JP2023058274A - Fuel assembly and core of nuclear reactor - Google Patents

Abstract

To provide a fuel assembly capable of suppressing an occurrence of dryout of a liquid film.SOLUTION: A fuel assembly 13 is formed by arranging fuel rods 1-4, 6 and 8 whose lower ends are supported by a lower tie plate and whose upper ends are supported by an upper tie plate, and fuel rods 5 and 7 (partial-length fuel rods) whose lower ends are supported by the lower tie plate and whose upper ends are not supported by the upper tie plate, in a channel box 17. The effective fuel length of the fuel rods 5 is 2/3 of that of the fuel rods 6, and the effective fuel length of the fuel rods 7 is 9/10 of that of the fuel rods 6. The fuel rods 5 are respectively arranged at four corners in a third row from the inner face of the channel box 17 in a fuel rod array. The fuel rods 7 are arranged at the third from the fuel rods 1 arranged at the respective corners in the second row in each of two orthogonal directions along the inner surface of the channel box 17, with the fuel rods 1 arranged at the respective corners in the second row from the inner surface of the channel box 17 used as base points in the fuel rod array.SELECTED DRAWING: Figure 1

Description

The present invention relates to fuel assemblies and nuclear reactor cores, and more particularly to fuel assemblies and nuclear reactor cores suitable for application to boiling water reactors.

In order to effectively utilize uranium resources, reduce the amount of radioactive waste generated, and effectively utilize plutonium, the development of derate-spectrum boiling water reactors is underway based on light water reactor technology. A low moderation spectrum boiling water reactor has multiple fuel rods filled with nuclear fuel material arranged in a rectangular channel box with a square cross section in the core. ing. This fuel assembly has a dense arrangement of these fuel rods, greatly reducing the proportion of moderator water in the channel box. For this reason, in the fuel assembly loaded in the core of a low-speed spectrum boiling water reactor, the fast neutrons generated by the fission of the fissile material contained in each fuel material in the fuel rod are not moderated so much, and relatively High-energy neutrons are used to fission fissile materials.

JP 2020-118526 describes fuel assemblies that are loaded into the core of a reduced-spectrum boiling water reactor. In this fuel assembly, a plurality of fuel rods filled with MOX fuel are arranged in an equilateral triangular lattice, the pitch between the fuel rods is small, and the amount of water present inside is small. When a plurality of fuel rods are arranged in an equilateral triangular lattice in a channel box with a square cross section, a gap with a triangular cross section is formed near the inner surface of the channel box and rises inside the channel box. Cooling water can easily rise through the gap. As a result, the flow rate of cooling water rising through the gap whose cross section is triangular increases. In the fuel assembly disclosed in Japanese Patent Application Laid-Open No. 2020-118526, a water removal rod attached to the inner surface of the channel box is arranged in each gap having a triangular cross section. This arrangement of the water exclusion rods further reduces the amount of water in the channel box. In such fuel assemblies loaded in the core of a low moderation spectrum boiling water reactor, the neutron spectrum is hardened to improve the conversion of uranium-238 to plutonium-239, making the void reactivity coefficient even more negative. be able to.

Japanese Patent Application Laid-Open No. 2019-178896 describes a fuel assembly in which a plurality of fuel rods filled with fissile plutonium are arranged in a channel box in 10 rows and 10 columns. This fuel assembly contains fissile plutonium and has two types of part-length fuel rods with different active fuel lengths. The effective fuel length of these part length rods is longer than the other.

JP 2020-118526 A JP 2019-178896 A

Japanese Patent Application Laid-Open No. 2020-118526 describes a fuel assembly to be loaded into the core of a low-speed spectrum boiling water reactor. Like this fuel assembly, generally, a fuel assembly loaded into the core of a low-speed spectrum boiling water reactor has a plurality of fuel rods arranged in an equilateral triangular lattice in a channel box having a square cross section. , a triangular gap is inevitably formed between the fuel rods arranged in the outermost region of the fuel rod array and the inner surface of the channel box. Due to the formation of this triangular gap, the cooling water rising inside the fuel assembly may be unevenly distributed (concentrated) in the gap and rise. may decrease and the heat removal performance of the fuel rods located in that central region may be degraded.

In order to solve such a problem, in JP-A-2020-118526, water removal rods are arranged in respective triangular gaps near the inner surface of the channel box. The arrangement of these water displacement rods requires that a large number of water displacement rods be provided, with the lower end of each water displacement rod supported by a lower tie plate and their upper end supported by an upper tie plate. . Furthermore, it is necessary to bundle the water exclusion rods and the fuel rods while maintaining a predetermined distance between the water exclusion rods and the fuel rods using a plurality of fuel spacers. As a result, the installation of water scavenging rods complicates the construction of the fuel assembly.

In consideration of the above circumstances, the inventors, as described in detail later, changed the fuel rod arrangement in the rectangular tubular channel box of the fuel assembly from an equilateral triangular lattice to Japanese Patent Application Laid-Open No. 2019-178896. A fuel assembly without water scavenging rods was considered, with the square lattice shown and the fuel rods closely spaced to harden the neutron spectrum during reactor operation. In this study, the inventors found that in the cross section of the fuel assembly, the fuel rods arranged in the second and third layers from the inner surface of the channel box in the fuel rod array were liquid film dried due to the increase in output. I found out that out may occur.

Therefore, realization of a fuel assembly capable of suppressing the occurrence of liquid film dryout is desired.

A first object of the present invention is to provide a fuel assembly and a core of a nuclear reactor that can suppress the occurrence of liquid film dryout.

A second object of the present invention is to provide a fuel assembly and a core of a nuclear reactor capable of suppressing a decline in the inventory of nuclear fuel material.

A first feature of the present invention for achieving the first object noted above is a lower fuel support member;
an upper fuel support member;
a channel box, which is a rectangular tube having a square cross section, attached at its upper end to the upper fuel support member and extending toward the lower fuel support member;
A fuel assembly comprising a plurality of fuel rods filled with nuclear fuel material and arranged in a square lattice in a channel box,
The fuel rods include a plurality of first fuel rods having lower ends supported by the lower fuel support members and upper ends supported by the upper fuel support members; a plurality of second fuel rods that are not supported by the fuel support members and have an active fuel length that is less than that of the first fuel rods;
The second fuel rods are arranged in the second and third rows from the inner surface of the channel box in the fuel rod arrangement in the cross section of the fuel assembly.

By arranging the second fuel rods in the second and third rows from the inner surface of the channel box in the fuel rod arrangement in the cross section of the fuel assembly, the occurrence of liquid film dryout in the fuel assembly is prevented. can be suppressed. By suppressing the occurrence of liquid film dryout in the fuel assembly, the thermal margin of the fuel assembly can be improved, and the thermal output during operation of the nuclear power plant can be increased.

A second feature of the present invention for achieving the above second object is that the plurality of second fuel rods are arranged in the third row and the second row. and a plurality of fourth fuel rods, wherein the active fuel length of the fourth fuel rods is longer than the active fuel length of the third fuel rods.

Since the effective fuel length of the fourth fuel rod is longer than the effective fuel length of the third fuel rod, it is possible to suppress the decrease in the amount of nuclear fuel material loaded in the fuel assembly.

Preferably, the fourth fuel rods are arranged in each of two orthogonal directions along the inner surface of the channel box, starting from the first fuel rods arranged at each of the four second corners of the second row, In addition to being arranged in the third from the first fuel rods arranged at the respective second corners, the above-mentioned In each of the two orthogonal directions, it is desirable that the first and second fuel rods are also arranged from the first fuel rod arranged at the respective second corner.

The fourth fuel rod is located third from the first fuel rod located at the respective second corner in each of said two orthogonal directions, and in addition, said orthogonal fuel rod in said second row. In each of the two directions, the first and second fuel rods are arranged from the first fuel rod arranged at the respective second corner, so that the degree of reduction in the nuclear fuel material loading in the fuel assembly is small.

According to the first feature of the present invention, it is possible to suppress the occurrence of liquid film dryout in the fuel assembly.

According to the second feature of the present invention, it is possible to suppress reduction in the amount of nuclear fuel material loaded in the fuel assembly.

FIG. 3 is a cross-sectional view (sectional view taken along line II in FIG. 3) of a fuel assembly of Example 1 applied to a low-speed spectrum boiling water reactor, which is a preferred example of the present invention; FIG. 2 is an explanatory diagram showing the enrichment of fissile plutonium in each fuel rod included in the fuel assembly shown in FIG. 1; FIG. 2 is a longitudinal sectional view (III-III sectional view in FIG. 1) of the fuel assembly shown in FIG. 1; 2 is a longitudinal cross-sectional view of a low-speed spectrum boiling water nuclear power plant loaded with the fuel assemblies shown in FIG. 1; FIG. FIG. 2 is an explanatory diagram for explaining the distribution of the heat generation coefficient of each fuel rod in the cross section of the fuel assembly shown in FIG. 1 in which all the partial length fuel rods are replaced with full length fuel rods; FIG. 4 is a cross-sectional view of a fuel assembly of Example 2 applied to a low-speed spectrum boiling water nuclear reactor, which is another preferred example of the present invention; FIG. 7 is an explanatory diagram showing the enrichment of fissile plutonium in each fuel rod included in the fuel assembly shown in FIG. 6; FIG. 7 is an explanatory diagram for explaining the distribution of the heat generation coefficient of each fuel rod in the cross section of the fuel assembly shown in FIG. FIG. 8 is an explanatory diagram showing the relationship between the flow rate of cooling water (core flow rate) and MCPR in each of the fuel assembly (with part-length fuel rods) and the fuel assembly not adopting the part-length fuel rods of Example 2; FIG. 3 is a cross-sectional view of a fuel assembly of Example 3 applied to a low-rate spectrum boiling water nuclear reactor, which is another preferred example of the present invention; Figure 11 is a cross-sectional view of the fuel rod shown in Figure 10; FIG. 3 is a cross-sectional view of a fuel assembly of a conventional example;

The inventors have studied suppression of liquid film dryout in a fuel assembly in which fuel rods are arranged in a dense square grid in a rectangular tubular channel box.

Therefore, in this study, the inventors assumed a conventional fuel assembly in which a plurality of fuel rods are densely arranged in a square lattice in a channel box, which is a rectangular tube with a square cross section. FIG. 12 shows this assumed conventional fuel assembly 13C. The fuel assemblies 13C are loaded into the core of the reactor of a reduced-rate spectrum boiling water nuclear power plant. The fuel assembly 13C has a plurality of fuel rods 11C arranged in 13 rows and 13 columns in a channel box 17, which is a rectangular tube having a square cross section. These fuel rods 11C are arranged in a square lattice.

As shown in FIG. 12, the four support rods 12 are arranged at each of the four corners of the outermost peripheral region (the first row from the inner surface of the channel box 17) of the fuel rod arrangement in the cross section of the fuel assembly 13C. Arranged by book. A water gap region 39 where saturated water exists is formed between the fuel assemblies 13C loaded in the core. The water gap region 39 exists on each of the four side surfaces of the channel box 17 of the single fuel assembly 13C, that is, the outer surface. A control rod 34 having a cruciform cross section is inserted into a water gap region 39 formed between the fuel assemblies 13C.

In this fuel assembly 13C, a plurality of fuel rods 11C highly enriched in fissile plutonium are arranged in the central portion of the cross section, and a plurality of other fuel rods 11C containing burnable poisons such as gadolinia are arranged in the outermost region. placed in In the central part of the cross section, since the fuel rods 11C containing highly enriched fissile plutonium are densely arranged, the amount of cooling water (light water) that acts as a moderator is reduced, and the thermal neutron flux is small (i.e. , the neutron spectrum becomes stiff).

On the other hand, in the outer periphery of the fuel assembly 13C, which is close to the water gap region existing between the fuel assemblies 13C loaded in the core, the neutron spectrum is softened under the influence of the cooling water in the water gap region 39. . For this reason, the nuclear fission reaction of fissile plutonium becomes active in the outer peripheral portion of the cross section of the fuel assembly 13C, and the respective fuel rod arrays arranged in the second and third rows from the inner surface of the channel box 17 of the fuel rod array. The power of fuel rod 11C is increased. Such conditions continue throughout the operating cycle of the nuclear plant. When the output of specific fuel rods 11C arranged on the outer periphery of the fuel assembly 13C increases, a phenomenon (liquid film dryout) occurs in which the outer surface of the fuel rods 11C dries up, leading to burnout of the fuel rods 11C. . For this reason, it becomes a bottleneck of the thermal margin in the operation of the nuclear reactor, and limits the rated operating output of the fuel assembly 13C.

As a result of studying measures to solve such a problem, the inventors have found that part-length fuel rods are arranged in the second row and the third row from the inner surface of the channel box in the fuel rod arrangement in the cross section of the fuel assembly. , it is possible to suppress the occurrence of liquid film dryout in the fuel rods in the fuel assembly.

Further, it is desirable to have a higher loading of nuclear fuel material in fuel assemblies having part length rods.

An embodiment of the present invention in consideration of the above study results will be described below. Embodiments of the present invention described below can be applied to a reduced-rate spectrum boiling water reactor, which is a boiling water reactor. This derate spectrum boiling water reactor uses cooling water as a coolant, and a recirculation pump causes the cooling water to flow out of the reactor pressure vessel and back into the reactor pressure vessel. A boiling water reactor (BWR) that circulates cooling water, an advanced boiling water reactor (ABWR) that has an internal pump and circulates cooling water inside a reactor pressure vessel, and an internal pump in the ABWR including reduced-rate spectrum boiling water reactor applications such as high economy simplified boiling water reactors (ESBWR) that do not use

A fuel assembly of Example 1 used in a low speed spectrum boiling water nuclear power plant, which is a preferred example of the present invention, will be described with reference to FIGS. 1, 2, 3 and 4. FIG.

first. The structure of the nuclear reactor of the low-rate spectrum boiling water nuclear power plant 20 in which the fuel assemblies of this embodiment are loaded into the core will be described with reference to FIG. The reactor 21 of this reduced speed spectrum boiling water nuclear power plant 20 is an ABWR.

This nuclear reactor 21 has a reactor pressure vessel 22, and a core 23 loaded with a plurality of fuel assemblies 13 (see FIGS. 1, 2 and 3) is arranged in the reactor pressure vessel 22. . Inside the reactor pressure vessel 22 , a cylindrical core shroud 24 surrounds a core 23 , and a shroud head 25 positioned above the core 23 is installed at the upper end of the core shroud 24 . A plurality of steam separators 28 are attached to the shroud head 25 and extend upwardly. Additionally, a steam dryer 29 is installed within the reactor pressure vessel 22 above the steam separator 28 . An annular downcomer 32 is formed between the outer surface of the core shroud 24 and the inner surface of the reactor pressure vessel 22 . An internal pump 26 located within the downcomer 32 extends downwardly through the bottom of the reactor pressure vessel 22 and is attached to the bottom of the reactor pressure vessel 22 . An internal pump 26 has an impeller 27 . A main steam line 37 and a feedwater line 38 are connected to the reactor pressure vessel 22 .

An upper grid plate 30 is positioned above the core 23 and attached to the inner surface of the core shroud 24 . A core support plate 31 is positioned below the core 23 and attached to the inner surface of the core shroud 24 . A plurality of fuel support fittings 33 are installed on the core support plate 31 .

A lower plenum 36 is formed within the reactor pressure vessel 22 below the core 23 . A plurality of control rod guide tubes (not shown) are disposed in the lower plenum 36 . Each control rod 34 having a plurality of neutron-absorbing rods filled with a neutron-absorbing material (e.g., boron carbide) and having a cross-sectional shape of a cross is separately arranged in a respective control-rod guide tube. . A plurality of control rod drive mechanisms 35 are installed at the bottom of the reactor pressure vessel 22 and extend downward from the bottom of the reactor pressure vessel 22 , each control rod drive mechanism 35 being separately connected to the control rods 34 . be.

As shown in FIG. 3, the fuel assemblies 13 loaded in the core 23 are composed of a plurality of fuel rods 11, a lower tie plate (lower fuel support member) 14, an upper tie plate (upper fuel support member) 15, and axially It has a plurality of fuel spacers 18 and channel boxes 17 arranged therein. The fuel assembly 13 is a MOX fuel assembly containing mixed oxide fuel (MOX fuel) of uranium oxide and plutonium oxide.

Each fuel rod 11 has a cladding tube (not shown), the lower end of which is closed with a lower end plug (not shown) and the upper end of the cladding tube is closed with an upper end plug (not shown). The cladding tube is filled with a plurality of fuel pellets (not shown) containing nuclear fuel material (MOX fuel). A gas plenum (not shown) is formed within the cladding tube above the nuclear fuel material loading region filled with those fuel pellets. A lower end of each fuel rod 11 is supported by a lower tie plate 14 and an upper end of each fuel rod 11 is supported by an upper tie plate 15 . A handle 16 is provided on the upper tie plate 15 .

Each fuel rod 11 is bundled by a plurality of fuel spacers 18 arranged in the axial direction. The bundled fuel rods 11 are placed in a downwardly extending channel box 17 having its upper end attached to an upper tie plate 15 . The channel box 17 is a square cylinder with a square cross section. 165 fuel rods 11 are arranged in a square lattice in the channel box 17 with 13 rows and 13 columns. Cooling water passages 19 are formed between the fuel rods 11 . The diameter of the fuel rod 11, that is, the outer diameter of the cladding tube is 8.0 mm.

The fuel assembly 13 has four support rods 12 in addition to the fuel rods 11 . As shown in FIG. 1, the four support rods 12 are arranged at each of the four corners of the outermost peripheral region (the first row from the inner surface of the channel box 17) of the fuel rod arrangement in the cross section of the fuel assembly 13. Arranged by book. Each support rod 12 is made of metal and is preferably made of a metal with a low neutron absorption cross section. Support rods 12 do not contain nuclear fuel material. The lower ends of the support rods 12 are threaded into and attached to the lower tie plate 14 . The upper ends of the support rods 12 are inserted into holes formed in the upper tie plate 15 and held by the upper tie plate 15 . Each support rod 12 serves to hold each fuel spacer 18 axially in place.

The lower tie plate 14 of the fuel assemblies 13 loaded in the core 23 is supported by fuel support fittings 33 installed on the core support plate 31 . Each fuel support fitting 33 supports the lower tie plates 14 of four fuel assemblies 13 . The upper ends of the channel boxes 17 of each of the four fuel assemblies 13 in which the lower tie plate 14 is supported by one fuel support fitting 33 are inserted into one square formed in the upper lattice plate 30. . The upper grid plate 30 is formed with a plurality of grids having a square cross section and passing through the upper grid plate 30 . The upper end portions of the four fuel assemblies 13 supported by the respective fuel support metal fittings are inserted into the respective squares, and the upper end portions of the respective fuel assemblies 13 are supported by the upper lattice plate 30 .

A water gap region 39 in which saturated water exists is formed between the fuel assemblies 13 loaded in the core 23, similarly to between the fuel assemblies 13C shown in FIG. This water gap region 39 exists facing each of the four side surfaces of the channel box 17 of the single fuel assembly 13 . A control rod 34 connected to a control rod drive mechanism 35 is inserted through a control rod insertion hole (not shown) which is formed in the center of the fuel support fitting 33 and penetrates the fuel support fitting 33 and has a cross-shaped cross section. The fuel assemblies 13 are inserted between the four fuel assemblies 13 supported by the fuel support fittings 33 through the holes. The insertion of the control rods 34 between the fuel assemblies 13 and the withdrawal of the control rods 34 from between the fuel assemblies 13 are performed by a control rod drive mechanism 35 . The control rods 34 inserted between the fuel assemblies 13 are present within the water gap region 39 .

The arrangement of the plurality of fuel rods 11 within the channel box 17 in the cross section of the fuel assembly 13 will be specifically described with reference to FIG. The plurality of fuel rods 11 within the fuel assembly 13 includes fuel rods 1, 2, 3, 4, 5, 6, 7 and 8. The respective number of fuel rods 1-8 in fuel assembly 13 is shown in FIG. 12 fuel rods 1, 16 fuel rods 2, 8 fuel rods 3, 8 fuel rods 4, 4 fuel rods 5, 89 fuel rods 6, 8 fuel rods 7 and fuel The number of bars 8 is twenty.

Of the plurality of fuel rods 11, the fuel rods 1 to 4, 6 and 8, which are full-length fuel rods, have their lower ends supported by the lower tie plate 14 and their upper ends supported by the upper tie plate 15. there is The axial length of the nuclear fuel material filling region in each of the fuel rods 1-4, 6 and 8, that is, the active fuel length is the same. The fuel rods 5 and 7 are supported by the lower tie plate 14 at their lower ends, but are not supported by the upper tie plate 15 at their upper ends. The active fuel length of each of fuel rods 5 and 7 is less than the active fuel length of each of fuel rods 1-4, 6 and 8, and each of fuel rods 5 and 7 is a part length fuel rod. The effective fuel length of each of the fuel rods 5 and 7 is longer than 1/2 of the effective fuel length of each of the fuel rods 1-4, 6 and 8, and the effective fuel length of the fuel rod 7 is longer than that of the fuel rod 5. Longer than the fuel effective length. The effective fuel length of the fuel rod 5 is 2/3 of the effective fuel length of the fuel rods 1-4, 6 and 8, and the effective fuel length of the fuel rod 7 is the effective fuel length of the fuel rods 1-4, 6 and 8. It is 9/10 of the fuel effective length.

When L1 is the effective fuel length of the long part-length fuel rod (for example, the fuel rod 7) and L2 is the effective fuel length of the short part-length fuel rod (for example, the fuel rod 5), the following equation (1) is obtained. Each of the effective fuel lengths L1 and L2 can be determined so as to satisfy

{(L−L1)×n1+(L−L2)×n2}/(L×n)≦0.072 (1)
However, L is the active fuel length of full-length fuel rods corresponding to each of fuel rods 1 to 4, 6 and 8, n1 is the number of long part-length fuel rods (for example, fuel rod 7), and n2 is The number of short part-length rods (eg, fuel rods 5 ) and n is the number of total fuel rods contained in the fuel assembly 13 .

Equation (1) calculates the sum of effective fuel lengths (L×n) of all full-length fuels in a fuel assembly in which there are no partial-length fuel rods. Long partial length so that the total reduction length of fuel effective length caused by arranging It represents setting the effective fuel length L1 of the fuel rod and the effective fuel length L2 of the short partial length fuel rod.

If the formula (1) is satisfied, the effective fuel length L1 of the long partial-length fuel rods may be other than 9L/10, and the effective fuel length L2 of the short partial-length fuel rods may be other than 2L/3. It may be the fuel effective length.

Specifics for determining each of the effective fuel length L2 of the short part-length rods (e.g., fuel rods 5) and the effective fuel length L1 of the long part-length rods (e.g., fuel rods 7) according to equation (1): method. By calculating the core performance for the fuel assembly, the distribution of the heating coefficient in the cross section of the fuel assembly (see, for example, FIG. 5) is obtained, and the short partial length fuel rods and the long length fuel rods Determine where to place each of the part length rods. Further, the number of short part-length fuel rods (n2) and the number of long part-length fuel rods are determined based on the determined positions for arranging the short part-length fuel rods and the long part-length fuel rods. Determine (n1). After that, the effective fuel length L of the full-length fuel rods, the number n of full-length fuel rods when all the fuel rods in the fuel assembly are full-length fuel rods, and the number of long partial-length fuel rods are added to the equation (1). By substituting n1 and the number n2 of short partial length fuel rods, and further substituting an arbitrary value for either one of the effective fuel length L1 and the effective fuel length L2 (for example, the effective fuel length L2), an arbitrary The other effective fuel length (for example, effective fuel length L1) for which no value is substituted is calculated. By carrying out several patterns of calculation of such formula (1), the lengths of the optimum effective fuel length L1 and the optimum fuel effective length L2 are comprehensively confirmed, and MCPR determines the reference value (design value). In the filled condition, the active fuel lengths L1 and L2 of the long and short part length rods, respectively, are determined to be maximized.

Each of the fuel rods 1 to 7 is filled with MOX fuel, which is a nuclear fuel material, in a sealed cladding tube, and the nuclear fuel material does not contain burnable poison. The fissile Pu enrichment of the MOX fuel in each of the fuel rods 1-7 is described below with reference to FIG. The fissile Pu enrichment in the fuel rod 1 is 3.6 wt%, the fissile Pu enrichment in the fuel rod 2 is 5.0 wt%, and the fissile Pu enrichment in the fuel rod 3 is 6.2 wt%, the fissile Pu enrichment in fuel rod 4 is 8.8 wt%, the fissile Pu enrichment in fuel rod 5 is 11.5 wt%, fuel rods 6 and 7 The fissionable Pu enrichment in each is 12.1 wt%. The fuel rods 8 are filled with nuclear fuel material such as depleted uranium, and contain burnable poison such as gadolinium (Gd). The concentration of gadolinium in the fuel rods 8 is 9.0 wt%. The fuel rods 8 do not contain fissile Pu.

The fuel rods 6 and 7 have the highest enrichment of fissile Pu in the fuel assembly 13 . Fuel rod 5 has the next highest enrichment of fissile Pu after fuel rods 6 and 7, respectively. The fissile Pu enrichment of each of the fuel rods 4, 3, 2 and 1 is smaller than the fissile Pu enrichment of the fuel rod 5, and the order of the fuel rods 4, 3, 2 and 1 is is smaller than

The above fissile Pu enrichment values and burnable poison concentration values are values in the fuel assembly 13 with a burnup of 0 GWd/t, which includes each of the fuel rods 1-8. The fuel assembly 13 with a burnup of 0 GWd/t is a new fuel assembly, and is a fuel assembly that has not experienced the operation of a nuclear reactor.

The arrangement positions of the fuel rods 1 to 8 in the cross section of the fuel assembly 13 will be explained.

The fuel rods 1 are arranged in the outermost peripheral region of the fuel rod arrangement in the channel box 17 and in the second row from the inner surface of the channel box 17 . At the corners facing the control rods 34 in the transverse cross section of the fuel assembly 13, perpendicular lines along the inner surface of the channel box 17 with the support rods 12 arranged at each of the four corners of the outermost peripheral region as base points In each of the two directions, fuel rod 1 is positioned second from support rod 12 in its outermost region. Furthermore, at the corners facing the control rods 34, the fuel rods 1 are also arranged at the corners of the second row. The arrangement of the outermost peripheral region and the second row of fuel rods 1 at the corner portion facing the control rods 34 is the same as the other three positions in the cross section of the fuel assembly 13 other than the corner portion facing the control rods 34 . It is also done in two corners.

At the corners facing the control rods 34, the fuel rods 8 are arranged between the support rods 12 and the fuel rods 1 in each of two orthogonal directions along the inner surface of the channel box 17 within the outermost peripheral region. be. Also, within the outermost peripheral region, other fuel rods 8 are arranged adjacent to the fuel rods 1 adjacent to the fuel rods 8 adjacent to the support rods 12 in the two orthogonal directions thereof. The arrangement of the fuel rods 8 in the outermost peripheral region at the corners facing the control rods 34 is different from the corners facing the control rods 34 at the other three corners in the cross section of the fuel assembly 13. is also being done. Other fuel rods 8 adjoin fuel rods 1 on opposite sides of fuel rods 8 adjoining support rods 12 and fuel rods 1, respectively, in the outermost region. , the fuel rods 8 are also arranged in the central part of each square piece formed by the four support rods 12 in their outermost peripheral region. In the fuel assembly 13, the fuel rods 8 are arranged only in the outermost peripheral region.

The fuel rods 3 are arranged in the second row from the inner surface of the channel box 17 in the fuel rod array in the cross section of the fuel assembly 13 . At the corner portion facing the control rod 34, the fuel rods 3 are positioned at the corners of the second row in each of the two orthogonal directions along the inner surface of the fuel rods 1 arranged at the corners of the second row. are arranged adjacent to the fuel rods 1 arranged in the . Besides the corners facing the control rods 34, the fuel rods 3 are arranged in the same manner as the corners facing the control rods 34 at the other three corners.

The fuel rods 4 are arranged in the second row from the inner surface of the channel box 17 in the fuel rod array in the cross section of the fuel assembly 13 . At the corner facing the control rod 34, in each of the two orthogonal directions with the fuel rod 1 arranged in the corner of the second row as the base point, the fuel rod 4 Located adjacent to bar 3 . At the three corners other than the corners facing the control rods 34, the fuel rods 4 are arranged in the same way as at the corners facing the control rods 34. FIG.

The fuel rods 5 are arranged in the third row from the inner surface of the channel box 17 in the fuel rod array in the cross section of the fuel assembly 13 . At the corners facing the control rods 34, the fuel rods 5 are arranged in the corners of the third row. Specifically, the fuel rods 5 are arranged in the same manner as the corners facing the control rods 34 at the three corners other than the corners facing the control rods 34 .

The fuel rods 7 are arranged in the second row from the inner surface of the channel box 17 in the fuel rod array in the cross section of the fuel assembly 13 . In the second row, the fuel rods 7 are arranged adjacent to the fuel rods 4 on the opposite side of the fuel rods 3 .

The plurality of fuel rods 6 are not arranged in the outermost peripheral region, and together with the plurality of fuel rods 7, the cross section of the fuel assembly 13, which does not include the above-described four corner portions in the cross section of the fuel assembly 13, It is arranged in the central region on the central axis side of the fuel assembly 13 from the second row of the fuel rod arrangement.

The fuel rods 1 to 8 arranged as described above are arranged point-symmetrically about the central axis of the fuel assembly 13 in the cross section of the fuel assembly 13 . At the four corners in the cross section of the fuel assembly 13, the enrichment of fissile Pu from the fuel rods 5 arranged at the corners of the third row from the inner surface of the channel box 17 in the fuel rod array toward the outermost peripheral region is decreasing. At each of the four corners are fuel rods 11 arranged at the corners of the third row, and fuel rods arranged at the corners of the second row on the second row from the inner surface of the channel box 17. 1, fuel rods 3 arranged adjacent to both sides of this fuel rod 1, and fuel rods 8 arranged adjacent to both sides of support rods 12 arranged at the corners in the outermost peripheral region, these fuel rods 1 arranged respectively adjacent fuel rods 8; fuel rods 8 respectively arranged on the opposite side of each fuel rod 1 from the support rods 12; There are fuel rods 2 that have been At each corner, the fissile Pu enrichment of each of the fuel rods 11 present in the third row, the fuel rods 11 present in the second row, and the fuel rods 11 present in the outermost peripheral region decreases in the order of the third row, the second row, and the outermost peripheral region. The fissile Pu enrichment of the plurality of fuel rods 11 arranged at these four corners is that of the fuel rods 11 existing in the third row, that is, the fuel rods 6 adjacent to the fuel rods 5. is smaller than In other words, the enrichment of fissile Pu of the fuel rods 11 arranged in each corner is equal to the enrichment of fissile Pu of each of the plurality of fuel rods 6 and the plurality of fuel rods 7 arranged in the central region. less than degrees.

In the low moderation spectrum boiling water nuclear power plant 20, all the control rods 34 are inserted into the core 23 at the end of one operation cycle, and the operation in that operation cycle ends. While the operation of the nuclear power plant 20 is stopped, fuel replacement and maintenance inspection of the nuclear power plant 20 are performed. In the fuel exchange, some of the fuel assemblies 13 loaded in the core 23 are removed as spent fuel assemblies from the reactor 21 to a fuel storage pool (not shown) and stored in the fuel storage pool. New fuel assemblies (fuel assemblies with a burnup of 0 GWd/t) are loaded into the core 23 in the reactor 21 . At the time of refueling, for example, when 1/4 of all the fuel assemblies 13 loaded in the core 23 are removed from the core 23 as spent fuel assemblies when the operation of the nuclear power plant 20 is stopped, the removed fuel assemblies are removed. The same number of new fuel assemblies 13 as the number of spent fuel assemblies that have been collected are loaded into the core 23 . As a result, before the operation of the nuclear power plant 20 with the new fuel assemblies 13 loaded in the core 23 starts in the next operation cycle, the core 23 does not experience the operation of the nuclear power plant 20. The fuel assembly 13 with a degree of 0 GWd/t, the fuel assembly 13 that has experienced operation in one operation cycle, the fuel assembly 13 that has experienced operation in two operation cycles, and the fuel assembly 13 that has experienced operation in three operation cycles. A fuel assembly 13 is present.

Since the fuel assembly 13 with a burnup of 0 GWd/t has fuel rods 8, it contains burnable poison. The concentration of the burnable poison contained in the fuel rods 8 present in the fuel assembly 13 with a burnup of 0 GWd/t is such that the burnable poison disappears at the end of one operation cycle due to the operation of the nuclear power plant 20. is set to Therefore, the remaining fuel assemblies 13 that have experienced operation in one or more operation cycles present in the core 23 will be: Contains no burnable poisons.

After the fuel replacement and the maintenance and inspection of the nuclear plant 20 are finished, the operation of the nuclear plant 20 is started in the next operation cycle. When the internal pump 26 is driven, the cooling water present in the downcomer 32 inside the reactor pressure vessel 22 is pressurized by the impeller 27 of the internal pump 26 . The pressurized cooling water is supplied from the fuel support fitting 33 through the lower plenum 36 into the channel box 17 of the fuel assembly 13, and rises through the cooling water passages 19 formed between the fuel rods 11 in the channel box 17. do. While ascending through the cooling water passages 19 , the cooling water is heated by heat generated by fission of fissile material (eg fissile Pu) contained in the nuclear fuel material present in the fuel rods 11 . Since part of the heated cooling water becomes steam, the cooling water becomes a gas-liquid two-phase flow containing water and steam.

This gas-liquid two-phase flow is discharged from the upper ends of the fuel assemblies 13 , that is, from the core 23 and flows into the steam separator 28 . Within the steam separator 28, the gas-liquid two-phase flow is separated into water and steam. The separated water is discharged from the steam separator 28 to the downcomer 32, descends through the downcomer 32 as cooling water, and is pressurized by the internal pump 26. - 特許庁The separated steam is led from the steam separator 28 to the steam dryer 29, and the moisture contained in the steam is removed by the steam dryer 29. The steam that has been dehumidified and discharged from the steam dryer 29 is discharged from the reactor pressure vessel 22 to the main steam line 37 and is led through the main steam line 37 to a steam turbine (not shown). The steam rotates a steam turbine, which in turn rotates a generator (not shown) coupled to the steam turbine. Power is generated by the rotation of the generator. Steam discharged from the steam turbine is condensed into water in a condenser (not shown). This condensed water is supplied to the reactor pressure vessel 22 as feed water through the feed water pipe 38 .

The width in the horizontal direction of the cooling water passages 19 formed between the fuel rods 11 is the inner surface of the fuel rods 11 arranged in the outermost peripheral region of the fuel rod array and the channel box 17 in the cross section of the fuel assembly 13. and the width of the water gap region 39 existing between the fuel assemblies 13 . Therefore, in the central region of the cross section of the fuel assembly 13, the amount of cooling water (light water) acting as a moderator per fuel rod 11 (fuel rods 6 and 7) is The amount of moderator cooling water per fuel rod located in the outermost region facing the gap between the fuel rods 11 located in the region and the inner surface of the channel box 17 and the water gap region 39. very little compared to Therefore, the neutron spectrum becomes harder toward the center of the aforementioned central region in the cross section of the fuel assembly 13 . On the other hand, the gap between the fuel rods 11 arranged in the outermost region of the fuel rod array and the inner surface of the channel box 17 and the outermost region near the water gap region 39 soften the neutron spectrum.

In the fuel assembly 13 of this embodiment, the neutron spectrum becomes softer toward the outer periphery as described above. On the other hand, the fuel rods 8 arranged in the outermost peripheral region of the fuel assembly 13 contain burnable poison. Therefore, the output of the fuel rods 11 in the outermost peripheral region of the fuel assembly 13 is suppressed, and the output of the fuel rods 11 arranged in the second row from the inner surface of the channel box 17 in the fuel rod array is highest in In addition, as the nuclear power plant 20 operates in a certain operation cycle and the burning of fissile Pu contained in each fuel material in the fuel rods 11 progresses, the Since the thermal output of the fuel rods 11 decreases, the thermal output of the fuel rods arranged in the third row from the inner surface of the channel box 17 becomes the highest in the fuel assembly 13 .

When designing a nuclear reactor, it is necessary to evaluate the MCPR (Minimum Critical Power Ratio). MCPR uses the thermal margin of the BWR as an evaluation index. The critical heat output is the fuel assembly heat output that causes the transition from nucleate boiling to film boiling. MCPR is the minimum critical power ratio (CPR) among the fuel assemblies loaded in the core, among the critical power ratios (CPR) defined by critical thermal power/fuel assembly generated thermal power.

The critical thermal output is a value that depends on the design of the fuel assembly, and becomes higher as the fuel assembly is more thermally hydraulically superior. It is necessary to determine the thermal power of the nuclear reactor during operation so that the critical power ratio is kept below the regulation value. That is, if fuel rods with a high thermal load are present in the fuel assembly when the fuel assembly is designed, the critical thermal output of the fuel assembly is reduced. As a result, in order to keep the critical power ratio below the regulation value, the thermal power of the nuclear reactor must be lowered during operation. This brings about economic problems due to reduced power generation of the reactor and safety problems such as ensuring sufficient operating margins.

The liquid film dryout that occurs within the fuel assembly is described below. To explain this liquid film dryout, the inventors replaced all of the partial length fuel rods 5 and 7 with full length fuel rods in the fuel assembly 13 of this embodiment shown in FIG. A fuel assembly having a configuration was assumed. This assumed fuel assembly is hereinafter referred to as the first assumed fuel assembly.

In the above-described first assumed fuel assembly, as in the fuel assembly 13 of the present embodiment, the neutron spectrum becomes harder toward the central portion in the cross section of the first assumed fuel assembly. In addition, the neutron spectrum becomes soft in the outermost region of the fuel rod array. Therefore, the closer the fuel rod is to the inner surface of the channel box 17, the higher the heat output. Since the fuel rods containing burnable poison are arranged in the outermost peripheral region of the fuel rod array, the output of the fuel rods in the outermost peripheral region of the first assumed fuel assembly is suppressed, and the inner surface of the channel box 17 of the fuel rod array The output of the fuel rods arranged in the second row from 1 is the highest in the first assumed fuel assembly. In addition, as the nuclear power plant 20 operates in a certain operation cycle and the burning of fissile Pu contained in each fuel material in the fuel rods 11 progresses, the Since the thermal output of the fuel rods is reduced, the thermal output of the fuel rods arranged in the third row from the inner surface of the channel box 17 is the highest in the first assumed fuel assembly.

In a boiling water nuclear reactor, the thermal margin of the fuel rods is maintained by removing heat from the fuel rods through the presence of sufficient cooling water on the surfaces of the fuel rods. During operation of the reactor, the cooling water flows into the channel box 17 from the lower tie plate of the first hypothetical fuel assembly loaded in the core and out of the upper tie plate, as described above. The cooling water boils in the first assumed fuel assembly due to the heat generated by the nuclear fission of the fissile Pu in the full-length fuel rods, and a part thereof becomes steam. Therefore, the gas-liquid two-phase flow rises in the first assumed fuel assemblies loaded in the core.

The flow pattern of the gas-liquid two-phase flow depends on the respective velocities and pressures of cooling water and steam, but is most affected by the volume ratio of cooling water and steam. On the upstream side of the first assumed fuel assembly, the volume ratio of steam to cooling water is small, and cooling water flows in most of the first assumed fuel assembly. Therefore, on the upstream side of the first assumed fuel assembly, a flow pattern of bubble flow is formed such that steam bubbles flow in the cooling water. On the downstream side of the first assumed fuel assembly, the heat generated in the fuel rods causes the cooling water rising between the fuel rods to sufficiently boil, and the volume ratio of steam to the cooling water becomes sufficiently large. Therefore, downstream of the first assumed fuel assembly, a gas-liquid two-phase flow flows between the fuel rods, and a flow pattern of an annular spray flow is formed such that a liquid film flows on the outer surface of the fuel rods.

Under conditions of high heat output of the fuel rods, it is conceivable that all the liquid film flowing on the outer surfaces of the fuel rods will evaporate. Such a phenomenon that all the liquid film on the outer surface of the fuel rod evaporates and the outer surface of the fuel rod dries is called liquid film dryout. When a liquid film is present on the outer surfaces of the fuel rods, the heat generated in the fuel rods can be substantially removed by boiling and vaporization of the liquid film. On the other hand, when liquid film dryout occurs, the only heat in the fuel rods is the convective removal of vapor containing surrounding droplets. Heat removal by convection has lower heat removal performance than heat removal by boiling and evaporation, and the heat generated in the fuel rod cannot be removed. This causes the temperature of the fuel rods to rise sharply, and this sudden rise leads to burning out of the fuel rods.

In the first assumed fuel assembly, the fuel rods arranged in the second and third rows from the inner surface of the channel box 17 in the fuel rod arrangement have the highest thermal output in the first assumed fuel assembly, These fuel rods experience liquid film dryout and determine the critical thermal power of the first hypothetical fuel assembly. That is, if the occurrence of liquid film dryout in the fuel rods 11 arranged in the second and third rows can be suppressed, the critical thermal power of the first assumed fuel assembly increases, and the economy of the reactor increases. It can improve security and safety.

As described above, the liquid film dryout occurs under the condition of an annular spray flow in which the volume ratio of steam to the cooling water is sufficiently large. It is assumed to occur near the upper end of the active fuel length of the full-length fuel rods in the first assumed fuel assembly, where the phase flow flows. Therefore, in the first assumed fuel assembly, liquid film dryout occurs near the upper end of the effective fuel length in the fuel rods arranged in the second and third rows from the inner surface of the channel box 17 in the fuel rod arrangement. expected to occur.

The inventors obtained the heat generation coefficient of each fuel rod in the cross section of the above-mentioned first assumed fuel assembly by core performance calculation. The determined heat generation coefficient of each fuel rod in the cross section of such a fuel assembly is shown in FIG. In the first assumed fuel assembly, the fissile Pu enrichment of the full-length fuel rods corresponding to the fuel rods 5 is the same as that of the fuel rods 5 in the fuel assembly 13, and the full-length fuel rods corresponding to the fuel rods 7 The fissile Pu enrichment is the same as that of the fuel rods 7 in the fuel assembly 13 . The first hypothetical fuel assembly subject to the core performance calculation does not contain part-length fuel rods. FIG. 5 shows the heat generation coefficient of each fuel rod in the cross section of the first hypothetical fuel assembly obtained by the core performance calculation. The heating coefficient of a fuel rod represents the relative heating value of individual fuel rods in the first assumed fuel assembly when the average heating value of the fuel rods in the first assumed fuel assembly is 1.

In FIG. 5, "1, 2, 3, . "1, 2, 3, . The position of each fuel rod is represented by (column number, row number). In FIG. 5, each position of (1,1), (1,13), (13,1) and (13,13) in the cross section of the first assumed fuel assembly corresponds to the fuel rod arrangement in that cross section. 4 shows the positions of the four corners in the outermost peripheral region of . Since the support rods 12 are arranged at these corner positions, the heat generation coefficient at those positions is "0.00". The corner portion of the cross section of the first assumed fuel assembly, where the support rod 12 arranged at the position (1, 1) exists, faces the control rod 34 inserted into the water gap region 39 .

In the cross section of the first assumed fuel assembly, the fuel rods arranged in the cross section are arranged point-symmetrically around the central axis of the fuel assembly (existing at the position of (7, 7)). there is

According to the heat generation coefficients of the fuel rods shown in FIG. becomes the largest at The heat generation coefficient of the fuel rods 11 arranged at positions (3,11), (11,3) and (11,11) is "1.42", and the heat generation coefficient of the fuel rods 11 arranged at position (3,3) is It is almost the same as the heat generation coefficient of the fuel rod 11. Each fuel rod 11 located at positions (3,3), (3,11), (11,3) and (11,11) is a channel of the fuel rod array in the cross section of the fuel assembly. It is arranged at each corner in the third row from the inner surface of the box 17 .

According to FIG. 5, the fuel rods 11 having the next largest heat generation coefficient after the fuel rods 11 arranged at positions (3,3), (3,11), (11,3) and (11,11) are , (2,5), (2,9), (5,2), (5,12), (9,2), (9,12), (12,5) and (12,9) The fuel rods 11 arranged in position. The heat generation coefficient of the fuel rods 11 arranged at positions (2, 5) and (5, 2) is "1.40". The heating coefficients of the fuel rods 11 located at positions (2,9), (5,12), (9,2), (9,12), (12,5) and (12,9) are 1.39, which is approximately the same as the heat generation coefficient of the fuel rods 11 located at positions (2,5) and (5,2). Positions (2,5), (2,9), (5,2), (5,12), (9,2), (9,12), (12,5) and (12,9) The fuel rods 11 arranged in the inner surface of the channel box 17 are based on the fuel rods 11 arranged in each corner of the second row from the inner surface of the channel box 17 in the fuel rod arrangement in the cross section of the fuel assembly. in each of the two orthogonal directions along the .

Considering the distribution of such exothermic coefficients, the above (3,3), (3,11), (11,3), (11,11), (2,5), (2,9), Fuel rods 11 (full length fuel rods) arranged at positions (5,2), (5,12), (9,2), (9,12), (12,5) and (12,9) , liquid film dryout may occur.

We have identified the fuel rod positions (3,3), (3,11), (11,3), (11,11), (2,5 ), (2,9), (5,2), (5,12), (9,2), (9,12), (12,5) and (12,9), respectively, liquid film dry In order to suppress the occurrence of out, I came up with the idea of arranging part-length fuel rods. As noted above, at each of the locations (3,3), (3,11), . , by arranging the part length fuel rods, it is possible to avoid arranging nuclear fuel material near the upper end of the active fuel length of the fuel assembly 13 where liquid film dryout is expected to occur at those positions; The occurrence of liquid film dryout can be suppressed. The effective fuel length of the fuel assembly 13 is the same as the effective fuel length of the full-length fuel rods in the fuel assembly 13 .

Specifically, as shown in FIG. 1, in the fuel rod arrangement of the cross section of the fuel assembly 13, each of (3,3), (3,11), (11,3) and (11,11) A fuel rod 5 is arranged at the position. Also, (2,5), (2,9), (5,2), (5,12), (9,2), (9,12), (12,5) and (12,9) A fuel rod 7 longer than the fuel rod 5 is arranged at each position. The effective fuel length of the fuel rods 7 can be longer than that of the fuel rods 5 because the heat coefficient at the locations where the fuel rods 7 are located is smaller than the heat coefficient at the locations where the fuel rods 5 are located.

In other words, the fuel rods 11 (long fuel rods) arranged at positions (3,3), (3,11), (11,3) and (11,11) shown in FIG. At the outer surface of the cladding tube, liquid film dryness can occur at the upper end of its active fuel length. Liquid film dryout occurs at points on the outer surface of the cladding tube. A liquid film dryout region where such liquid film dryout may occur extends downward from the upper end of the effective fuel length in the axial direction of the fuel rod 11 . It is considered that the length of the liquid film dryout occurrence region in the axial direction downward from the upper end of the effective fuel length corresponds to the heat generation coefficient of the fuel rods 11 arranged at the above positions. The larger the exothermic coefficient, the longer the length of the liquid film dryout occurrence region. In order to suppress the occurrence of liquid film dryout, the fuel rods 11 arranged at positions (3,3), (3,11), (11,3) and (11,11) are The fuel rod 5 (partial length fuel rod) having an effective fuel length shorter than the effective fuel length by the length of the liquid film dryout occurrence region may be used. Also, (2,5), (2,9), (5,2), (5,12), (9,2), (9,12), (12,5) and (12,9) The effective fuel length of the fuel rods 7 arranged at each position, which is longer than the fuel rods 5, has heat generation coefficients of (3, 3) and (3, 11) at the above positions where the fuel rods 7 are arranged. , (11, 3) and (11, 11) are longer than the effective fuel length of the fuel rod 5.

That is, the heat generation coefficients at locations (3,3), (3,11), (11,3) and (11,11), and (2,5), (2,9), (5,2 ), (5,12), (9,2), (9,12), (12,5) and (12,9), and (3,3), (3 , 11), (11,3) and (11,11), the active fuel lengths L2 of the part-length fuel rods arranged at the positions of (11,11) and 2,5), (2,9), (5,2), To determine the effective fuel length L1 of the part length fuel rods arranged at positions (5,12), (9,2), (9,12), (12,5) and (12,9), As described above, several patterns of calculation of formula (1) are performed, and the effective fuel lengths L1 and L2 are determined to be maximized while MCPR satisfies the reference value (design value).

Also, (2,5), (2,9), (5,2), (5,12), (9,2), (9,12), (12,5) and (12,9) The length of the liquid film dryout occurrence region in the axial direction on the outer surface of the cladding tube of the fuel rods 11 (full-length fuel rods) arranged at each position depends on the heat generation coefficient of the fuel rods 11 arranged at those positions. It is considered to correspond. Since the heat coefficient of the fuel rods 11 placed at positions (2, 5) etc. is smaller than that of the fuel rods 11 placed at positions (3, 3) etc., The length of the liquid film dryout occurrence region in the axial direction of the arranged fuel rods 11 is shorter than that of the fuel rods 11 arranged at positions such as (3, 3). In order to suppress the occurrence of liquid film dryout, the fuel rods 11 arranged at the positions (2, 5), etc., have a length corresponding to the length of the liquid film dryout occurrence region rather than the effective fuel length of the fuel rods 6. The fuel rods 7 (part-length fuel rods) having an active fuel length that is short but longer than the active fuel length of the fuel rods 5 may be used.

According to this embodiment, since the part-length fuel rods are arranged in the second and third rows from the inner surface of the channel box in the fuel rod arrangement, liquid film dryout occurs in the fuel assembly 13. can be suppressed. By suppressing the occurrence of liquid film dryout in the fuel assembly 13, the thermal margin of the fuel assembly can be improved, and the thermal output during operation of the nuclear power plant 20 can be increased.

In particular, the fuel rods 5, which are part-length fuel rods, are arranged at each corner of the third row from the inner surface of the channel box 17 in the fuel rod arrangement in the cross section of the fuel assembly 13, and the fuel in the cross section of the fuel assembly 13 is arranged. In each of the two orthogonal directions along the inner surface of the channel box 17, starting from the fuel rods 1 arranged at the corners of the second row from the inner surface of the channel box 17 of the rod array, each corner of the second row A fuel rod 7, which is a part-length fuel rod, is arranged as the third fuel rod from the fuel rods 1 arranged in . By arranging the fuel rods 5 and 7 in this manner, the occurrence of liquid film dryout in the fuel assembly 13 can be further suppressed.

The arrangement of the fuel rods 5 having a shorter active fuel length than the fuel rods 7 from the inner surface of the channel box 17 to each corner of the third row of the fuel rod arrangement in the cross section of the fuel assembly 13 is This can contribute to further suppression of the occurrence of liquid film dryout.

Two orthogonal directions along the inner surface of the channel box 17 with the fuel rods 11 arranged at each corner of the second row from the inner surface of the channel box 17 as the base point of the fuel rod arrangement in the cross section of the fuel assembly 13 In each case, the effective fuel length from the fuel rod 11 arranged at each corner of the second row to the fuel rod 7 arranged in the third row of the fuel rod array of the second row is the channel box 17 of the fuel rod array. Since it is longer than the effective fuel length of the fuel rods 5 arranged at the corners of the third row from the inner surface, it is possible to suppress reduction in the loading amount of nuclear fuel material in the fuel assembly. Suppressing the reduction of the nuclear fuel material loading, that is, increasing the Pu loading, ultimately reduces the volume of the spent nuclear fuel.

Fuel rods 5 arranged at each corner ((3, 3), etc.) of the third row from the inner surface of the channel box 17 of the fuel rod array in the cross section of the fuel assembly 13, and the fuel rod array of the fuel rod array, the channel box Fuel rods arranged at each corner of the second row in each of two orthogonal directions along the inner surface of the channel box 17, with the fuel rods 1 arranged at each corner of the second row from the inner surface of the channel box 17 as the base point. When the liquid film dryout does not occur in the fuel rods 7 arranged in the first to third positions ((5, 2), etc.), the liquid film dryout occurs at each corner of the second row from the inner surface of the channel box 17. In each of the two orthogonal directions along the inner surface of the channel box 17 with the fuel rods 1 arranged in the second row as the base point, It is presumed to occur in fuel rods 3 and 4 located respectively in the first ((3,2), etc.) and second ((4,2), etc.) fuel rods 1 to 1, respectively.

That is, the fuel rods 5 arranged at each corner of the third row from the inner surface of the channel box 17, and the fuel rods 1 arranged at each corner of the second row from the inner surface of the channel box 17 of the fuel rod arrangement are the base points. In each of the two orthogonal directions along the inner surface of the channel box 17, the effective fuel length of each of the fuel rods 1 arranged at each corner of the second row to the fuel rod 7 arranged in the third row is Even if the length is shorter than the length that can avoid the occurrence of liquid film dryout, from the fuel rods 1 arranged at each corner of the second row to the fuel rods 3 and 4 arranged in the first and second rows, respectively Since liquid film dryout occurs, it does not contribute to improvement of thermal margin. Furthermore, it is feared that the shorter the effective fuel length of the part-length fuel rods, the smaller the volume of nuclear fuel material in the fuel assembly and the smaller the fuel loading in the fuel assembly. For this reason, it is desirable to ensure that the active fuel length of the part length fuel rods is as long as possible if liquid film dryout can be avoided in the relevant fuel rods. In this embodiment, the third corner ((3, 3), etc.) of the fuel rod array from the inner surface of the channel box 17 in the cross section of the fuel assembly 13, where the heat generation coefficient is the largest and the thermal load is large. Part-length fuel rods with a short effective fuel length were used as the fuel rods placed in the . Also, in each of the two orthogonal directions along the inner surface of the channel box 17, with the fuel rods arranged at the corners of the second row from the inner surface of the channel box 17 in the above fuel rod arrangement, two rows A fuel rod located third from the fuel rods located at each corner of the row, i.e., a fuel rod having a lower heating coefficient and a smaller thermal load than the fuel rods located at the corners of the third row. Part-length fuel rods having a longer effective fuel length than the part-length fuel rods arranged at the corners of the third row were used as the fuel rods. As a result, the effective fuel length of the part length rods can be maximized while avoiding liquid film dryout, minimizing the reduction in fuel loading of the fuel assembly due to the use of part length rods. can be done.

In this embodiment, the support rods containing no nuclear fuel material located at the four corners of the outermost region of the fuel rod array reduce the power output at the corners of the fuel assembly 13, thus reducing the traversal of the fuel assembly 13. It contributes to flattening the output distribution on the surface. That is, the cross-sectional area of the portion formed between the corners of the four fuel assemblies 13 facing each other in the water gap region 39 is equal to the cross-sectional area of the adjacent fuel in the water gap region 39 The water gap region 39 is wider than the cross-sectional area of the portion formed between the sidewalls of the channel box 17 facing each other in the assembly 13, and the water gap region 39 between the corners of the four fuel assemblies 13. More thermal neutrons are generated in the part formed in the For this reason, the power output at the corners of the cross section of the fuel assembly 13 tends to increase. is reduced, and the power distribution in the cross section of the fuel assembly 13 is flattened.

At the four corners in the cross section of the fuel assembly 13, the enrichment of fissile Pu from the fuel rods 5 arranged at the corners of the third row from the inner surface of the channel box 17 in the fuel rod array toward the outermost peripheral region is decreasing. At each corner in the cross section of the fuel assembly 13, the fissile Pu enrichment from the fuel rods 5 arranged at the corners of the third row from the inner surface of the channel box 17 to the outermost peripheral region of the fuel rod array. A reduction in the degree of flattening results in a further flattening of the power distribution across the cross section of the fuel assembly 13 .

Placing the burnable poison in the outermost region of the fuel rods 8 can further increase the value of the burnable poison in the fuel rods 8 .

A fuel assembly of Example 2 used in a low speed spectrum boiling water nuclear power plant, which is another preferred example of the present invention, will be described with reference to FIGS. 3, 4, 6 and 7. FIG.

The nuclear reactor 21 of the slow spectrum boiling water nuclear power plant 20 in which the fuel assemblies of this embodiment are loaded into the core is an ABWR.

The arrangement of the plurality of fuel rods 11 within the channel box 17 in the cross section of the fuel assembly 13A of this embodiment will be specifically described with reference to FIG. The plurality of fuel rods 11 in the fuel assembly 13A includes fuel rods 1, 2, 5, 6, 7, 8, 9 and 10. Each of the fuel rods 1, 2, 5, 6, 7 and 8 arranged in the fuel assembly 13A are identical to the fuel rods 1, 2, 5, 6 and 7 arranged in the fuel assembly 13 of Example 1. and 8, respectively. That is, the fissile Pu enrichment and the number of the fuel rods 1, 2, 5, 6 and 7 arranged in the fuel assembly 13A are the same as the fuel arranged in the fuel assembly 13 of the first embodiment. Same as fissile Pu enrichment and number of rods 1, 2, 5, 6 and 7, respectively. The fuel rods 8 arranged in the fuel assembly 13A are filled with nuclear fuel material such as depleted uranium and contain burnable poison such as gadolinium (Gd). The gadolinium concentration and the number of the fuel rods 8 are the same as the gadolinium concentration and the number of the fuel rods 8 arranged in the fuel assembly 13 of the first embodiment. The fuel assembly 13A of this embodiment also has the fuel rods 11 arranged in a square grid of 13 rows and 13 columns.

The arrangement of the fuel rods 1, 2, 5, 6, 7 and 8 in the fuel assembly 13A is the same as the arrangement of the fuel rods 1, 2, 5, 6, 7 and 8 in the fuel assembly 13 in Example 1. is the same as Arrangement of the fuel rods 9 and 10 used in the fuel assembly 13A will be described. Each of the fuel rods 9 and 10 is arranged in the second row from the inner surface of the channel box 17 in the fuel rod array in the cross section of the fuel assembly 13 . At the corners facing the control rods 34, the fuel rods 9 are arranged at the corners of the second row in each of the two orthogonal directions with the fuel rods 1 as the base point in the second row from the inner surface of the channel box 17. adjacent to the fuel rods 1. Besides the corners facing the control rods 34, the fuel rods 9 are arranged in the same manner as in the corners facing the control rods 34 at the other three corners. At the corner facing the control rods 34, in each of its two orthogonal directions in its second row, the fuel rods 10 are arranged adjacent to the aforementioned fuel rods 9 arranged in its second row. . At the three corners other than the corners facing the control rods 34, the fuel rods 4 are arranged in the same way as at the corners facing the control rods 34. FIG.

Each of fuel rods 9 and 10 is a part length fuel rod. The active fuel length of each of fuel rods 9 and 10 is the same as that of fuel rod 7, which is 9/10 of the active fuel length of each of fuel rods 1-4, 6 and 8. The fissile Pu enrichment of the fuel rods 9 is 6.2 t%, and the fissile Pu enrichment of the fuel rods 10 is 8.8 t%. Further, eight fuel rods 9 and eight fuel rods 10 are present in the fuel assembly 13A.

The inventors have proposed a fuel assembly having a configuration in which all of the fuel rods 5, 7, 9 and 10, which are part-length fuel rods, in the fuel assembly 13A of this embodiment shown in FIG. 6 are replaced with full-length fuel rods. assumed. This assumed fuel assembly is hereinafter referred to as a second assumed fuel assembly. The inventors obtained the heat generation coefficient of each fuel rod in the cross section of the above-mentioned second assumed fuel assembly by core performance calculation. FIG. 8 shows the calculated heat generation coefficient of each fuel rod in the cross section of such a second hypothetical fuel assembly.

In the cross section of the second assumed fuel assembly, the fuel rods arranged in the cross section are arranged point-symmetrically around the central axis of the fuel assembly (located at (7, 7)). there is

According to the heat generation coefficients of the fuel rods shown in FIG. becomes the largest at The heat generation coefficient of the fuel rods 11 arranged at positions (3,11), (11,3) and (11,11) is "1.38", and the heat generation coefficient of the fuel rods 11 arranged at position (3,3) is It is almost the same as the heat generation coefficient of the fuel rod 11. Each fuel rod 11 located at positions (3,3), (3,11), (11,3) and (11,11) is a channel of the fuel rod array in the cross section of the fuel assembly. They are arranged at the corners of the third row from the inner surface of the box 17 . Fuel rods 5, which are part-length fuel rods (the effective fuel length of the fuel rods is full-length fuel rods, for example, 2 of the effective fuel length of the fuel rods 6) are placed at the positions of the corners of the third row of the fuel assembly 13A. /3) are arranged.

According to FIG. 8, (2,5), (2,9), (5,2), (5,12), (9,2), (9,12), (12,5) and (12 , 9) has a heat generation coefficient of 1.38 or 1.37. Positions (2,5), (2,9), (5,2), (5,12), (9,2), (9,12), (12,5) and (12,9) The fuel rods 11 arranged in the inner surface of the channel box 17 are based on the fuel rods 11 arranged in each corner of the second row from the inner surface of the channel box 17 in the fuel rod arrangement in the cross section of the fuel assembly. In each of the two orthogonal directions along , the fuel rods 11 are arranged third from the fuel rods 11 arranged at each corner of the second row.

According to FIG. 8, (2,4), (2,10), (4,2), (4,12), (10,2), (10,12), (12,4) and (12 , 10) is 1.38 or 1.39. Positions (2,4), (2,10), (4,2), (4,12), (10,2), (10,12), (12,4) and (12,10) In each of the two orthogonal directions along the inner surface of the channel box 17, with the fuel rods 11 arranged at each corner of the second row of the above fuel rod array as the base point, It is arranged second from the fuel rod 11 arranged at each corner of the second row.

Further, according to FIG. 8, (2,3), (2,11), (3,2), (3,12), (11,2), (11,12), (12,3) and The heat generation coefficient of the fuel rod 11 arranged at each position (12, 11) is "1.36" or "1.37". Positions (2,3), (2,11), (3,2), (3,12), (11,2), (11,12), (12,3) and (12,11) In each of the two orthogonal directions along the inner surface of the channel box 17, with the fuel rods 11 arranged at each corner of the second row of the above fuel rod array as the base point, It is arranged first from the fuel rods 11 arranged at the corners of the second row.

Two orthogonal directions along the inner surface of the channel box 17, starting from the fuel rods 1 arranged at each corner of the second row from the inner surface of the channel box 17, of the fuel rod arrangement in the cross section of the fuel assembly 13A In each of them, the fuel rod 1 placed at each corner of the second row to the fuel rod 7 placed at the third row, the fuel rod 1 placed at each corner of the second row to the fuel rod 10 placed at the second row. , and the fuel rods 1 located at each corner of the second row to the fuel rods 9 located at the first row are part length fuel rods. The active fuel length of each of fuel rods 7, 9 and 10 is 9/10 of the active fuel length of, for example, fuel rod 6, which is a full length fuel rod.

A second assumption that the fuel rods 1, 2, 6 and 8 are full length fuel rods in the fuel assembly 13A and the fuel rods 5, 7, 9 and 10 that are part length fuel rods are full length fuel rods. The structure of the fuel assembly has fuel rods 1, 2, 3, 4, 6 and 8 which are full-length fuel rods in the fuel assembly 13 used in Example 1, and fuel rod 5 which is a part-length fuel rod. and 7 are full-length fuel rods, respectively. This is because the enrichment of each of the fuel rods 1, 2 and 6 and the number of each of the fuel rods 1, 2 and 6 in the second assumed fuel assembly are 2 and 6, respectively, and the same as the number of fuel rods, 1, 2, and 6, respectively. The gadolinium concentration of the fuel rods 8 and the number of fuel rods 8 in the second assumed fuel assembly are the same as the gadolinium concentration of the fuel rods 8 and the number of fuel rods 8 in the first assumed fuel assembly. The enrichment of each full-length fuel rod corresponding to each of the fuel rods 5, 7, 9 and 10 in the second assumed fuel assembly and the number of each of these full-length fuel rods are the fuel in the first assumed fuel assembly Each full length fuel rod corresponding to each of rods 5 and 7, and the enrichment of each of fuel rods 3 and 4 being full length fuel rods, and each full length fuel rod corresponding to each of fuel rods 5 and 7, and each full length. It is the same as the number of fuel rods 3 and 4, which are fuel rods.

In this way, although the configuration of the second assumed fuel assembly is the same as the configuration of the first assumed fuel assembly, the The value of each heat generation coefficient at the arrangement position of each fuel rod is higher than the value of each heat generation coefficient at the arrangement position of each fuel rod shown in FIG. generally smaller. The reason why the value of each heat generation coefficient at the arrangement position of each fuel rod in the second assumed fuel assembly is smaller than the value of each heat generation coefficient at the arrangement position of each fuel rod in the first assumed fuel assembly It is explained below.

A period of one operation cycle in a low-rate spectrum boiling water nuclear power plant in which a plurality of second assumed fuel assemblies are loaded in the core is reduced-rate spectrum boiling water in which a plurality of first assumed fuel assemblies are loaded in the core. It is different from the duration of one operation cycle in a type nuclear power plant. Specifically, the period of one operation cycle in the reduced speed spectrum boiling water nuclear power plant using the second assumed fuel assemblies is equal to that of the reduced speed spectrum boiling water nuclear power plant using the first assumed fuel assemblies. is longer than The burnup of the second assumed fuel assembly loaded in the core of the low-moderation spectrum boiling water nuclear power plant is also different from the burnup of the first assumed fuel assembly loaded in the core of the slow-moderation spectrum boiling water nuclear power plant. ing. Specifically, the burnup of the second assumed fuel assembly loaded in the core of the slow spectrum boiling water nuclear power plant is is greater than the burnup of

As described above, the period and burnup of one operation cycle when using the second assumed fuel assembly are different from those when using the first assumed fuel assembly, and the total length in the second assumed fuel assembly The power of the fuel rods differs from the power of the full length fuel rods in the first assumed fuel assembly. As a result, the value of each heat generation coefficient at the position of each fuel rod shown in FIG. 8 for the second assumed fuel assembly is shown in FIG. It is generally smaller than the value of each heat generation coefficient at the arrangement position of each fuel rod.

In the fuel assembly 13A, each of the fuel rods 1, 2, 6 and 8 has its lower end supported by the lower tie plate 14 and its upper end supported by the upper tie plate 15. As shown in FIG. Each of fuel rods 5, 7, 9 and 8 has the same active fuel length. Each of the fuel rods 5 , 7 , 9 and 10 has its lower end supported by the lower tie plate 14 but its upper end is not supported by the upper tie plate 15 . The active fuel length of each of fuel rods 5, 7, 9 and 10 is less than the active fuel length of each of fuel rods 1, 2, 6 and 8, and each of 5, 7, 9 and 10 has a partial length. fuel rods. The active fuel length of each of fuel rods 5, 7, 9 and 10 is longer than half the active fuel length of each of fuel rods 1, 2, 6 and 8, and the length of fuel rods 7, 9 and 10 is Each active fuel length is longer than the active fuel length of the fuel rods 5 . The active fuel length of fuel rod 5 is 2/3 of the active fuel length of each of fuel rods 1-4, 6 and 8, and the active fuel length of fuel rods 7, 9 and 10 is each of fuel rods 1-4, 9/10 of the effective fuel length of 6 and 8 respectively.

The inventors evaluated the MCPR of the fuel assembly 13A of this example using a cooling water flow evaluation method. FIG. 9 shows the MCPR evaluation results. In the evaluation results shown in FIG. 9, the solid line represents the first case in which fuel assemblies having part-length fuel rods were loaded in the core, and the broken line represents the second case in which fuel assemblies having part-length fuel rods were loaded into the core. is shown. In the first case, the fuel assemblies loaded into the core are the fuel assemblies 13A of the second embodiment. In the second case, the fuel assemblies loaded into the core have a configuration in which the part length rods in fuel assembly 13A are replaced by full length rods. The horizontal axis of FIG. 9 indicates the core flow rate (%), and the core flow rate is set to 100% during rated operation of the reactor with a reactor output of 100%. The MCPR was evaluated by changing the core flow rate in the range of 80% to 120%. The vertical axis in FIG. 9 indicates the MCPR obtained by changing the core flow rate.

The MCPR in the first case, in which the core is loaded with fuel assemblies having part-length rods, is greater than the MCPR in the second case, in which the core is loaded with fuel assemblies without part-length rods. Fuel bundles containing part length rods have increased MCPR. The MCPR in the first case increases by about 5% compared to the MCPR in the second case. In order to improve MCPR, it is better to load the core with fuel assemblies containing part-length fuel rods.

This embodiment can obtain each effect produced in the first embodiment. Furthermore, in this embodiment, the inner surface of the channel box 17 is based on the fuel rods 1 arranged at each corner of the second row from the inner surface of the channel box 17 in the fuel rod arrangement in the cross section of the fuel assembly 13A. Part-length fuel rods 10 are arranged in the first row from the fuel rods 1 arranged in each corner of the second row in each of the two orthogonal directions along, and the fuel rod array of the channel box 17 From the fuel rods 1 arranged at the corners of the second row in each of two orthogonal directions along the inner surface of the channel box 17 with the fuel rods 1 arranged at the corners of the second row from the inner surface as the base point Since the fuel rods 9 (the fuel rods 9 are adjacent to the fuel rods 10), which are part-length fuel rods, are arranged in the first rod, the occurrence of the liquid film dryout in the fuel assembly 13A can be prevented from occurring in the fuel assembly of the first embodiment. 13 can be further suppressed.

In the fuel assembly 13A of this embodiment, the fuel rods 9 and 10, which are part-length fuel rods, are arranged in addition to the fuel rods 5 and 7, which are part-length fuel rods. The loading of nuclear fuel material is reduced. However, since the effective fuel length of each of fuel rods 9 and 10 is longer than that of fuel rod 5, the degree of reduction in the loading of nuclear fuel material in fuel assembly 13A is small.

A fuel assembly of Example 3 used in a low speed spectrum boiling water nuclear power plant, which is another preferred example of the present invention, will be described with reference to FIGS. 3, 4, 10 and 11. FIG.

The nuclear reactor 21 of the slow spectrum boiling water nuclear power plant 20 in which the fuel assemblies of this embodiment are loaded into the core is an ABWR.

In the fuel assembly 13B of this embodiment, the fuel rods 11B are also arranged in a square grid of 13 rows and 13 columns. The arrangement of the fuel rods 11B in the fuel assembly 13B is the same as the arrangement of the fuel rods 11 in the fuel assembly 13A of the second embodiment. In the fuel assembly 13B, the fuel rods 5, 7, 9 and 10, which are part length fuel rods used in the fuel assembly 13A of the second embodiment, are arranged at the same positions as the fuel assembly 13A. As shown in FIG. 11, the fuel rod 11B is constructed by filling a cladding tube 41 with a plurality of fuel pellets 40 made of nuclear fuel material.

In the fuel assembly 13A of Example 2, there are four fuel rods 5 that are partial length fuel rods with an effective fuel length of 2/3, and fuel rods 5 and 7 that are partial length fuel rods with an effective fuel length of 9/10. , 9 and 10 in total, the loading amount of the nuclear fuel material is reduced by 2.3% as compared with the fuel assembly 13 of the first embodiment. In low moderation spectrum boiling water reactors, it is desirable to have a high loading of nuclear fuel material in the fuel assemblies.

Therefore, in this embodiment, fuel The diameter of the rod 11B is set to 8.1 mm, which is larger than the diameter of the fuel rod 11 of the fuel assembly 13A of the second embodiment, which is 8.0 mm. As a result, a decrease in the fissile material loading of the fuel assembly due to an increase in the number of part length fuel rods can be compensated for by an increase in the nuclear fuel material loading due to an increase in the diameter of the fuel rods.

The increased fuel rod diameter _dw to compensate for the reduced fissile material loading of the fuel assembly depends on the active fuel length of the part length rods placed in the fuel assembly, so It may be set so as to satisfy the relationship of the formula (2) shown.

π×{(d _w −2×δ _w )/2} ² ×Ls _w
−π×{(d _w/o −2×δ _w/o )/2} ² ×Ls _w/o >0 (2)
where _Lsw is the sum of the active fuel lengths of all the fuel rods arranged in a fuel assembly with part length rods, _dw is the diameter of the fuel rods arranged in a fuel assembly with part length rods, _δw is the cladding thickness of the fuel rods arranged in fuel assemblies with part length rods, and Lsw _/o is the fuel of all fuel rods arranged in fuel assemblies without part length rods. total effective length, dw _/o is the diameter of the fuel rods arranged in fuel assemblies without part length rods, and .delta.w _/o is the diameter of the fuel rods arranged in fuel assemblies without part length rods. is the thickness of the cladding of the fuel rods that have been processed.

(2), the diameter _dw of the fuel rods arranged in the fuel assembly with part-length rods, excluding the thickness of the cladding tube 41, is the same as that of the fuel without part-length rods. By setting it to be larger than the diameter _dw/o of the fuel rods arranged in the assembly, the loading of nuclear fuel material in the fuel assembly 13B with part-length rods can be reduced to can be increased over the nuclear fuel material loading in unloaded fuel assemblies.

According to this embodiment, each effect produced in the second embodiment can be obtained. Furthermore, in this embodiment, since the diameter of the fuel rods 11B arranged in the fuel assembly 13B having part-length fuel rods is set so as to satisfy the equation (2), the fuel assembly having part-length fuel rods The loading of nuclear fuel material in body 13B can be increased over the loading of nuclear fuel material in fuel assemblies that do not have part length rods.

1, 2, 3, 4, 6, 8 ... fuel rods (full-length fuel rods), 5, 7, 9, 10 ... fuel rods (part-length fuel rods), 11, 11B ... fuel rods, 12 ... support rods, 13 , 13A, 13B... fuel assembly, 14... lower tie plate (lower fuel support member), 15... upper tie plate (upper fuel support member), 17... channel box, 20... low speed spectrum boiling water nuclear power plant, 21 22 Reactor pressure vessel 23 Core 34 Control rod 39 Water gap region 40 Fuel pellet 41 Cladding tube.

Claims (12)

a lower fuel support member;
an upper fuel support member;
a channel box, which is a rectangular tube having a square cross-section and having an upper end attached to the upper fuel support member and extending toward the lower fuel support member;
A fuel assembly comprising a plurality of fuel rods filled with nuclear fuel material arranged in a square lattice in the channel box,
The fuel rods include a plurality of first fuel rods having lower ends supported by the lower fuel support member and upper ends supported by the upper fuel support member, and lower ends supported by the lower fuel support member. a plurality of second fuel rods not supported by said upper fuel support member and having an effective fuel length less than that of said first fuel rods;
A fuel assembly according to claim 1, wherein the second fuel rods are arranged in the second row and the third row from the inner surface of the channel box in the fuel rod arrangement in the cross section of the fuel assembly. The plurality of second fuel rods includes a plurality of third fuel rods arranged in the third row and a plurality of fourth fuel rods arranged in the second row,
2. The fuel assembly according to claim 1, wherein the active fuel length of said fourth fuel rod is longer than the active fuel length of said third fuel rod. The third fuel rods are arranged at each of the four first corners of the third row, and the fourth fuel rods are arranged at each of the four second corners of the second row. 3. The fuel assembly according to claim 2, wherein in each of two orthogonal directions along the inner surface of the channel box, the fuel assembly is arranged third from the first fuel rod arranged at the respective second corner . The effective fuel length of the fourth fuel rod is L1, the number of the fourth fuel rods in the fuel assembly is n1, the effective fuel length of the third fuel rod is L2, and the length of the third fuel rod in the fuel assembly is When the number of fuel rods is n2, the effective fuel length of the first fuel rod is L, and the total number of fuel rods included in the fuel assembly is n, the effective fuel length of the fourth fuel rod is L1 and the third 4. The fuel assembly according to claim 2 or 3, wherein the effective fuel length L2 of the fuel rod satisfies the relationship of the following formula (1).
{(L−L1)×n1+(L−L2)×n2}/(L×n)≦0.072 (1) The active fuel length of the third fuel rod is 2/3 of the active fuel length of the first fuel rod, and the active fuel length of the fourth fuel rod is 9/10 of the active fuel length of the first fuel rod. 5. The fuel assembly of claim 4. The fourth fuel rods are arranged in each of two orthogonal directions along the inner surface of the channel box from the first fuel rods arranged at the four second corners of the second row. In addition to being arranged at the third from the first fuel rod arranged at the second corner, the above-mentioned 4. The fuel assembly of claim 3, wherein in each of two orthogonal directions, the fuel rods are also arranged first and second from said first fuel rods arranged at respective second corners. 7. The fuel assembly of claim 6, wherein said third fuel rods are positioned adjacent to a portion of said plurality of fourth fuel rods positioned in said second row. the plurality of first fuel rods includes a plurality of fifth fuel rods containing the nuclear fuel material and no burnable poison, and a plurality of sixth fuel rods containing the nuclear fuel material and the burnable poison;
8. The fuel assembly according to any one of claims 1 to 7, wherein said sixth fuel rod is arranged in the outermost peripheral region of the fuel rod arrangement in the cross section of said fuel assembly. a support rod having a lower end supported by the lower fuel support member and an upper end supported by the upper fuel support member, holding a fuel spacer for bundling the plurality of fuel rods, and a support rod containing no nuclear fuel material; 8. A fuel assembly according to any one of claims 1 to 7, arranged at each of the four corners of the outermost peripheral region of the fuel rod array in the cross section of said fuel assembly. Lsw _is the sum of the active fuel lengths of all the fuel rods arranged in a fuel assembly having part length rods; dw _is the diameter of the fuel rods arranged in said fuel assembly having said part length rods; _w is the cladding thickness of the fuel rods disposed in the fuel assemblies having part length rods; Lsw _/o the total length, dw _/o the diameter of the fuel rods disposed in the fuel assembly not having the part length rods, and the fuel rods not having the part length rods. 10. The fuel rod according to any one of claims 1 to 9, wherein the diameter _dw of the fuel rod satisfies the following formula (2) where .delta.w _{/o is} the thickness of the cladding tube of the fuel rod arranged in the assembly. or the fuel assembly according to item 1.
π×{(d _w −2×δ _w )/2} ² ×Ls _w
−π×{(d _w/o −2×δ _w/o )/2} ² ×Ls _w/o >0 (2) 11. The fuel assembly of any preceding claim, wherein the plurality of fuel rods in the fuel rod array within the channel box are arranged in 13 rows and 13 columns. A core of a nuclear reactor, comprising a plurality of fuel assemblies according to any one of claims 1 to 11.

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