JPH0378955B2 - - Google Patents

Description

[Detailed Description of the Invention] [Industrial Application Field] The present invention relates to a fuel assembly, and in particular to a fuel assembly disposed in the core of a boiling water reactor, which is used to reduce core reactivity gain and pressure loss. The present invention relates to a fuel assembly with suitable water rods.

[Prior Art] A core of a boiling water nuclear reactor includes fuel assemblies arranged at a certain distance from each other and control rods inserted between the fuel assemblies.

FIG. 3 shows the structure of a fuel assembly used in a boiling water reactor, and FIG. 4 shows the structure of a fuel rod used in the fuel assembly. The fuel assembly 10 is
It has an upper tie plate, a lower tie plate 14, and a fuel rod 16. The lower end of the channel box 12 attached to the upper tie plate surrounds the lower tie plate 14. A plurality of holes are bored in the lower tie plate 14, and lower end plugs 18 of the fuel rods 16 are inserted into the holes. The upper end plug 20 of the fuel rod 16 is retained in the upper tie plate. The fuel rod 16 is pressed downward by an expansion spring 22 attached to the upper end plug 20, and its lateral movement is restricted by a spacer 26 at the intermediate portion. A handle 28 is also provided on the upper tie plate.

The fuel rod 16 has a cylindrical fuel cladding tube 30 filled with fuel pellets 32 . This fuel pellet 32 is pressed downward by a spring 34 within the fuel cladding tube 30.

FIG. 5 shows a cross section of the fuel assembly 10. The fuel rods 16 and water rods 17 are arranged in a grid of 8 rows and 8 columns.

Water, which is a coolant, flows within the fuel assembly 10 .
This water removes thermal energy generated by the reaction of fissile material, such as uranium-235, present within the fuel rods 16. Water also flows into the gaps between the fuel assemblies outside the channel box.

On the other hand, water also acts as a moderator for neutrons, and the high-energy fast neutrons produced by nuclear fission are moderated by the moderator, water, and become low-energy thermal neutrons. These thermal neutrons are released into the fissile material in the fuel rods 16, such as uranium-235.
When absorbed, a nuclear fission reaction occurs, producing energy.

However, in a boiling water reactor, voids are generated inside the channel box 12 due to heat generation from the fuel rods 16, whereas no voids are generated outside the channel box 12. Therefore,
The density distribution of water is characterized by being high outside the channel box 12 and low near the center of the fuel assembly. In addition, while the fuel absorbs thermal neutron flux inside the channel box 12, the nuclear fuel does not absorb thermal neutrons outside the channel box 12, and water, which is a moderator, is abundant, so the thermal neutron flux The distribution of is low in the center of the fuel assembly and high on the outside near the channel box. This causes an imbalance in the power distribution within the fuel assembly.

As a countermeasure to this problem, methods such as increasing the number of water rods installed in the center of the fuel assembly were proposed in 1983-
It is described in Publication No. 40986. In recent years, there has been a trend toward increasing the burnup of fuel in order to improve the economic efficiency of nuclear power plants. In order to increase the amount of water in the center of the fuel assembly in response to the increase in enrichment, it is being considered to increase the diameter of the water rod and place it in the center of the fuel assembly. .

[Problem to be solved by the invention]

However, increasing the diameter of the water rod
This results in a narrowing of the coolant flow path area of the fuel assembly. This results in the problem of increased pressure loss and decreased thermal margin. It became clear that countermeasures were needed to address this problem.

An object of the present invention is to provide a fuel assembly that can suppress an increase in pressure loss and increase reactivity.

[Means to solve the problem]

The features of the present invention that achieve the above object include a plurality of fuel rods filled with nuclear fuel material inside, and a fuel assembly that is arranged in the center of a fuel assembly and occupies an area where a plurality of fuel rods can be arranged. and a water rod having a diameter larger than the diameter of the fuel rod, the axial length of the water rod being shorter than that of the fuel rod, and the water rod in the fuel rod adjacent to the water rod. The area above the upper end of the water rod is lower than the planar enrichment in that area of the outermost fuel rod.

[Effect]

By making the axial length of the water rod shorter than that of the fuel rod, increase in pressure loss can be suppressed.
In the region above the upper end of the water rod, the average enrichment of the outermost fuel rods is greater than that of the fuel rods adjacent to the water rod, so that the reactivity can be increased.

[Embodiments of the invention]

Before explaining a fuel assembly which is a preferred embodiment of the present invention, the content of the study that led to the present invention will be explained. The water rod used in the embodiment of the present invention is arranged to occupy a space in which a plurality of fuel rods are arranged, and is characterized by having a larger diameter than the fuel rods, and is called a large-diameter water rod. The cross-sectional area of the large-diameter water rod is 3 to 5 times larger than the cross-sectional area of the fuel rod. This large diameter water rod is placed in the center of the fuel assembly. The length of the large diameter water rod is shorter than the fuel rod, and the height of the upper end of the thicker diameter part of the fuel rod is the part of the fuel rod that is filled with nuclear fuel material (referred to as the effective fuel part). , its axial length is called the effective fuel length).
Such a large diameter water rod is called a partial length water rod.

Placing a large-diameter water rod or a container for storing a moderator such as water in the center of the fuel assembly is disclosed in Japanese Patent Application Laid-Open Nos. 50-40986, 1972-13981 and 1972. -Described in Publication No. 65792, etc. It is known that the arrangement of the water rods homogenizes the thermal neutron flux distribution within the fuel assembly, which has the effect of simplifying the enrichment distribution and improving fuel economy by increasing reactivity.

However, placing a large-diameter water rod in the center of the fuel assembly reduces the coolant flow area, which worsens the cooling characteristics of the fuel rod and reduces the thermal margin of the fuel. Become.

The cooling characteristics of fuel rods are characterized by the pressure drop of the coolant flowing within the fuel assembly. Increased pressure drop deteriorates fuel rod cooling characteristics,
Conversely, reducing pressure drop is known to improve the cooling characteristics of fuel rods.

The flow in the boiling water reactor core is a gas-liquid two-phase flow, and its frictional pressure loss ΔP _f is expressed as ΔP _f =W ² /2 _g 〓·f·L/D _H ·A ² _CH φ _TPF .

Here, ΔP _f is friction pressure drop, W is channel flow rate, g is gravitational acceleration, ρ is water density, D _H is channel hydraulic diameter, A _CH is channel flow area, L
is the length, f is the friction pressure loss coefficient, and φ _TPF is the two-phase flow friction pressure loss multiplier.

Remove the four fuel rods in the center of the fuel assembly,
If a large-diameter water rod is placed instead, the channel flow area A _CH decreases, so the friction pressure loss ΔP _f
increases. In order to prevent an increase in the frictional pressure loss ΔP _f , it is conceivable to shorten the length L of the pressure loss portion.
In this embodiment, in order to shorten the length L, the length of the large diameter water rod is shortened and made into a partial length. The length of the large-diameter water rod is determined by the length of the large-diameter water rod because the original purpose of the large-diameter water rod is to improve the nuclear properties in the fuel effective part, that is, to flatten the thermal neutron flux distribution of the fuel assembly. It is sufficient to reach the upper limit of .

However, if the pressure loss due to the large diameter water rod is large, the upper end of the large diameter water rod must be located below the upper end of the fuel effective portion in order to shorten the length L. Therefore, at the upper end of the fuel effective portion where there is no large-diameter water rod, the thermal neutron flux distribution is greatly distorted and reactivity is lost.

In order to improve the deterioration of nuclear properties caused by the adoption of part-length water rods, there is an idea to use low-enrichment fuel, such as natural uranium, in the entire region located at the upper end of the part-length water rod. However, when the region is made of natural uranium, a reactivity gain occurs when the axial length of the region is within about 45 cm, or within 3/24 of the effective length of the fuel. If the part-length water rod needs to be made shorter than this, it is no longer appropriate to use natural uranium above the upper end of the part-length water rod.

In this embodiment, as a countermeasure against this problem, the following measures were taken in the axial division of the enrichment degree.

FIG. 6 shows the relationship between the axial enrichment classification and the partial length water rod for two types of fuel rods.

The part-length water rod 103 has an axial length shorter than the effective fuel length, and its upper end is shorter than the upper end of the fuel rod. Assuming a boundary surface that divides the area at the upper end of the part-length water rod, the area below this boundary surface is defined as a first area, and the area above the boundary surface is defined as a second area.
The fuel rods are classified at the above boundary surface into fuel rods 102 that have enrichment area divisions and fuel rods 101 that do not have enrichment area divisions. The fuel rod 102 is
the degree of enrichment in the first region is lower than the degree of enrichment in the second region;
It is arranged around the large diameter water rod 103 with a partial length.
In FIG. 6, e ₁ , e ₂ , and e ₃ indicate the degree of enrichment of each region, and there is a relationship of e ₁ >e ₃ and e ₂ >e ₃ . The relationship between e ₁ and e ₂ is not particularly restricted, but generally e ₁ ≦e ₂ .

These enrichment region divisions operate as follows. FIG. 7 schematically shows a cross section of the first area in FIG. 6, showing the water gap part outside the channel box,
The water rod part and the part in between where the fuel rods are arranged are shown divided into two parts. The purpose of using large-diameter water rods is to provide a water region in the center of the fuel assembly to flatten the distribution of thermal neutron flux, so the enrichment of the fuel rods is approximately equal on the inside and outside. can do. In relation to FIG. 6, the fuel rods arranged inside are fuel rods 102.
The fuel rods arranged on the outside are fuel rods 101. When the area of the water area in the large-diameter water rod is approximately equal to the area of the water area in the water gap, it is also possible to set e ₁ =e ₂ . However, since the area of the water region in the large-diameter water rod is small, the fuel rod output is generally flattened by setting e ₁ ≦e ₂ . The reason why the inner side is a highly enriched region is to effectively absorb the neutrons heated in the large-diameter water rod region into the fuel.

On the other hand, FIG. 8 shows a cross section of the second region where large-diameter water does not exist. In the second region, the center becomes a void region, so placing the high enrichment region on the outside and the low enrichment region on the inside will result in placing the high enrichment fuel rods near the water gap. , the reactivity can be improved. In addition, when natural uranium is used as the low enrichment region, a large amount of uranium-238 will be placed in contact with the high void region, which will increase the conversion to plutonium, and by further burning it, reaction will occur. The degree can be increased.

In addition, by configuring the second region into a high enrichment region and a low enrichment region, the average enrichment of the second region is lower than the average enrichment of the first region, and the reactor has a margin for reactor shutdown. It has the effect of improving

In addition, Japanese Patent Application No. 115269/1984 describes a technology to flatten the axial power distribution by dividing the axial enrichment of a fuel rod into two regions, and this technology can be used in combination with the above-mentioned technology. can do. in this case,
For example, fuel rod 104 in FIG. 9 will be used as fuel rod 102. By providing the enrichment dividing surface A of the fuel rod 104 below the upper end of the large diameter water rod 103, the same effect as in the case of FIG. 6 can be obtained.

Furthermore, Japanese Patent Application No. 58-46077 describes a technique for arranging natural uranium over a length of 15 to 30 cm at the upper or lower ends of fuel rods in order to improve fuel economy. It is also possible to use this technique in combination.

Hereinafter, a fuel assembly which is an embodiment of the present invention will be explained with reference to FIGS. 1, 2, and 10.

FIG. 10 shows the structure of a fuel assembly having part-length water rods in this embodiment. The large diameter water rod 8 has a shorter axial length than the fuel rod 16. Furthermore, as shown in FIG. 1, the large diameter water rod 8 is disposed in the center of the fuel assembly in an area that can be occupied by four fuel rods, and its outer diameter is equal to that of the fuel rod 16.
is larger than that of

FIG. 2 shows the enrichment distribution and gadolinia (Gd ₂ O ₃ ) distribution of the fuel rods 16. Fuel rods 1-7
and the water rod 8 are arranged as shown in FIG.
In FIG. 2, the shaded area represents the natural uranium area, and the other areas are the enriched uranium area. The enrichment level of the enriched uranium region is shown numerically in the figure. For example, the enrichment level of most of the fuel rod 1 is 3.0.
%. Fuel rods 6 and 7 contain gadolinia. Fuel rod 6 contains 5% gadolinia, fuel rod 7 contains 3.5% gadolinia in the upper region and 4.5% gadolinia in the lower region. The total number of fuel rods 1 to 7 is 60
In a book, these are 8x8 surrounding one water rod 8.
are arranged in a grid pattern.

In this embodiment, the upper end of the water rod 8 is located below the upper end of the fuel effective portion by 5/24 of the total length of the fuel effective portion. The fuel rods 5 and 6 surrounding the water rod 8 have an enrichment classification boundary at a position 5/24 of the total length of the effective fuel part below the upper end of the effective fuel part, and enrichment in the area above the boundary. degree is 0.71%
(natural uranium), and the area below the boundary (excluding the 1/24th area from the bottom of the fuel effective part) is
4.5% and 3.9%. Fuel rods 5 and 6 are fuel rod 1
Corresponds to 02. Fuel rods 1 to 4 correspond to fuel rod 101 because they are arranged not adjacent to water rod 8 and have no enrichment division at a height of 5/24. In addition, the fuel rod 3 has an enrichment division point at a height of 9/24 from the lower end of the effective fuel, and the axial power distribution is flattened by the technology shown in Japanese Patent Application No. 115269/1983. There is. In addition, a technique (described in Japanese Patent Application No. 1983-46077) has also been adopted to improve the reactivity by filling 1/24 of the upper and lower ends of each fuel rod with natural uranium.

In this embodiment, the pressure loss of the fuel assembly can be reduced by making the large-diameter water rod 8 a partial length, and the upper ends of the fuel rods 5 and 6 are made of natural uranium, as shown in FIG. With this configuration, reactivity gain can be obtained. Furthermore, the reactor shutdown margin is also improved.

Furthermore, the present invention is not limited to an 8 x 8 lattice as in this embodiment, but may be implemented using, for example, a 9 x 9 lattice arrangement of fuel rods as shown in FIGS. 11 and 12. Application is also possible in fuel assemblies. In particular, in a fuel assembly arranged in a 9 x 9 arrangement, the increase in pressure loss increases due to the increase in the number of fuel rods. It becomes more necessary to use the partial length as the partial length. In the example of FIG. 11, as shown in FIG. 12, two partially long and large diameter water rods are arranged at the center of the assembly. In both fuel rods 5 and 6 adjacent to this, the region above the upper end of the large diameter water rod of partial length is made of natural uranium.

〔Effect of the invention〕

According to the present invention, it is possible to suppress an increase in pressure loss of a fuel assembly, and it is also possible to increase reactivity.

[Brief explanation of drawings]

FIG. 1 is a cross-sectional view of a fuel assembly that is a preferred embodiment of the present invention, FIG. 2 is an explanatory diagram showing the enrichment and gadolinia distribution of each fuel rod shown in FIG. 1, and FIG. 4 is a structural diagram of the fuel rod shown in FIG. 3, FIG. 5 is a cross-sectional view of the fuel assembly shown in FIG. 3, and FIG. 6 is shown in FIG. 1. An explanatory diagram showing the concept of the embodiment, FIG. 7 is a schematic diagram showing the cross-sectional configuration of the first region shown in FIG. 6, and FIG.
FIG. 9 is an explanatory diagram showing the relationship between a fuel rod and a partial length water rod to which the concept shown in Japanese Patent Application No. 115269/1980 is applied; Figure 10 is a longitudinal sectional view of the fuel assembly in Figure 1;
FIG. 11 is a longitudinal cross-sectional view of a fuel assembly according to another embodiment of the present invention, and FIG. 12 is a cross-sectional view of the fuel assembly shown in FIG. 11.

Claims (1)

[Scope of Claims] 1. A fuel assembly having a plurality of fuel rods filled with nuclear fuel material, and a fuel assembly arranged in the center of the fuel assembly, occupying an area where the plurality of fuel rods can be arranged, and a water rod having a diameter larger than the diameter of the fuel rod, the axial length of the water rod is shorter than that of the fuel rod, and the axial length of the water rod is shorter than the upper end of the water rod in the fuel rod adjacent to the water rod. A fuel assembly characterized in that the upper region has a lower average enrichment than the average enrichment in that region of the outermost fuel rods. 2. The fuel assembly according to claim 1, wherein the fuel rod adjacent to the water rod has a region above the upper end of the water rod made of natural uranium.