second flange portion (720) meet are fastened to each other by a stud bolt (B) and a nut (N).

[0132] To this end, the second flange portion (720) is configured such that a bolt fastening hole (H) is formed at the same position as the bolt fastening hole (H), so that they can be accurately aligned and fastened to each other.

[0134] Below, the detailed configuration of the control circuit support structure (700) will be described, with reference to FIG. 17 and FIG. 18.

[0135] As illustrated in FIGS. 17 and 18, the control circuit support structure (700) is configured to include a cooling water discharge port (730), a cooling water flow guide (740), a cooling water intake port (750), an elastic plate (760), and a nozzle (770).

[0136] The cooling water discharge port (730) is formed radially on the outer circumference of one side of the barrel portion (710), and is formed adjacent to and aligned with the cooling water discharge hole (410) to minimize fluid resistance, as shown in FIG. 9.

[0137] The cooling water flow guide (740) is located below the cooling water discharge port (730) and is formed to be inclined upward on the inner circumference of one side of the barrel portion (710), and the cooling water intake port (750) is formed at the end of the cooling water flow guide (740) and configured to communicate with the cooling water intake port (953) of the cooling water pump (900) to be described later.

[0138] That is, the flow of cooling water rising from the core is guided by the cooling water flow guide (740) to the cooling water intake port (750) so that it can be introduced into the cooling water pump (900) to be described later.

[0139] Additionally, the cooling water flow guide (740) is provided with a cooling water pump lower support (741) that extends horizontally at the end, thereby providing a space for the lower housing (951) of the cooling water pump (900), which will be described later, to be seated.

[0140] The elastic plate (760) is provided on the inner bottom surface of the second flange portion (720) with a material having a certain elastic force, configured to generate elastic force through bending deformation during reactor assembly, and plays a role in generating assembly load of the main components during reactor assembly.

[0141] Specifically, when the flange portions of the reactor vessel (100), steam generator (200), control circuit support structure (700), and reactor cover (800) are fastened using stud bolts (B) and nuts (N), an assembly load is generated by the elastic force of the elastic plate (760), thereby preventing the parts from detaching or being damaged due to earthquakes, flow, vibration, etc., and maintaining their original functions.

[0142] The assembly load acts in a direction that compresses the barrel portion (710) of the control circuit support structure (700) downward, and this assembly load is continuously transferred to the upper flange portion (630) of the control rod cylinder housing (600) and the control rod cylinder (610) placed inside it. Subsequently, the assembly load is transferred to the upper fuel support plate (540) of the integrated fuel assembly (500) located below the control rod cylinder housing (600), and then the load is transferred to the upper flange portion (330) of the core shroud (300). This load is transferred to the first flange (210) through the core support barrel (400) coupled to the upper side of the upper flange portion (330) of the core shroud (300), and the assembly load is continuously transferred to the upper flange portion (110) of the reactor vessel (100), and the assembly load is sequentially transferred to the lower flange portion (810) and the second flange portion (720) of the reactor cover through the stud bolt (B) and nut (N), thereby completing the load transfer path.

[0143] Meanwhile, since the elasticity of the elastic plate (760) is a key design element for maintaining the reliability of the connection and seal between reactor components, the thickness of the elastic plate (760) must be determined through precise structural analysis and experimentation.

[0144] Generating a fastening force using the elastic plate (760) presented above is one example, and assembly spring force can be generated in other ways. Since the purpose of generating such assembly load is to alleviate the load caused by the difference in thermal expansion between the internal equipment of the reactor, unnecessary stress concentration should not occur due to excessive assembly load.

[0146] Next, the nozzle (770) is spaced out along the outer surface of the second flange portion (720), and in detail, the nozzle (770) is coupled to a plurality of circuits connecting the inside and outside of the reactor, such as a control rod control system (control circuit), a pressurization system, a shutdown cooling system, a safety injection system, and a reactor coolant system. Accordingly, the number of the nozzle (770) must be equal to the number of circuits connecting the inside and outside of the reactor.

[0147] Additionally, if necessary, the above multiple circuits can be bundled and connected commonly to a single nozzle (700). This provides the effect of reducing the number of nozzles (770) by introducing a multi-connector, etc., to allow multiple circuits connecting the inside and outside of the reactor to pass through a single nozzle (770).

[0148] In detail, various pipes may be connected, including control circuit connection pipes for guiding the inflow of the control circuit (i.e., hydraulic fluid and coaxial cable) of the control rod drive unit, as well as heater connection pipes and sprinkler connection pipes, cooling water filling and discharge pipes for controlling the cooling water level inside the reactor, shutdown cooling system pipes for core cooling when the reactor is shut down, and safety injection system pipes for injecting cooling water into the reactor at high pressure in the event of a reactor emergency or a severe accident.

[0149] Below, we will explain in detail the case where a control circuit connecting pipe, a heater connecting pipe, and a sprinkler connecting pipe are connected to the nozzle (770).

[0151] First, when the nozzle (770) is connected to the control rod control system, the nozzle (770) has an inner end connected to an internal pipe (781) that guides the operating fluid to the control rod cylinder (610), and an outer end connected to an external pipe (783) that guides the inflow of the operating fluid to the internal pipe (781).

[0152] Specifically, the control rod drive unit is configured to control individual control rods using a hydraulic cylinder drive method, and accordingly, the control circuit connecting pipe includes a coaxial cable (C) for controlling the control rods to be withdrawn from or inserted into the core, and a pipe through which hydraulic fluid flows.

[0153] At the top of the above external pipe (783), a gas collection device (790) is formed to collect and discharge the steam-state working fluid generated in the above internal pipe (781).

[0154] The above gas collection device (790) includes a gas collector (791) in which an internal space is provided to collect gas from which the working fluid has evaporated, a discharge pipe (793) connected to a cooling water storage tank, and a purge valve (795) formed on one side of the discharge pipe (793).

[0155] Meanwhile, preferably, since the above-mentioned operating fluid uses reactor coolant, the gas collected in the gas collection device (790) is cooled and recovered into the reactor coolant storage tank to prevent leakage of radioactive material.

[0156] The purge valve (795) is provided as an outlet to discharge steam generated inside the control rod cylinder (610) due to the high temperature inside, and when pressure rise is detected due to the formation of steam or when the generation of gas is detected by measuring the internal dryness of the gas collector (791), the purge valve (795) is opened to discharge steam, and is operated to maintain a closed state during normal operation.

[0158] Next, the pressurizer (830) space formed between the control circuit support structure (700) and the reactor cover (800) performs the function of regulating the pressure of the primary system, as shown in FIG. 9.

[0159] More specifically, in FIG. 17, a heater (840) and a sprinkler (850) are installed on the upper side of an elastic plate (760) formed on a control circuit support structure (700), and the heater (840) is connected to a nozzle connected to a heater connecting pipe, and the sprinkler (850) is connected to a nozzle connected to a sprinkler connecting pipe.

[0160] When the amount of heat generated in the core, that is, the integrated nuclear fuel assembly (500), is greater than the amount of heat transferred to the secondary system through the steam generator (200), the temperature of the cooling water rises and expands due to heat accumulation inside the reactor, and as a result, the water level of the primary system cooling water rises. This is manifested as a rise in the water level of the cooling water inside the pressurizer (830).

[0161] At this time, in order to lower the internal pressure of the reactor, the sprinkler (850) is activated to spray cold water onto the top of the pressurizer (830) to lower the temperature and lower the primary system pressure.

[0162] And, conversely, when the internal pressure of the reactor decreases and the water level of the primary system coolant drops, the heater (840) is operated to heat the coolant to raise its temperature and expand the coolant to raise the water level, thereby controlling the pressure by raising the internal pressure of the reactor.

[0164] Below, the configuration and connection relationship of the reactor cover (800) and the cooling water pump (900) will be described in detail, with reference to FIG. 19 and FIG. 20.

[0165] FIG. 19 is a conceptual diagram illustrating the combination of a reactor cover (800) and a cooling water pump (900) according to the present invention, and FIG. 20 is a cross-sectional view illustrating the combination of a reactor cover (800) and a cooling water pump (900) according to the present invention.

[0166] The reactor cover (800) is coupled to the upper part of the second flange portion (720), as shown in FIG. 8.

[0167] As shown in FIG. 8, it is preferable that the flange surfaces of the reactor cover (800) and the second flange portion (720) be joined together by stud bolts (B) and nuts (N).

[0168] To this end, the reactor cover (800) is configured such that a bolt fastening hole (H) is formed at the same location as the bolt fastening hole (H) so that they can be accurately aligned and fastened together.

[0170] Next, the coolant pump (900) is coupled to the upper center of the reactor cover (800) and is configured to include a main motor (910), a drive shaft (930), a pump housing (950), and an impeller pump (970).

[0171] The main motor (910) is a canned motor coupled to the upper center of the reactor cover (800), and in detail, consists of a rotor (911) located on the motor shaft (917) and a stator (electromagnet, 913) formed on the outside of the rotor (911) to induce a magnetic field.

[0172] In particular, the rotor (911) is housed and assembled in a sealing housing (915) that serves as a reactor pressure boundary, and to maintain the reactor pressure boundary, the sealing housing (915) is fastened to the reactor cover (800) to maintain a complete seal.

[0173] Preferably, the seal is performed using an omega welding method to ensure structural reliability and prevent leakage of cooling water or other fluids.

[0174] The present invention simplifies pump maintenance by eliminating the need for a seal on the motor shaft (917) by applying a sealed main motor (910). In particular, since the main motor (910) is installed at a position higher than the pressurizer (830) space of the reactor cover (800), the cooling water does not penetrate into the pressure boundary of the motor, and thus the fluid resistance generated when the motor rotates is reduced, thereby improving operating efficiency.

[0175] The drive shaft (930) is installed through the center of the reactor cover (800) and is rotated by the sealed main motor (910).

[0176] In detail, the drive shaft (930) is housed inside the drive shaft housing (931) and is coupled through the shaft coupling (933) at the lower end of the motor shaft (917), and the lower end of the drive shaft (930) is coupled to the impeller shaft (972) of the impeller pump (970) to be described later. The motor shaft (917), the drive shaft (930), and the impeller shaft (972) must be installed so as to be precisely aligned to prevent vibration of the impeller (971).

[0177] In addition, the impeller pump (970), including the drive shaft (930), is secured by being fastened to the housing bracket (935) on the upper inner side of the reactor cover (800).

[0178] The pump housing (950) is coupled to the lower part of the drive shaft (930) and consists of a lower housing (951) having a cooling water intake port (953) formed along the axial direction, and an upper housing (955) formed at a certain distance above the lower housing (951), thereby protecting the high-speed rotating impeller (971) and serving to efficiently guide the cooling water to the impeller.

[0179] The lower housing (951) is combined with the cooling water flow guide (740) and the cooling water pump lower support (741) of FIG. 18 to prevent the cooling water from bypassing and to allow the cooling water to be efficiently guided to the impeller (971).

[0180] The above impeller pump (970) is housed inside the pump housing (950) and discharges the cooling water sucked in through the cooling water intake port (953) in a radial manner so that it flows into the upper interior of the steam generator (200) through the cooling water discharge port (730), as shown in FIG. 9.

[0181] FIG. 9 is a diagram illustrating the flow of primary system cooling water according to the present invention.

[0182] As shown in FIG. 9, the primary system cooling water absorbs heat generated in the core and rises, and is sucked into the cooling water intake port (953) by the impeller pump (970) and then discharged radially.

[0183] The discharged cooling water passes through the cooling water discharge port (730) provided at the top of the barrel section (710) of the control circuit support structure (700) and the cooling water discharge port (410) of the core support barrel (400), flows into the top of the steam generator (200), and descends along the outside of the heat transfer tube (250), transferring heat to the secondary system cooling water circulating inside the heat transfer tube (250).

[0184] Afterwards, the cooling water that reaches the lower header (260) of the steam generator (200) is recirculated into the core from the bottom of the steam generator (200).

[0186] This concludes the description of the configuration of the integrated pressurized light water reactor according to the present invention.

[0187] The integrated pressurized water reactor according to the present invention relates to the shape of the main equipment constituting the primary system of the reactor, and the dimensions of the equipment must be determined according to changes in the output and size of the reactor, and detailed design can be performed by applying the concept of the present invention.

[0188] In addition, the integrated pressurized water reactor according to the present invention can provide a method for implementing a system in which the primary system can efficiently transfer heat to the secondary system by effectively configuring the core within a limited installation space and integrating the steam generator, cooling water pump, and reactor control system within a single reactor.

[0189] In addition, the integrated pressurized water reactor according to the present invention is configured to form a sufficient distance between the innermost heat transfer tube and the outermost heat transfer tube of the steam generator, thereby enabling shielding of neutron rays generated in the core, which minimizes the deterioration of the reactor vessel and maximizes the lifespan of the reactor.

[0191] Although the present invention has been described above with reference to the embodiments illustrated in the drawings, this is merely for the purpose of explaining the invention, and those skilled in the art will understand that various modifications or equivalent embodiments are possible from the detailed description of the invention.

[0192] Therefore, the true scope of rights of the present invention must be determined by the technical concept of the patent claims. Explanation of the symbols

[0194] 10 : Core 20: Upper guide structure 30: Reactor cover 40: Control rod drive unit 50 : Reactor superstructure 60 : Nuclear fuel rods 65 : Nuclear fuel pressing spring 70: Control rod guide tube 80:4-element control rod assembly 90:12-element control rod assembly 100: Reactor vessel 110 : Flange section 200: Steam generator 210: 1st flange section 211 : Inlet groove 220: Steam exhaust nozzle 230: Coolant supply nozzle 240 : Top header 241 : High temperature header 243 : Cold header 245 : Bulkhead 247 : Upper header tube plate 250 : Heat transfer tube 260 : Subheader 261 : Connection header 263 : Lower header tube plate 265 : Bulkhead 300: Core Shroud 310: Neutron reflector 330 : Upper flange 350 : Lower flange section 370 : Space part 400 : Core support barrel 410: Coolant drain port 430 : 3rd flange section 500: Integrated nuclear fuel assembly 510: Nuclear fuel rod 520: Control rod guide tube 530 : Support grid 530A : Top support grid 530B : Intermediate support grid 530C : Bottom support grid 531 : Horizontal grid 532 : Left-facing grid 533 : Right-sided grid 534 : 1st Lattice 535 : Second lattice 536 : Third Lattice 540: Nuclear fuel top support plate 541: Coolant passage 550: Control rod guide tube top nut 560 : Nuclear fuel bottom support plate 561: Coolant passage 570: Control rod guide tube bottom cap nut 600 : Control rod cylinder housing 610: Control rod cylinder 630 : Upper flange 700 : Control circuit support structure 710 : Barrel part 720 : 2nd flange section 730 : Coolant outlet 740 : Coolant flow guide 741: Coolant pump lower support 750: Coolant intake 760 : Elastic plate 770 : Nozzle 781 : Internal piping 783 : External piping 790 : Gas capture device 791 : Gas collector 793 : Discharge pipe 795 : Purge valve 800: Reactor cover 810: Lower flange 830 : Pressurizer 840 : Heater 850 : Sprinkler 900: Coolant pump 910: Main motor 911: Rotor 913 : Stator 915 : Sealed housing 917: Motor shaft 919 : Sealing bracket 930 : Drive shaft 931 : Drive shaft housing 933 : Axis coupling 935 : Housing bracket 950 : Pump housing 951 : Lower housing 953 : Coolant intake 955 : Upper housing 970 : Impeller pump 971 : Impeller 972: Impeller shaft

Claims

Claim 1 An integrated pressurized light water reactor characterized by comprising: a reactor vessel having a flange portion formed on its upper side; a steam generator installed inside the reactor vessel and having a first flange portion formed to be coupled to the upper side of the flange portion; a core shroud coupled to the lower side inside the steam generator; a core support barrel coupled to the upper side of the core shroud; an integrated fuel assembly provided inside the core shroud and having control rod guide tubes installed between a plurality of fuel rods; a control rod cylinder housing coupled to the upper side of the integrated fuel assembly to be received inside the core support barrel and accommodating a control rod cylinder coupled to the upper side of the control rod guide tube; a control circuit support structure having a barrel portion coupled to the upper side of the control rod cylinder housing to be received inside the core support barrel, and a second flange portion formed on the upper side of the barrel portion and coupled to the upper side of the first flange portion; a reactor cover coupled to the upper side of the second flange portion; and a cooling water pump coupled to the center of the reactor cover. Claim 2 The integrated pressurized light water reactor according to claim 1, wherein the integrated nuclear fuel assembly has a cylindrical or hexagonal prism shape and a unit grid of a certain shape is continuously formed to accommodate a nuclear fuel rod and a control rod guide tube inside, and comprises: a plurality of support grids that support the nuclear fuel rod and the control rod guide tube; a nuclear fuel upper support plate into which a control rod guide tube is inserted and installed, protruding upward from the uppermost support grid among the plurality of support grids; a control rod guide tube upper nut fixed to the upper end of the control rod guide tube protruding upward from the nuclear fuel upper support plate; a nuclear fuel lower support plate into which a control rod guide tube is inserted and installed, protruding downward from the lowermost support grid among the plurality of support grids; and a control rod guide tube lower cap nut fixed to the lower end of the control rod guide tube protruding downward from the nuclear fuel lower support plate. Claim 3 In paragraph 2, the integrated pressurized light water reactor is characterized in that the upper support plate of the nuclear fuel is seated and supported on the upper surface of an upper flange portion formed on the upper side of the core shroud. Claim 4 In paragraph 2, the support grid is characterized by being capable of inserting a control rod or an in-core instrument into the control rod guide tube, in an integrated pressurized light water reactor. Claim 5 The steam generator according to claim 1 comprises: a steam discharge nozzle spaced apart along the outer surface of the first flange portion; a cooling water supply nozzle spaced apart along the outer surface of the first flange portion and formed alternately with the steam discharge nozzle; an upper header partitioned by a bulkhead, wherein a high-temperature header communicating with the steam discharge nozzle and a low-temperature header communicating with the cooling water supply nozzle are partitioned; a heat transfer tube formed on the lower side of the upper header, wherein a high-temperature side heat transfer tube communicating in a straight line with the high-temperature header and a low-temperature side heat transfer tube communicating in a straight line with the low-temperature header are provided; and a lower header formed on the lower side of the heat transfer tube, wherein a connecting header connecting the low-temperature side heat transfer tube and the high-temperature side heat transfer tube, arranged adjacently so that cooling water discharged from the low-temperature side heat transfer tube flows into the high-temperature side heat transfer tube, is partitioned by a bulkhead. Claim 6 An integrated pressurized light water reactor according to claim 5, wherein the inner and outer diameters of the upper header and lower header are spaced apart to form a sufficient distance between the innermost heat transfer tube and the outermost heat transfer tube, thereby shielding neutron rays emitted from the core and minimizing the neutron irradiation effect on the reactor vessel. Claim 7 An integrated pressurized light water reactor according to claim 1, wherein the core shroud has an outer diameter smaller than the inner diameter of the steam generator and is provided with a space portion having an inner surface formed on the inside that has a shape corresponding to the outer surface of the integrated nuclear fuel assembly. Claim 8 An integrated pressurized light water reactor according to claim 1, wherein the core support barrel has an outer diameter equal to the outer diameter of the core shroud and includes: a cooling water discharge hole formed radially on the upper outer surface; and a third flange portion supported on the upper surface of the first flange portion. Claim 9 In claim 8, the control circuit support structure further comprises: a cooling water discharge port formed radially on the outer surface of one side of the barrel portion to be positioned adjacent to the cooling water discharge port; a cooling water flow guide portion located below the cooling water discharge port and formed to be inclined upward on the inner surface of one side of the barrel portion; a cooling water intake port formed at the end of the cooling water flow guide portion; an elastic plate formed on the inner bottom surface of the second flange portion; and a nozzle formed spaced apart along the outer surface of the second flange portion; wherein the flow of cooling water rising from the core is guided to the cooling water intake port along the cooling water flow guide portion. Claim 10 An integrated pressurized light water reactor according to claim 9, wherein the nozzle is each connected to a plurality of circuits connecting the inside and outside of the reactor, such as a control rod control system, a pressurization system, a shutdown cooling system, a safety injection system, and a reactor coolant system, and wherein, if necessary, the plurality of circuits can be commonly connected to a single nozzle. Claim 11 An integrated pressurized light water reactor characterized in that, in the case where the nozzle is connected to the control rod control system, the inner end is coupled to an internal pipe that guides the operating fluid to the control rod cylinder, the outer end is coupled to an external pipe that guides the inflow of the operating fluid into the internal pipe, and at the uppermost end of the external pipe, a gas collection device is formed to collect the operating fluid in a steam state generated in the internal pipe. Claim 12 In claim 9, the cooling water pump comprises: a sealed main motor coupled to the upper center of the reactor cover; a drive shaft installed through the center of the reactor cover and rotated by the main motor; a pump housing coupled to the lower part of the drive shaft and including a lower housing having a cooling water intake port formed along the axial direction and an upper housing formed at a certain distance above the lower housing; and an impeller pump housed inside the pump housing and radially discharging the cooling water sucked in through the cooling water intake port so that it flows into the upper interior of the steam generator through the cooling water discharge port.

Images