JP2011075496A - Fuel assembly - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a fuel assembly having a 10×10 or larger array which enables high burnup extension for BWR fuels, and significant improvement in the stability and reliability of a core, by providing a fuel deficit portion or non heat-generating fuel portion in a lower part of the core, without changing a basic core structure. <P>SOLUTION: In a fuel assembly for a boiling water reactor having fuel rods 51 arranged in a N×N square lattice form, a water box 52 is arranged at the center, having two or more fuel rods in the fuel rods being fuel rods 51a, and each of which having the fuel deficit portion in its lower part. <P>COPYRIGHT: (C)2011,JPO&INPIT

Description

  The present invention relates to a fuel assembly such as a boiling water reactor, and more particularly to a high burnup fuel assembly of 10 × 10 or more with improved stability and reliability.

  In a boiling water nuclear power plant (hereinafter referred to as “BWR”), a reactor pressure vessel 2 is provided in a reactor containment vessel 1 as shown in FIGS. A reactor core 3 including a coolant 4 (for example, water), a plurality of fuel assemblies 30, a control rod 32, and the like is accommodated therein. The coolant 4 is forcibly circulated by the recirculation system 8 and is in a state where saturated water and saturated steam are mixed by receiving heat generated by nuclear fission of uranium 235 (hereinafter referred to as “U235”) in the core 3. Move to the top of the core. And it will be in a dry state with the steam-water separator and steam dryer which are not shown in figure, and will be sent to the turbine 10 via the main steam piping system 9 connected to the reactor pressure vessel 2, and will drive a turbine. The generator 15 is rotated by the driving of the turbine to generate power. The steam that has worked in the turbine is introduced into the condenser 16 to become condensed water. After the temperature is raised by the feed water heater 18, the steam is supplied again into the reactor pressure vessel by the feed water pump 19.

  FIG. 16 shows a typical operating characteristic diagram of the BWR configured as described above. Normal operation is performed on the lines of the rated output curve, design flow control curve, stability limit curve, minimum pump speed curve, cavitation limit curve, maximum pump speed curve and in the area surrounded by them and on the natural circulation curve . In the example of FIG. 16, the rated output is achieved when the core flow rate is 85% (point A) to 105% (point B). Reactor operation is usually near the 85% flow rate at the beginning of the cycle, moves to a higher flow rate as the cycle burnup progresses to take advantage of the reactivity gain due to the increased coolant flow rate, and 105% flow rate at the end of the cycle. It becomes a neighborhood. In addition, the core nuclear power design is usually designed so that the core axial power distribution has a lower strain at the beginning of the cycle, and the void ratio at the core outlet is increased. This promotes the generation of plutonium 239 (hereinafter referred to as “Pu239”) which is a fissile material from uranium 238 (hereinafter referred to as “U238”). From the middle to the end of the cycle, Pu239 is burned in addition to U235, so the power distribution in the axial direction of the core tends to be upwardly distorted. The core with improved economic efficiency is optimized by combining the core flow rate adjustment and the nuclear design that enables the axial power distribution adjustment.

  More specifically, the core 3 of the boiling water reactor is configured as shown in FIG. A plurality of fuel assemblies 30 are loaded in the core, and each of the fuel assemblies 30 is covered with a channel box 31 as shown in FIG. The channel box 31 of the BWR is a thin rectangular tube that is long in the vertical direction, and its cross section is usually a pseudo square having a quadrant circle having a radius of about 1 cm at the corner.

  A plurality of local output region monitors 33 (hereinafter referred to as “LPRM33”) are arranged in the core to detect neutron flux. Furthermore, each channel box 31 is supported by a core support plate 34 and an upper lattice plate 35 and surrounded by a cylindrical shroud 36 as shown in FIG. The coolant 4 flows from below into the channel box 31 via the orifice of the fuel support bracket 37 and the lower tie plate, is heated by the fuel assembly 30, generates steam (void) by boiling, and gas-liquid Two-phase flow. The effective fuel assembly length of a commercial BWR that is currently operating is about 3.7 m.

  A typical fuel assembly 30 employed in a conventional boiling water reactor includes an upper tie plate 61 and a lower tie plate 62, as in the example of the 9 × 9 fuel assembly shown in FIG. A plurality of fuel rods and water rods 52 held at both ends by these tie plates, a fuel spacer 63 that bundles these fuels, and a channel box 31 that surrounds the fuel rod bundle and is attached to the upper tie plate 61 (Patent Document) 1).

  In the example shown in FIG. 19, a plurality of fuel rods 51 employ partial length fuel rods 51 b whose height does not reach the upper tie plate. The fuel assembly 30 is usually loaded in a lattice shape on the core of a nuclear reactor. When the fuel assembly 30 is loaded in a lattice shape, the distribution of water as a moderator increases relatively in the portion between the water rod portion and the adjacent channel box 31 (water gap portion) in the upper part of the core. .

  As the fuel assembly 30 in the boiling water reactor as described above, since commercial power generation has been carried out in Japan, the 7 row × 7 column type, the 8 row × 8 column type, the improved 8 row × 8 column type, High burn-up 8 rows x 8 columns and high burn-up 9 rows x 9 columns have been adopted.

  These improvements will increase the capacity of fissile material per fuel assembly, and the optimization of the distribution of enrichment in the assembly and the optimal arrangement of the flammable poisons will result in higher burnup and longer operating cycles. The economics of the core is improving. In the high burnup 8 × 8 fuel to 9 × 9 fuel, the total heat transfer area increases as the number of fuel rods increases, so that the limit output characteristics are improved.

  In addition, deterioration of stability based on a nuclear factor due to an increase in the absolute value of the void reactivity coefficient accompanying an increase in burnup / long-term operation cycle is prevented by the use of two large diameter water rods 52 and the like.

  It should be noted that the deterioration of the stability based on the thermohydraulic factor due to the increase of the two-phase pressure loss accompanying the increase of the assembly lattice is due to the reduction of the spacer pressure loss coefficient, the partial length fuel rod of the eight upper defects, and the lower part of the high pressure loss type This is prevented by adopting a tie plate. That is, the reduction of the spacer pressure loss coefficient and the eight partial length fuel rods having a length that is about 2/3 of the full length fuel rod contribute to the reduction of the two-phase flow pressure loss. This contributes to an increase in fluid pressure loss. As a result, the ratio of single-phase flow pressure loss pressure loss to total pressure loss is increased, and the stability of the core is improved.

  Here, the stability means that when the operating point becomes a low flow rate / high output state at the time of starting or stopping the plant, or when a transient change such as tripping of one recirculation pump occurs in the plant, the operating point is low. It means the damping characteristic of neutron flux vibration when the flow rate / high power is shifted.

  The core is desirably stable in the entire operation region, and is confirmed by analyzing that the reduction ratio, which is a stability determination parameter, is less than 1.0. On the contrary, an operation region having a small margin with respect to the reduction ratio of 1.0 is excluded by a selected control rod (hereinafter referred to as “SRI”) or a stability limiting curve.

  The types of stability include channel stability with particular attention to the thermo-hydraulic stability of the highest output channel, core stability (fundamental mode stability), which is neutron flux oscillation with the entire core phase aligned, and core There is region stability (stability of higher-order modes) that is neutron flux oscillation that has an axis of symmetry in the circumferential direction and is 180 degrees out of phase. The sensitivity of each stability to the axial power distribution is more severe as the core stability is generally flat, and the channel stability and region stability are more severe as the lower peak distribution. In the core stability, the sensitivity to the axial power distribution differs from other stability. The core stability has a large effect of nuclear feedback, which has a large effect when the power peak is high at a high void fraction. This is because it appears as

  FIG. 20 shows the core stability analysis result of a BWR plant loaded with all 9 × 9 uranium fuel. Here, the stability analysis is performed using a frequency domain stability analysis code that models a BWR core including a coolant recirculation system. The core stability reduction ratio is the largest at the maximum output point on the minimum pump speed curve, and the reduction ratio is 0.67 in the core loaded with all 9 × 9 uranium fuel. This analysis was performed by applying a peaking coefficient of 1.15 (5/24 node position) representing a flat distribution as the power distribution in the axial direction of the core.

  In recent years, a high burnup 10-row × 10-column fuel assembly has also been studied, but here again, in order to ensure the stability of the core, it is necessary to use a plurality of partial fuel rods with upper defects. It has been proposed (Patent Document 2).

On the other hand, a technique for preventing an excessive increase in internal pressure of a fuel rod accompanying a high burnup (Patent Document 3) and a technique for preventing an inflow of foreign matter into a fuel assembly (Patent Document 4) are also proposed.
As described above, in the BWR core, the deterioration of stability has been prevented by various design improvements so far. On the other hand, the high burnup of light water reactor fuel is aimed at effectively using uranium resources and reducing the amount of spent fuel. Along with this, the number of fuel lattices increases (for example, 10 × 10 or more), and the uranium enrichment tends to increase. Adoption of mixed oxide fuel is also in full swing.

JP 2006-184174 A JP 2005-134179 A Japanese Patent Laid-Open No. 5-1340669 JP 2003-172790 A

  It is known that the stability of BWR is generally stabilized when the core flow rate is high. This is partly because the time required for the voids to pass through the core (fuel channel) is shortened.

  Since the coolant is not saturated at the core inlet, the coolant is a single-phase flow at the bottom of the core. The above stability can be understood as indicating the responsiveness when the disturbance generated at the core inlet part returns to the core inlet part as a disturbance through various feedback loops formed in the core and RPV. it can. Generally, since the stabilization occurs when the response delay is small, the core is stabilized when the core inlet flow rate is high. When viewed as the flow velocity in the channel, the margin for stability increases when the flow velocity at the inlet of the assembly where the single-phase flow is dominant is high. In order to increase the flow velocity at the assembly inlet, for example, the flow area in the channel may be reduced.

  However, a decrease in the channel area in the channel results in an increase in the pressure loss in the channel. In particular, the pressure loss in the two-phase flow portion leads to an increase in the gain of the loop, and thus acts on the side of reducing the margin for stability.

  In addition to the above, as a means for improving the stability of the core, it is known to increase the ratio of the single-phase flow pressure loss to the total pressure loss. An example of a fuel assembly that uses this effect is a 9 × 9 fuel assembly of a conventional design, and the stability is improved by adopting an upper deficient partial length fuel rod and a high-pressure loss lower tie plate. (Patent Documents 1 and 2).

  However, this measure provides channel stability under conditions where the differential pressure between the upper and lower cores is almost constant, or region stability under conditions where the total core inlet flow rate is almost constant (high-order mode stability). The core stability (fundamental mode stability), which deals with the propagation of density waves along the coolant flow in the pressure vessel and the core core feedback is dominant It is known that a large improvement effect cannot be expected. For example, the degree of improvement in the stability reduction ratio at the minimum pump speed maximum power point is about 10% for region stability and 20% for core stability when it is 20% for channel stability. It is about 5%.

  On the other hand, as a means for further improving the economics of the BWR core, when the high burnup 10 × 10 fuel is introduced, the fuel pellet enrichment becomes high in order to achieve high burnup, As the nuclear feedback effect increases and the number of fuel rods increases, the assembly pressure loss increases. Therefore, it is necessary to propose not only a measure for preventing the core stability from being lowered but also a measure for improving the fundamental stability (particularly the core stability). Furthermore, there is a need for early introduction of core fuel that can utilize high enrichment (greater than 5%) uranium oxide fuel or mixed oxide fuel without reducing stability margins. Further, development of a fuel assembly capable of suppressing an increase in the internal pressure of the fuel rod or preventing foreign matter from flowing into the fuel assembly while improving the stability is awaited.

  Thus, as shown in Patent Documents 1 and 2, conventionally, proposals have been made to improve the stability of the core even when the fuel loaded in the BWR core is increased in burnup. However, it does not sufficiently secure the stability and reliability of high burnup fuel assemblies of 10x10 or higher that will be put to practical use in the future. It was.

  The present invention has been made to solve the above-mentioned problems, and is based on the BWR fuel design and manufacturing technology cultivated so far, such as partial-length fuel rods and large-diameter water rods. An object of the present invention is to provide a fuel assembly with a high burn-up of 10 × 10 or more that has improved stability and reliability by adopting an assembly structure.

  In order to solve the above problems, a fuel assembly of a boiling water reactor according to the present invention is a boiling water atom having fuel rods arranged in an N × N square lattice and a water box arranged in the center. In the fuel assembly of the furnace, a plurality of fuel rods among the fuel rods are lower deficient fuel rods.

  Also, a fuel assembly for a boiling water reactor according to the present invention is a fuel assembly for a boiling water reactor having fuel rods arranged in an N × N square lattice and a water box arranged in the center. The fuel rod is characterized in that a fuel pellet defect region is formed in a lower region, and a cylindrical portion provided with a large number of small holes is disposed in the defect region.

  Also, a fuel assembly for a boiling water reactor according to the present invention is a fuel assembly for a boiling water reactor having fuel rods arranged in an N × N square lattice and a water box arranged in the center. The fuel rod is composed of a plurality of full length fuel rods loaded with natural uranium or deteriorated uranium pellets in the lower fuel pellet defect region and a plurality of upper defect fuel rods formed with the fuel pellet defect region in the upper part. It is characterized by.

  According to the present invention, by providing a fuel deficient part or a non-heat generating fuel part at the bottom of the core without changing the basic core structure, it is possible to greatly increase the burnup of BWR fuel and the stability and reliability of the core. A fuel assembly of 10 × 10 or more can be provided.

  Further, when the cylindrical space portion is disposed in the fuel deficient region below the core, it is possible to obtain the effect of suppressing the increase in the fuel rod internal pressure and the effect of preventing the inflow of foreign matter, in addition to the stability improvement effect.

  As described above, according to the present invention, the stability and reliability of a BWR core that employs a fuel assembly of 10 × 10 or more are greatly improved, and the economics due to the high burnup of the fuel assembly are also improved. be able to.

(A) is a lower cross-sectional view of the fuel assembly according to the first embodiment of the present invention, and (b) is an upper cross-sectional view. (A) is the longitudinal cross-sectional view of A line of the fuel assembly of FIG. 1, (b) is the longitudinal cross-sectional view of B line. The upper tie plate structure figure of the fuel assembly concerning a 1st embodiment of the present invention. The core stability analysis result figure which concerns on the 1st Embodiment of this invention. (A) is a lower cross-sectional view of a fuel assembly according to a second embodiment of the present invention, and (b) is an upper cross-sectional view. (A) is a longitudinal cross-sectional view of the fuel assembly of FIG. 5 along line A, and (b) is a longitudinal cross-sectional view of line B. The lower structure figure of the lower defect | deletion fuel rod which concerns on the 2nd Embodiment of this invention. (A) is a lower cross-sectional view of a fuel assembly according to a third embodiment of the present invention, and (b) is an upper cross-sectional view. (A) is the longitudinal cross-sectional view of A line of the fuel assembly of FIG. 8, (b) is the longitudinal cross-sectional view of B line. The lower-structure figure of the fuel assembly which concerns on the 4th Embodiment of this invention. (A) is a lower cross-sectional view of a fuel assembly according to a third embodiment of the present invention, (b) is a central cross-sectional view, and (c) is an upper cross-sectional view. (A) is a longitudinal cross-sectional view of line A of the fuel assembly of FIG. 11, and (b) is a vertical cross-sectional view of line B. (A) is a lower cross-sectional view of a fuel assembly according to a fifth embodiment of the present invention, (b) is a central cross-sectional view, and (c) is an upper cross-sectional view. (A) is the longitudinal cross-sectional view of A line of the fuel assembly of FIG. 13, (b) is the longitudinal cross-sectional view of B line. The whole block diagram of a boiling water nuclear power plant. Operation characteristic diagram of boiling water reactor. A sectional view of the entire core of a boiling water reactor. The relationship diagram of the core shroud and assembly arrangement of a boiling water reactor. The block diagram of the conventional 9x9 fuel assembly. The stability analysis result figure of the conventional fuel assembly (9x9) loading core.

First, the basic configuration and function of the fuel assembly according to the present invention will be described.
The basic structure of the fuel assembly according to the present invention is characterized in that a fuel deficient region where no fuel pellets are arranged or a non-heat generating fuel region is provided in a lower region of the fuel assembly.

  In other words, in the present invention, the axial output near the core inlet (approximately 1/4 or less of the effective length) is relatively small, and that amount is downstream from the vicinity of the inlet (approximately 1/4 or more of the effective length). By compensating for the increase in output, the boiling start point is moved to the downstream side, so that the single-phase flow length in the fuel assembly channel is longer than the single-phase flow length in the conventional BWR fuel assembly channel. Thereby, the effect of shortening the channel passage time of the void is obtained, and the stability of the core (core stability, region stability and channel stability) is improved.

  At this time, the loading amount of the nuclear material can be increased by about 8% by setting the lattice shape of the 9 × 9 fuel assembly adopted in the recent BWR to 10 × 10 fuel. Therefore, high burnup is achieved using 10 × 10 fuel, and the core stability is effectively improved by providing a fuel deficient region or a non-heat generating fuel region. Analysis using the frequency domain stability solution code confirms that the core stability reduction ratio is improved by about 10% by deleting the lower part of the 10 × 10 fuel assembly for about 1/24 minutes. Yes. This corresponds to a fuel deficiency ratio of about 3-4%.

  In the present invention, considering the core stability, region stability, and channel stability, the ratio of the fuel deficient region is obtained by adopting the lower deficient fuel rod in the region of 0/24 to 6/24 below the fuel assembly. 3 to 8%. Furthermore, a defect region is also formed in the upper region of the fuel assembly (effective height 10/24 to 24/24) according to the number of lower defect fuel rods and the position of the defect. The upper limit value of the ratio of the total fuel deficient area to the upper deficient area is set to 8%.

  By adopting such a configuration, the core stability reduction ratio is reduced by 10% or more as compared with the 9 × 9 fuel core, and at the same time, a higher burnup than that of the 9 × 9 fuel assembly is achieved. Can do.

Further, the improvement rate of the core stability reduction ratio varies somewhat depending on the combination of the number and length of the defective fuel rods in which the fuel pellets are partially omitted, but the lower region of the assembly is reduced to 1/4 in 10 × 10 fuel. When the number of defects is reduced, if the number of defects is reduced to about 1/5 of 10 × 10, the same effective stability improvement effect as described above can be obtained. At this time, the effect of improving the stability is obtained by arranging the deficient fuel rods almost evenly in the cross section of the fuel assembly, for example, by arranging them on the diagonal of the channel box diagonal line perpendicular to the control rod center line. At the same time, the output distribution in the cross section can be flattened.
The above-described combination can also be applied to a fuel assembly having a larger number of lattices than 10 × 10 (for example, an 11 × 11 fuel assembly).

  In addition, the lower fuel deficient part can be used as a fuel internal pressure rise suppression space by connecting the cylinder part of the upper end opening lower end closed to the fuel rod, and the cylinder part with a small hole on the side wall is connected at the upper end closing lower end opening By doing so, it can also be used as a foreign matter removal space. The above means can be realized without any special modification to existing plant equipment and furnace structures.

Hereinafter, embodiments of a fuel assembly according to the present invention will be described with reference to the drawings.
(First embodiment)
The fuel assembly according to the first embodiment will be described with reference to FIGS. 1 to 4.
1A and 1B are schematic cross-sectional views of a 10 × 10 lattice fuel assembly 30 according to the first embodiment. FIG. 1A is a lower portion of the fuel assembly 30 and FIG. FIG. 2A and 2B are schematic longitudinal sectional views of the fuel assembly 30. FIG. 2A is a cross-sectional view (A cross section) of the outermost periphery of the fuel assembly 30, and FIG. B cross section).

  Referring to FIGS. 1A, 1B, and 2B, the center of the fuel assembly 30 has a circular cross section for one fuel rod in the lower region and nine fuel rods in the upper region. A water box 52 having a square cross section is arranged. Here, the total length of the fuel rod is about 4 m, the effective length (value for the full length fuel rod) is about 3.7 m, and the diameter is about 10 mm. The fuel rod is fixed to the lower tie plate 62 and the upper tie plate 61 by a lower end plug and an upper end plug having a screw structure, respectively. Further, the fuel rods 51 are held at predetermined intervals by seven fuel spacers 63. The diameter of the fuel pellet loaded in each fuel rod 51 is about 9 mm.

  The water box 52 is closed at the lower end and the upper end, and is supported and fixed to the lower tie plate and the upper tie plate. The coolant inflow hole 52a is several tens of centimeters downstream from the upper surface of the lower tie plate, and The outflow hole 52b is formed several tens of centimeters upstream from the lower surface of the upper tie plate (see FIGS. 2B and 3). The water box 52 configured as described above can substantially eliminate the increase in the absolute value of the void coefficient when the average enrichment of the fuel pellets loaded in the 10 × 10 lattice fuel assembly is increased.

  FIG. 3 is an enlarged schematic view of the upper tie plate 61 used in the present invention. The basic structure is such that round cells having an inner diameter smaller than the outer diameter of the fuel rod are connected in the same manner as the upper tie plate of the conventional 9 × 9 fuel assembly, and a fuel rod upper end plug is inserted into the cell. At this time, the lower deficient fuel rod 51a is threaded near the tip (about 5 cm), and is fixed to the upper tie plate 61 by the nut 61b after the fixed washer 61a is inserted from above the upper tie plate.

  The fuel assembly 30 is covered with a rectangular channel box 31 that forms a coolant passage in the vertical direction, and the coolant is inserted into the lower tie plate of the fuel rod lower end plug. It flows in from a plurality of small holes secured separately from the holes.

  In the present embodiment, among the fuel rods 51 of the fuel assembly 30, some of the plurality of fuel rods 51a (22 in total) are missing (in this embodiment, the missing position is from the lower end of the full length fuel rod. The effective length is about 6/24 and substantially corresponds to the lower end of the second spacer), and the ratio of the lower defect region is about 6%. The total pressure loss of the assembly is almost the same as that of the conventional 9 × 9 fuel assembly. Since the fuel assembly of the present embodiment is a 10 × 10 lattice, even when the lower deficient fuel rod 51a in which the fuel pellet does not exist in the lower portion described above is used, compared to the conventional 9 × 9 lattice fuel, U235, etc. Since the filling amount of the fissile nuclide is about 2%, it is possible to achieve high burnup.

  FIG. 4 is a diagram showing a core stability analysis result of the fuel assembly 30 configured as described above. In this analysis, the peaking coefficient 1.25 (8/24 node position) is applied to the core axis direction power distribution, compared to the peaking coefficient 1.15 (5/24 node position) applied in the conventional 9 × 9 fuel analysis. is doing. In the fuel assembly of this embodiment, since the boiling start point moves to the downstream side (fuel assembly outlet side), the passage time of bubbles (voids) generated in the assembly in the channel is shortened. Acts in the direction of improving stability.

  In this embodiment, even when 22 lower deficient fuel rods are employed, there are 69 fuel rods in the lower region, and the ratio of the single-phase flow pressure loss to the total pressure loss due to the increase in the single-phase flow region is 9 The stability of the core is improved because there is almost no change from the x9 fuel and the total pressure loss is almost the same as that of the conventional 9x9 fuel, and the core stability reduction ratio, which is an index of fundamental mode stability, is about Decrease by 15% (maximum value is 0.57). For the same reason, the region stability and the channel stability are also improved as compared with the conventional 9 × 9 fuel.

  In this embodiment, the number of lower deficient fuel rods is 22 and the lower deficit position is 6/24 of the effective length from the lower end. However, the present invention is not limited to this, and the ratio of the lower deficient region is about 3-8. %, On the premise of ensuring the same stability as in the above embodiment, the number of lower deficient fuel rods is appropriately changed within a range of about 20-30% of all fuel rods, and the deficit position is also It can be changed as appropriate.

  According to the first embodiment, by using a plurality of lower deficient fuel rods, the BWR fuel can be increased in burnup and the stability and reliability of the core can be greatly improved without changing the basic core structure. A fuel assembly of 10 × 10 or more can be provided.

(Second Embodiment)
A fuel assembly according to a second embodiment of the present invention will be described with reference to FIGS.
In the second embodiment, as shown in FIGS. 5 and 6, the cylindrical portion of the upper end opening lower end closing having the same outer diameter as the fuel cladding tube is formed in the deficient region of the lower deficient fuel rod 51 a described in the first embodiment. 51g is connected by the connection part 51c. As the material of the cylindrical portion 51g, for example, a Zircaloy alloy is used.

  The connecting portion 51c simultaneously closes the lower end of the lower deficient fuel rod 51a loaded with the mixed oxide fuel pellet 51n (or uranium oxide fuel pellet 51m) and the upper end of the cylindrical portion 51g, and has an internal communication hole 51s. (See FIG. 6), gas generated at the time of nuclear fission in the fuel rod portion flows into a cylindrical portion 51g whose lower end is closed by an end plug 51z. Even in this case, since the boiling start point moves to the downstream side, the passage time of bubbles (voids) generated in the aggregate in the channel is shortened, so that the stability of the core is improved. Furthermore, in this embodiment, since the gaseous fission product generated during the combustion period in the fuel rod loading portion moves to the space portion, an excessive increase in the internal pressure of the fuel rod can be prevented.

  Even when the above-mentioned number of lower-deficient fuel rods are used, the filling amount of fissile nuclides such as U235 is about 2% higher than that of the conventional 9 × 9 lattice fuel. Combined with the use of pellets, it is possible to achieve further higher burnup while avoiding a decrease in stability and an excessive increase in fuel rod internal pressure.

  According to the second embodiment, by using a plurality of lower deficient fuel rods and arranging the cylindrical portion in the deficient region, the high burnup of BWR fuel can be achieved without changing the basic core structure. It is possible to provide a fuel assembly of 10 × 10 or more that can greatly improve the stability and reliability of the reactor and the core, and can prevent an excessive increase in the internal pressure of the fuel rods.

(Third embodiment)
A fuel assembly according to a third embodiment of the present invention will be described with reference to FIGS.
In the third embodiment, the defective area of the fuel pellet is formed from the upper surface height of the lower tie plate to 1/24 effective length with respect to all 91 fuel rods, and the defective ratio is about 4%. Further, as shown in FIGS. 8 and 9, a cylindrical portion 51h having the same outer diameter as the fuel cladding tube and having an upper end closed lower end opening is connected by a connecting portion 51d.

  The connecting portion 51d completely closes the lower end of the fuel rod and the cylindrical portion 51h at the same time. The cylindrical portion 51h is provided with a large number of small holes 51p. The cylindrical portion 51h to which the fuel rod 51 is connected upward is vertically supported by the support hole of the lower tie plate with its lower end open. At this time, as shown in FIG. 10, a flange portion 51q having a cylindrical diameter of 13 mm is formed in an upper portion of about 3 cm from the lower end.

  Note that no holes other than those for supporting the fuel rods 51 are provided in the upper surface of the lower tie plate. Even in this case, since the boiling start point moves downstream, the passage time of bubbles (voids) generated in the assembly is shortened, which acts to improve the stability of the core.

  Further, in the present embodiment, the coolant flows into the fuel assembly through the cylindrical portion 51h having a small hole on the side surface, so that it is possible to prevent foreign matter from flowing into the fuel rod assembly. it can.

  In addition, even when the number of the above-mentioned number of lower fuel pellet-deficient fuel rods is adopted, the filling amount of fissile nuclides such as U235 is about 4% higher than that of the conventional 9 × 9 lattice fuel, which improves the stability. It is possible to achieve further higher burnup while avoiding the inflow of foreign matter.

  According to the third embodiment, the fuel pellet defect region is formed in the lower part of all the fuel rods, and the cylindrical portion is arranged in the defect region, so that the BWR can be changed without changing the basic core structure. It is possible to provide a fuel assembly of 10 × 10 or more that can significantly increase the burnup of the fuel and improve the stability and reliability of the core and can prevent an excessive increase in the internal pressure of the fuel rod.

(Fourth embodiment)
A fuel assembly according to a fourth embodiment of the present invention will be described with reference to FIGS.
In the fourth embodiment, in the 10 × 10 fuel assembly having one water box with a maximum cross section approximately circumscribing 3 × 3 fuel rods at the center, the lower end of the full length fuel rods (lower tie plate upper surface height) The natural pellets or deteriorated uranium pellets, which are non-heated fuel, are loaded in the fuel pellet deficient region having an effective length of 1/24. The defect ratio of the fuel pellet defect region is about 4%.

  And in the upper region, about 12 fuel rods loaded with pellets with enrichment over 5%, the upper deficit is in the region from about 18/24 to 24/24 of its effective length and up to the upper tie plate. A region is formed, and the total defect ratio is about 7%.

  The twelve missing fuel rods are provided with an upper plenum to suppress an excessive increase in internal pressure. Even in this case, the filling amount of fissionable nuclides such as U235 is almost the same as that of the conventional 9 × 9 lattice fuel. (In this embodiment, since the two-phase flow pressure loss is reduced, the region stability and the channel stability are greatly improved), and it is possible to achieve a high burnup.

  According to the fourth embodiment, natural uranium or deteriorated uranium pellets are arranged in the lower fuel pellet defect region of all the fuel rods, and the upper part of the plurality of fuel rods is used as the fuel pellet defect region. 10 × 10 or more, which can greatly increase the burnup of BWR fuel, improve the stability and reliability of the core, and prevent an excessive increase in the internal pressure of the fuel rods without changing the core structure. A fuel assembly can be provided.

(Fifth embodiment)
A fuel assembly according to a fifth embodiment of the present invention will be described with reference to FIGS. 13 to 14.
In the fifth embodiment, in a 10 × 10 fuel assembly in which two large-diameter water rods having a maximum cross section of 2 × 2 fuel are arranged in the central portion, from the lower end of the full length fuel rod (lower tie plate upper surface height). Natural uranium or depleted uranium pellets are loaded in the fuel pellet deficient area of 1/24 effective length. The defect rate in the lower region is about 4%.

  In the upper region, after loading pellets with enrichment exceeding 5%, the eight fuel rods provided with the upper plenum region have an effective length of about 20/24 to 24/24. In the area up to the tie plate, and for the six fuel rods labeled with S, the upper defective area is provided in the area from about 10/24 to about 24/24 of the effective length and up to the upper tie plate. Provided. The total defect rate is about 8%.

  Even in this case, compared with the conventional 9 × 9 lattice fuel, the filling amount of fissile nuclides such as U235 is about 1% more, so combined with the adoption of fuel pellets with enrichment over 5%, stability is improved. It is possible to achieve high burn-up after improvement (in this example, two-phase flow pressure loss is reduced, so that region stability and channel stability are greatly improved).

  According to the fifth embodiment, natural uranium or deteriorated uranium pellets are arranged in the lower defect region of all the fuel rods, and the upper portion of the plurality of fuel rods is used as the defect region, thereby providing a basic core structure. A fuel assembly of 10 × 10 or more that can significantly increase the burnup of the BWR fuel, improve the stability and reliability of the core, and prevent an excessive increase in the internal pressure of the fuel rod without changing the fuel Can be provided.

  Although the fuel assembly of the present invention has been described above with reference to the first to fifth embodiments, the scope of the present invention may be appropriately combined with the first to fifth embodiments described above, for example, fuel rods. When the defect region is from the lower end (lower tie plate upper surface height) to the 6/24 region, the upper end closed / lower end open tube with a small hole in the side wall is formed from the lower end to 3/24. Up to 24 is connected to the hollow tube with a connecting portion with a communication hole, and a fuel rod (loaded with uranium oxide fuel pellets with a enrichment of more than 5% or mixed oxide fuel pellets) is connected to the upper portion thereof, thereby connecting the fuel rod It is also possible to increase the margin of core stability after obtaining the effect of avoiding the increase in internal pressure and the effect of preventing the inflow of foreign matter.

  Even if the fuel rod defect region is replaced with a fuel rod loaded with depleted uranium or a zircaloy alloy rod, the core has the same structure as a conventional fuel rod, compared with the conventional 9 × 9 fuel. The stability margin is improved.

DESCRIPTION OF SYMBOLS 1 ... Reactor containment vessel, 2 ... Reactor pressure vessel, 3 ... Core, 4 ... Coolant, 30 ... Fuel assembly, 31 ... Channel box, 32 ... Control rod, 33 ... LPRM, 34 ... Core support plate, 35 ... upper lattice plate, 36 ... shroud, 37 ... fuel support fitting, 51 ... fuel rod, 51a ... lower deficient fuel rod, 51b ... upper deficient fuel rod, 51c ... connection portion, 51d ... connection portion, 51g ... cylindrical portion, 51h ... Cylindrical part, 51m ... Uranium oxide fuel pellet, 51n ... Mixed oxide fuel pellet, 51p ... Small hole, 51s ... Communication hole, 51q ... Hut part, 51x ... Fuel rod upper plenum, 51z ... Fuel rod lower end plug, 52 ... Water box, 52a ... Coolant inflow hole, 52b ... Coolant outflow hole, 61 ... Upper tie plate, 61a ... Fixed washer, 61b ... Nut, 62 ... Lower tie plate, 63 ... Spare .

Claims (14)

In a fuel assembly of a boiling water reactor having fuel rods arranged in an N × N square lattice and a water box arranged in the center,
A fuel assembly for a boiling water reactor, wherein a plurality of fuel rods among the fuel rods are lower deficient fuel rods.   The boiling water reactor according to claim 1, wherein the number of the lower deficient fuel rods is about 20 to 30% of the fuel rods, and the lower deficient region is 3 to 8% of the total fuel rods. Fuel assembly.   The fuel assembly for a boiling water reactor according to claim 1 or 2, wherein the deficient position of the lower deficient fuel rod is about 6/24 of the effective length from the lower end of the full length fuel rod.   The fuel assembly for a boiling water reactor according to any one of claims 1 to 3, wherein the lower deficient fuel rod is fixed only by an upper tie plate.   The fuel assembly for a boiling water reactor according to any one of claims 1 to 4, wherein the lower deficient fuel rod is supported by a lower tie plate through a cylindrical portion connected to a lower end portion thereof. .   6. The fuel assembly for a boiling water reactor according to claim 5, wherein the inside of the cylindrical portion is a gas plenum space.   6. The fuel assembly for a rising water reactor according to claim 5, wherein a plurality of small holes are provided in the cylindrical portion. In a fuel assembly of a boiling water reactor having fuel rods arranged in an N × N square lattice and a water box arranged in the center,
A fuel assembly for a boiling water reactor, wherein the fuel rod has a fuel pellet defect region formed in a lower region, and a cylindrical part having a large number of small holes provided in the defect region.   9. The fuel assembly for a boiling water reactor according to claim 8, wherein a ratio of the fuel pellet defect region is about 4% of all fuel rods. In a fuel assembly of a boiling water reactor having fuel rods arranged in an N × N square lattice and a water box arranged in the center,
The fuel rod comprises a plurality of full length fuel rods loaded with natural uranium or deteriorated uranium pellets in a lower fuel pellet defect region and a plurality of upper defect fuel rods formed with a fuel pellet defect region in the upper part. A fuel assembly for a boiling water reactor.   The fuel assembly for a boiling water reactor according to claim 10, wherein the upper and lower fuel pellet defect regions are about 8% or less of the total fuel rods.   The fuel assembly for a boiling water reactor according to claim 10 or 11, wherein two water rods having a large diameter are used instead of the water box.   13. The fuel rod according to claim 1, wherein the fuel rod is a fuel rod containing at least one of uranium oxide fuel pellets having a degree of enrichment exceeding 5% and mixed oxide fuel pellets. Boiling water reactor fuel assembly.   14. The fuel assembly for a boiling water reactor according to claim 1, wherein the fuel assembly satisfies N ≧ 10.

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