JP2009243905A - Fuel assembly and core of boiling water reactor - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To elongate a fuel, while suppressing a pressure loss or increase of the length of a control rod. <P>SOLUTION: This fuel assembly having a fuel effective length of 3.9 m or longer includes a lower tie plate, a plurality of fuel rods each lower end of which is held by the lower tie plate, one or a plurality of water rods each lower end of which is held by the lower tie plate similarly, a fuel spacer for holding the fuel rods and the water rods, and a channel box for enclosing the fuel spacer. The length of a part including a nuclear fuel material in the fuel rod is 3.9 m or longer, and a domain of the fuel rod including the nuclear fuel material is divided into two domains between the lower tie plate and an upper tie plate for holding each upper end of the fuel rods, and an output per unit height in the second domain is set smaller than that in the first domain through an operation period. <P>COPYRIGHT: (C)2010,JPO&INPIT

Description

  The present invention relates to a fuel assembly and a core used in a boiling water reactor.

In general, in a boiling water reactor (BWR), as shown in Patent Document 1, a coolant (light water) is boiled by heat generated by fission of fissionable materials contained in a plurality of fuel assemblies loaded in a reactor core. Steam generated at the time is taken out and sent to the turbine, and the turbine is directly rotated by the steam to generate power.

  The fuel assembly includes a plurality of fuel rods and water rods, and a lower tie plate and an upper tie plate that respectively support the lower and upper portions thereof, and a fuel spacer that horizontally supports the fuel rods and the water rod, and surrounds them. It consists of a channel box.

  In the BWR, it is desired to increase the energy generated per fuel assembly from the viewpoint of effective utilization of uranium resources and reduction of the number of spent fuel assemblies. For this purpose, it is desired to increase the burnup to increase the energy generated per unit fuel weight and to increase the uranium weight per fuel assembly. In order to increase the burnup, it is necessary to increase the average enrichment of uranium 235 in the fuel assembly (hereinafter simply referred to as the average enrichment).

  In the fuel assemblies described in Patent Document 2 and Non-Patent Document 1, a water channel is disposed in an area where nine fuel rods in the central portion can be disposed, and cooling water is caused to flow therein. It corresponds to the increase. Further, in the fuel assembly described in Non-Patent Document 2, two large water rods are arranged in an area where seven fuel rods in the central part can be arranged, and other fuel rods (long fuel rods). Eight short fuel rods having a shorter length are arranged.

  FIG. 9 is a schematic diagram showing a longitudinal sectional structure of the fuel rod 2. In the figure, the fuel rod 2 has a structure in which fuel pellets 8 are stacked in a fuel cladding tube 14 and are sealed by an upper end plug 16 and a lower end plug 17 while being pressed by a spring 15. The length of the portion filled with fuel pellets is called the effective fuel length, and the space in which the spring 15 is housed is called the plenum. The plenum is provided to keep the internal pressure of the fuel rod within an allowable range when a gaseous fission product (FP gas) generated by the fission reaction is released from the pellet.

In the case of the fuel assembly shown in Non-Patent Document 2, the pellet diameter is 0.96 cm, 66 long fuel rods having an effective fuel length of 371 cm, and 8 short fuel rods having an effective fuel length of 216 cm are provided. That is, the total effective fuel length is about 262 m, and the total volume of fuel pellets is about 18970 cm ³ . In the fuel assembly of Non-Patent Document 2, by using short fuel rods, the flow path area in the two-phase flow region at the top of the fuel assembly having a high pressure loss ratio is expanded. Thereby, the pellet diameter is increased without increasing the overall pressure loss. Therefore, although the total effective fuel length is shorter than that of Patent Document 2 and Non-Patent Document 1, the total volume of fuel pellets is large and the weight of uranium is large.

On the other hand, in the case of the fuel assemblies shown in Patent Document 2 and Non-Patent Document 1, 72 fuel rods having a pellet diameter of 0.94 cm and an effective fuel length of 371 cm are provided, the total effective fuel length is about 267 m, and the total volume of fuel pellets Is about 18540 cm ³ . Therefore, since the total fuel effective length is longer than that of the fuel assembly of Non-Patent Document 2, the average linear power density defined by (heat output of fuel assembly / total fuel effective length) is larger than that of Non-Patent Document 2. About 2% lower.

  When the length of the heat generating portion of the fuel rod is increased, the total fuel effective length can be extended by increasing the length of the short fuel rod in the fuel assembly of Non-Patent Document 2. On the other hand, in the fuel assemblies of Patent Document 2 and Non-Patent Document 1, since all the fuel rods have the same length, it is necessary to extend the effective fuel length inside the fuel rods.

  In recent years, in order to improve the plant utilization rate and reduce the power generation cost, in a boiling water reactor (BWR), a drastic improvement in output and a long-term cycle operation of 13 months or more have been studied. In these operations, since the number of fuel exchangers is increased as compared with the conventional case, the average burnup of the fuel assembly tends to decrease and the fuel cycle cost tends to increase.

  As a countermeasure, as shown in Patent Document 3, the effective fuel length is made longer than the conventional 3.7 m to increase the amount of fuel loaded into the fuel assembly, thereby reducing the number of fuel exchangers. However, all of the above-described known techniques have been devised on the assumption that the total length of the fuel assembly is not changed, targeting the fuel assembly loaded in the existing boiling water reactor core. As a result, the increase in the area (fuel effective length) of the fissile material in the fuel rod of the fuel assembly is not always sufficient, and the improvement effect of the core performance is limited.

Japanese Patent No. 2058634 Japanese Unexamined Patent Publication No. 64-91088 JP 2004-286560 A "About 9x9 BWR fuel" NLR-15, April 1994 “Boiling water nuclear power plant 9 × 9 fuel” HLR-048 Rev. 1, February 1998

  In order to cope with the prolonged operation period and the improvement in output in recent years, further increase in the effective fuel length is desired. When a fuel assembly loaded in an existing boiling water reactor core is targeted, the upper lattice plate and the shroud head that support the fuel assembly are factors that limit the total length of the fuel assembly.

  The effective fuel length of the existing fuel assembly is approximately 3.6 m to 3.8 m, and the upper end of the fuel assembly is approximately 4.4 m. In addition, an upper grid plate is provided in a region of 3.9 m or more from the lower end of the effective fuel length, and the gap between the upper end of the fuel assembly and the shroud head is about 0.7 m. Therefore, there is a possibility that the effective fuel length can be increased from the current standard specification of 3.7 m to 3.9 m to 4.4 m by considering neutron irradiation of the upper support plate or the like. That is, in the new plant, the fuel assemblies elongated from 3.9 m to 4.4 m can be easily loaded by optimizing the arrangement of the upper grid plate and the like.

  Increasing the effective length of the fuel assembly not only increases the fuel load, but also increases the critical heat output (increases the thermal margin) by increasing the heat transfer area. On the other hand, in order to reduce the amount of neutron irradiation, it is necessary to reduce the output in the area of 3.9 m or more from the lower end of the effective fuel length. Will be limited.

  Further, if the effective fuel length is increased, the pressure loss in the fuel assembly increases, and the channel stability is lowered accordingly. Accordingly, the increase of the effective length has not been applied in practice. Furthermore, in existing boiling water reactors, the area above 3.9 m has an upper grid plate, and the effective fuel length is shorter than that due to reasons such as not increasing the amount of neutron irradiation on the upper grid plate. Yes.

  An object of the present invention is to provide a fuel assembly capable of suppressing pressure loss while maintaining the soundness of an upper lattice plate in a fuel assembly having an effective fuel length of 3.9 m or more.

  A lower tie plate, a plurality of fuel rods having a lower end held by the lower tie plate and having a fuel spacer, one or more water rods also having a lower end held by the lower tie plate, and the fuel rod And a channel box surrounding the fuel spacer, wherein the length of the portion containing the nuclear fuel material in the fuel rod is 3.9 m or more and 4.4 m or less, and from the lower end of the fuel region, from the lower end of the fuel region In a fuel assembly in which a region having a height of less than 3.9 m is a first region, and a region above the first region is a second region, the output per unit height in the second region is output throughout the first period. To make it smaller.

According to the present invention, it is possible to increase the amount of fissile material in the core while maintaining the soundness of the upper lattice plate and suppressing an increase in the pressure loss of the fuel assembly. It becomes possible to stay in the reactor core and fuel economy is improved. Further, the thermal output of the core can be increased without increasing the number of fuel assemblies. Furthermore, the increase in the two-phase flow pressure loss can be suppressed by lowering the output of the fuel upper region where the proportion of steam is large, and since the output of the second region is small, the shutdown characteristics of the core can be improved even if the control rod does not enter the upper end of the fuel. Can be maintained.

  As described above, the inventors conducted various studies on the fuel assembly that increases the loading amount of the fissile material while maintaining the soundness of the upper lattice plate and suppressing the increase in the pressure loss of the fuel assembly. . As a result, it is desirable for the inventors to set the effective fuel length of the fuel assembly to 3.9 m or more, and in the region where the effective fuel length is 3.9 m or more, the power distribution does not impair the thermal margin. I found a new thing. The above study results will be specifically described.

  One way to improve the fuel economy of the core is to increase the length of the fuel assembly. However, there is a limitation due to the upper grid plate described above, and an increase in the pressure loss of the fuel assembly must be avoided. . The soundness of the upper grid plate is achieved by intentionally creating a low-power region in the region where the effective fuel length is 3.9 m or more, but the pressure loss is generally increased by increasing the length of the fuel assembly. To do. In order to make the pressure loss equivalent to the conventional one, it can be realized by reducing the coolant flow rate of the fuel assembly, but if the output distribution is equivalent to the conventional one, the thermal allowance and stability margin of the fuel assembly will not be Decrease.

  Therefore, the inventors increased the thermal margin of the fuel assembly by making the region having an effective fuel length of 3.9 m or more as a region that does not impair the thermal margin, and lengthened the pressure loss to be the same as the conventional one. I came up with the idea.

  The inventors examined an output distribution that does not impair the thermal margin, divides the fuel assembly 1 shown in FIG. 1 into a lower first region and an upper second region, and the output distribution as shown in FIG. did. FIG. 3 shows the thermal margin in the fuel axis direction for the fuel rod 2 with the severest thermal margin at this time. The thermal margin of the fuel rod 2 is an index of the liquid film thickness that covers the fuel rod 2 when the coolant is in an annular spray flow state including voids, and when the thermal margin of the fuel rod 2 becomes 1. The part of the liquid film covering the fuel rod 2 has a thickness of 0 and is dried out. When this thermal margin is verified in the direction of the fuel axis, the thermal margin is reduced because the water forming the liquid film has a large effect in the heat generating portion. On the other hand, in the fuel spacer portion, a large number of droplets in the annular spray flow adhere to the liquid film covering the fuel rod 2 when the flow is disturbed by the fuel spacer, so that the thermal margin increases.

  In the present invention having the output distribution as shown in FIG. 2, the thermal margin does not decrease in the second region as shown in the explanatory diagram of the thermal margin in FIG. That is, when the thermal margin reduction effect due to heat generation in the region between the fuel spacers or between the fuel spacer and the fuel effective length upper end is VA, and the thermal margin improvement effect by the fuel spacer is SP, the output is a certain value. The inventors have newly found that SP ≧ VA when the value becomes smaller.

  FIG. 4 is an explanatory diagram of the thermal margin ratio of the fuel assembly. That is, in the region where the void ratio is high, the gas-liquid two-phase flows in an annular spray flow state, and between the fuel spacers or between the fuel spacer and the upper end of the fuel effective length, the output ratio in the total output of the fuel assembly in the region The exit / inlet thermal margin ratio of the region to The thermal margin tends to decrease as the output of the region increases, but the thermal margin improves only when the output of the region is 8% or less.

  The fuel assembly 1 according to the first embodiment of the present invention reflecting the above examination results will be described. In FIG. 1, the lower end of the fuel assembly 1 is held by the lower tie plate 6 and the lower tie plate 6, and the lower end of the fuel tie 2 having the fuel spacer 4 is held by the lower tie plate 6. One or a plurality of water rods 3 and a channel box 7 surrounding the fuel rods 2 and the fuel spacers 4 are provided. The length of the portion containing the nuclear fuel material in the fuel rod 2 is 3.9 m or more, and a region less than 3.9 m from the lower end of the region of the fuel rod 2 containing the nuclear fuel material is defined as the first region, 3 An area of 9 m or more is set as the second area. FIG. 5 shows a horizontal sectional view of the fuel assembly 1 in the first region, and FIG. 6 shows a horizontal sectional view of the fuel assembly 1 in the second region. The density of the fuel rods 2 in FIG. 6 is set to be smaller than 1/2 of FIG.

  As a result, as shown in FIG. 2, the output distribution can be lowered in the upper second region as compared with the conventional case. In this case, the result of evaluating the change in the thermal margin at the upper part of the fuel is shown in FIG. As shown in FIG. 3, there is a region where the thermal margin does not decrease in the height direction at the upper part when the output becomes lower than a certain level at the upper part. This is because there is an output in which the thermal margin increasing effect by the fuel spacer 4 and the evaporation of the liquid film due to the heat generation of the fuel rod 2 are balanced or the thermal margin increasing effect by the fuel spacer 4 is larger. In the present invention, it has been discovered that when the above newly discovered phenomenon is used, there is a great merit that has not been conventionally known when the fuel assembly 1 is lengthened.

  If the first region height is less than 3.9 m and the effective fuel length is increased, the pressure loss in the fuel assembly 1 increases. Regarding this, as a means for reducing the output of the region where the pressure loss is large due to the mixed flow of water and steam above the fuel, the pressure loss can be reduced by thinning the fuel rods in the second region, etc. Since the number of fuel rods is reduced, the target output can be reduced naturally. Further, the coolant passage area may be expanded by reducing the number of fuel rods in the second region to reduce the number of fuel rods. Also in this case, since the fuel decreases in the second region, the output naturally decreases.

  In addition, when the effective fuel length is increased, the channel stability decreases. This also reduces pressure loss by thinning out the fuel rods in the second region, etc., and reduces the number of fuel rods by reducing the number of fuel rods in the second region. This can be solved to some extent by enlarging the coolant channel area. However, since the channel stability is not satisfied when the length exceeds a certain length, this is the upper limit for increasing the effective fuel length.

  Next, a second embodiment for reducing the output distribution of the second region will be described. By adjusting the fuel concentration, the same effect as in the first embodiment is obtained. For example, it is effective that the average uranium enrichment in the second region is less than or equal to the average uranium enrichment in the first region. Specifically, the average uranium enrichment in the second region is ½ of the average uranium enrichment in the first region. Furthermore, a blanket region having a natural uranium enrichment or less can be provided in the second region.

  Next, a third embodiment for reducing the output distribution of the second region will be described. The length of the portion in which the nuclear fuel material is contained in the fuel rod 2 is 3.9 m or more and 4.4 m or less, and an area less than 3.9 m from the lower end of the fuel area is a first area, an area of 3.9 m or more The second region has at least one fuel spacer in the second region, and is a region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or the most downstream fuel. If the output of the region surrounded by the spacer and the other fuel spacer provided upstream one stage is 8% or less of the total output of the fuel assembly throughout the operation period, a desired output distribution can be obtained.

  The following are effective as specific means for reducing the output. That is, in the second region, the region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or the most downstream fuel spacer and another fuel spacer provided on the upstream side of the first region. The average number of fuel rods in the enclosed region is set smaller than the average number of fuel rods in the first region. In the second region, a region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or another fuel spacer provided on the upstream side with the most downstream fuel spacer. The average number of fuel rods in the enclosed area is ½ of the average number of fuel rods in the first area. Furthermore, in the second region, a region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or another fuel spacer provided on the upstream side with the most downstream fuel spacer. The average uranium enrichment of the enclosed region is set lower than the average uranium enrichment of the first region, and specifically, it is set to ½ of the average uranium enrichment of the first region. Further, the second region is surrounded by the region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or the most downstream fuel spacer and another fuel spacer provided on the upstream side of the first region. A blanket region that is below the enrichment of natural uranium may be included in the region.

  When the output of the second region is lowered in this way, conventionally, the control rod 91 inserted from below to the vicinity of the upper end of the effective fuel length may be extended to the vicinity of the upper end of the first region as shown in FIG. As a result, when the effective fuel length is increased, the control rod length, which normally increases in proportion, can be made smaller than the proportion. Further, the upper end of the control rod is not necessarily near the upper end of the first region, but may be in the middle of the second region. Further, when the output of the second area is sufficiently small, it may be below the upper end of the first area. As a result, the number of constituent members of the core of the boiling water reactor can be reduced, and economic efficiency can be improved.

  Further, in this embodiment, the thermal margin is large in the second region, and it is considered that strict monitoring is unnecessary. In general, a plurality of neutron detectors 19 in the neutron instrumentation tube 18 provided for monitoring the output of the core and the like are installed at equal intervals (FIG. 11). However, if strict monitoring of the second region is not required, the neutron detector 19 in the neutron instrumentation tube 18 can be disposed only in the first region as shown in FIG. Thereby, the monitoring performance of the first region is improved by using the same number of neutron detectors, and the performance of the core can be improved.

  Further, in this embodiment, the effective fuel length is lengthened, and the effect when the technique of Patent Document 1 is applied becomes larger, and the fuel economy can be improved more effectively. As described above, according to the present embodiment, various economic improvements can be made in addition to the fuel economic improvement due to the increase in the amount of fuel loaded in the core.

It is a longitudinal cross-sectional view which shows the structure of the fuel assembly of the Example of this invention. It is explanatory drawing which shows the axial direction output distribution of the fuel assembly of the Example of this invention. It is explanatory drawing which shows the axial direction thermal margin of the fuel assembly of the Example of this invention. It is explanatory drawing which shows the thermal margin ratio with respect to the output ratio between the fuel spacers of this invention. It is a cross-sectional view which shows the structure of the fuel assembly of the Example of this invention. It is a cross-sectional view which shows the structure of the fuel assembly of the Example of this invention. It is a longitudinal cross-sectional view which shows a part of core of the Example of this invention. It is a longitudinal cross-sectional view which shows a part of core of the Example of this invention. It is a schematic diagram which shows the structure of the conventional fuel rod. It is a longitudinal cross-sectional view which shows a part of core of the conventional boiling water reactor. It is a longitudinal cross-sectional view which shows a part of core of the conventional boiling water reactor.

Explanation of symbols

  DESCRIPTION OF SYMBOLS 1 ... Fuel assembly, 2 ... Fuel rod, 3 ... Water rod, 4 ... Fuel spacer, 5 ... Upper tie plate, 6 ... Lower tie plate, 7 ... Channel box, 8 ... Fuel pellet, 12 ... Rise pipe, 13 ... Down pipe, 14 ... Fuel cladding pipe, 15 ... Spring, 16 ... Upper end plug, 17 ... Lower end plug, 91 ... Control rod

Claims (21)

  A lower tie plate, a plurality of fuel rods having a lower end held by the lower tie plate and having a fuel spacer, one or more water rods having a lower end held by the lower tie plate, the fuel rods, A channel box surrounding the fuel spacer, wherein the length of the portion of the fuel rod containing nuclear fuel material is 3.9 m to 4.4 m, and the height from the lower end of the fuel region to the lower end of the fuel region Is set to be a first area, an area above the first area is a second area, and an output per unit height in the second area is set to be smaller than the first area throughout the operation period. A fuel assembly.   A lower tie plate, a plurality of fuel rods having lower ends held by the lower tie plate and having fuel spacers, one or more water rods having lower ends held by the lower tie plate, the fuel rods, A channel box surrounding the fuel spacer, wherein the length of the fuel rod containing the nuclear fuel material is 3.9 m or more and 4.4 m or less, and the height from the lower end of the fuel region to the lower end of the fuel region is A region less than 3.9 m is defined as a first region, a region above the first region is defined as a second region, and the average number of fuel rods in the second region is equal to or less than the average number of fuel rods in the first region. Fuel assembly.   A lower tie plate, a plurality of fuel rods having lower ends held by the lower tie plate and having fuel spacers, one or more water rods having lower ends held by the lower tie plate, the fuel rods, A channel box surrounding the fuel spacer, wherein the length of the fuel rod containing the nuclear fuel material is 3.9 m or more and 4.4 m or less, and the height from the lower end of the fuel region to the lower end of the fuel region is A region less than 3.9 m is defined as a first region, a region above the first region is defined as a second region, and an average uranium enrichment in the second region is equal to or less than an average uranium enrichment in the first region. Fuel assembly.   A lower tie plate, a plurality of fuel rods having lower ends held by the lower tie plate and having fuel spacers, one or more water rods having lower ends held by the lower tie plate, the fuel rods, A channel box surrounding the fuel spacer, wherein the length of the fuel rod containing the nuclear fuel material is 3.9 m or more and 4.4 m or less, and the height from the lower end of the fuel region to the lower end of the fuel region is An area less than 3.9 m is a first area, an area above the first area is a second area, and the average uranium enrichment of the second area is ½ of the average uranium enrichment of the first area. A fuel assembly characterized by   A lower tie plate, a plurality of fuel rods having lower ends held by the lower tie plate and having fuel spacers, one or more water rods having lower ends held by the lower tie plate, the fuel rods, A channel box surrounding the fuel spacer, wherein the length of the fuel rod containing the nuclear fuel material is 3.9 m or more and 4.4 m or less, and the height from the lower end of the fuel region to the lower end of the fuel region A fuel assembly, wherein a region of less than 3.9 m is a first region, a region above the first region is a second region, and the second region includes a blanket region that is less than or equal to the enrichment of natural uranium body.   The fuel assembly design method according to any one of claims 1 to 5, wherein the thermal margin improvement effect by the fuel spacer in the second region is SP and the thermal by evaporation of water determined from the average line output of the upper region. When the margin reduction effect is VA, the average line output of the second region is determined so that SP ≧ VA, then the length of the first region is determined, and then the length of the second region is set to the channel stability. A fuel assembly design method characterized by being determined from the limits.   A fuel assembly designed by the designing method according to claim 6.   A reactor core including at least one type of fuel assembly among the fuel assemblies according to any one of claims 1 to 5 or claim 7. A lower tie plate, a plurality of fuel rods having lower ends held by the lower tie plate and having fuel spacers, one or more water rods having lower ends held by the lower tie plate, the fuel rods, A channel box surrounding the fuel spacer, the length of the portion containing the nuclear fuel material in the fuel rod is 3.9 m or more and 4.4 m or less, and a region less than 3.9 m from the lower end of the fuel region is a first region When the region of 3.9 m or more is defined as the second region, the second region has at least one fuel spacer, and is located between the most downstream fuel spacer in the second region and the upper end of the effective fuel length. The power output in a certain region or the region surrounded by the most downstream fuel spacer and the other fuel spacer provided upstream one stage is 8% or less of the total output of the fuel assembly throughout the operation period. Fuel assembly, characterized the door.   The second region is surrounded by a region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or the most downstream fuel spacer and another fuel spacer provided upstream one stage. 10. The fuel assembly according to claim 9, wherein the average number of fuel rods in the region is smaller than the average number of fuel rods in the first region.   The second region is surrounded by a region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or the most downstream fuel spacer and another fuel spacer provided upstream one stage. 11. The fuel assembly according to claim 10, wherein the average number of fuel rods in the region is ½ of the average number of fuel rods in the first region.   The second region is surrounded by a region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or the most downstream fuel spacer and another fuel spacer provided upstream one stage. The fuel assembly according to claim 9, wherein the average uranium enrichment in the region is equal to or less than the average uranium enrichment in the first region.   The second region is surrounded by a region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or the most downstream fuel spacer and another fuel spacer provided upstream one stage. 13. The fuel assembly according to claim 12, wherein the average uranium enrichment in the region is ½ of the average uranium enrichment in the first region.   The second region is surrounded by a region between the most downstream fuel spacer and the upper end of the effective fuel length in the second region, or the most downstream fuel spacer and another fuel spacer provided upstream one stage. 14. The fuel assembly according to claim 13, wherein the region includes a blanket region that is less than or equal to the enrichment of natural uranium.   The fuel assembly according to any one of claims 9 to 14, wherein the water rod is a spectrum shift type water rod provided with an ascending pipe and a descending pipe communicating with an internal flow path at an upper part inside the fuel assembly. A fuel assembly characterized by being.   A reactor core including at least one type of fuel assembly among the fuel assemblies according to any one of claims 9 to 15.   A boiling water reactor core including at least one fuel assembly according to any one of claims 9 to 15, wherein a control rod inserted from a lower part of the core is inserted only in the lower region. The core of a boiling water reactor.   In the core of a boiling water reactor including at least one fuel assembly according to any one of claims 9 to 15, a control rod inserted from the lower part of the core is not inserted to the upper end of the upper region. The core of a boiling water reactor.   A boiling water reactor core including at least one fuel assembly according to any one of claims 9 to 15, wherein the neutron detector in the neutron instrumentation tube in the core is only at a position corresponding to the lower region. A boiling water reactor core characterized by comprising.   16. The fuel assembly according to any one of claims 9 to 15, wherein at least one pair of an ascending pipe and a descending pipe communicating with an internal flow path at an upper part is present inside the fuel assembly, and a lower end of the ascending pipe is a lower part. It is below the tie plate, has an opening for guiding the cooling water to the ascending pipe below the lower tie plate in the previous period, the lower end of the descending pipe is above the lower tie plate, and has a coolant outlet near its lower end. A fuel assembly characterized by   A core of a boiling water reactor including at least the fuel assembly according to claim 20.

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