JPH0564756B2 - - Google Patents

Description

[Detailed Description of the Invention] [Field of Application of the Invention] The present invention relates to a light water-moderated nuclear reactor, and in particular to a light water-moderated nuclear reactor, in which the density of fuel rods is varied depending on the location in the reactor core to obtain a high degree of burnout, and the effective use of fuel is achieved. Regarding moderated nuclear reactors.

[Background of the invention]

The methods of using fuel materials in light water-moderated nuclear reactors (hereinafter referred to as light water reactors) are broadly classified into the once-through method and the reprocessing and recycling method. In the once-through reactor, enriched uranium is used in the reactor, and none of the components of the fuel material contained in the spent fuel rods removed from the reactor are reused (recycled) in the light water reactor. This approach is advantageous in terms of fuel cycle costs when the cost of fuel reprocessing exceeds the cost of uranium enrichment. On the other hand, in the reprocessing recycling method, the fuel material contained in spent fuel rods is reprocessed to create new fuel rods, and the fuel rods are loaded into a light water reactor and the fuel material is reused. .

One method for effectively utilizing fuel materials in the once-through method is to increase the burnup taken out from the fuel assembly, that is, to achieve a high burnup. A fuel assembly is composed of a large number of fuel rods. In order to achieve a high burnup, it is necessary to increase the degree of enrichment of enriched uranium, but as the degree of enrichment increases, the following problems arise. Due to the high enrichment of the new fuel assembly and the large extraction burnup of the fuel assembly, combustion assemblies with widely different infinite neutron multiplication factors coexist in the core of a light water reactor, and each fuel assembly shares the output power. A difference in proportion occurs, resulting in larger output mismatch and increased output peaking.
Additionally, as the enrichment increases, the surplus reactivity that must be controlled in the early stages of combustion increases, and in a conventional combustion assembly using combustion rods containing gadolinia, it is necessary to increase the number of combustion rods containing gadolinia.

On the other hand, from the perspective of effective use of uranium resources, a light water reactor has been proposed that improves the conversion of uranium-238 to fissile material (plutonium-239). “Schematic characteristics of an improved pressurized water reactor using improved combustion utilization” by Oldekop et al. in Nucl. Technol., 59 , 212 (1982).
“General features of advanced pressurized
water reactors with improved fuel
The document titled ``utilization'' describes a core in which the water-to-combustion volume ratio between the cores of a light water reactor has been lowered from the conventional 2.0 to 5.0, the average energy of neutrons has been increased, and the plutonium conversion rate has been increased to 0.9 or higher.
A dense lattice structure is used to achieve this water-to-fuel ratio of 0.5. In conventional high-conversion light water reactors, including this example, spent fuel assemblies taken out of the reactor core are reprocessed and recycled back into the reactor core in order to effectively utilize the fuel material. The fuel cycle including processing must be completed.

[Purpose of the invention]

An object of the present invention is to provide an efficient light water reactor in which the core is divided into a plurality of regions with different fuel rod densities, which flattens the linear power density distribution of the core and achieves uranium savings and high burnup. That's true.

[Summary of the invention]

The inventors have previously developed a reactor core that eliminates the drawbacks of the above-mentioned conventional technology and achieves uranium savings while achieving high burnup using a once-through method. Since this prior invention forms the basis of the present invention, its outline will be described below to aid in understanding the present invention.

This core is designed to have regions with varying fuel rod densities in the radial direction, and to shuffle fuel from regions with high fuel rod density to regions with low fuel rod density during its life, thereby making effective use of the fuel. Figure 7 shows the concept of the core configuration. 12 in the figure is the reactor core; 12
1, 122, 123, . . . , 12N are radial regions having different fuel rod densities. In the following, the hydrogen to uranium atomic ratio H/U will be taken up as a parameter corresponding to the fuel rod density. Generally, H/U becomes small in a region where the fuel rod density is high, and conversely becomes large in a region where the fuel rod density is low.

FIG. 8 shows changes in the neutron multiplication factor in each region shown in FIG. 7. Now, the hydrogen to uranium atomic ratio in the 121 region is a ₁ , similarly 1
In the 2N region, a _N is set such that a ₁ <a ₂ <a ₃ <...<a _N (1). Here, a ₁ is a smaller value than the hydrogen to uranium atomic ratio (approximately 5.0) in current light water reactors, and a _N is, conversely, a larger value. New fuel is initially loaded into the 121 region and burns to E ₁ . Here the next 12
It is transferred to area 2 and burns up to E ₂ . It is then sequentially transferred to other areas and taken out of the reactor at a burnup of E _N.

In this way, in the first half of the fuel's life, the fuel is placed in an area where the conversion ratio is high and the hydrogen to uranium atomic ratio is small, in order to accumulate plutonium, and in the second half, the fuel is placed in an area where the hydrogen to uranium atomic ratio is high. When placed in the area, it becomes possible to efficiently burn the accumulated fissile material. In other words, the new fuel is
If the fuel is placed in a region where the neutron spectrum is hard and the conversion ratio is high but the neutron multiplication factor is low, and the fuel that has been burnt is placed in the region where the neutron spectrum is soft and the conversion ratio is low but the neutron multiplication factor is high.
In addition to controlling surplus reactivity and reducing output mismatch, it is possible to save uranium while improving neutron economy.

FIG. 9 shows changes in the conversion ratio with respect to the burnup when the conversion ratio with respect to the burnup is used as a parameter when the atomic ratio of hydrogen to uranium is used as a parameter. When the hydrogen to uranium atomic ratio is lowered to about 1.0,
Even after burning 30 Gwd/t, approximately 95% of the fissile material of the new fuel remains, indicating a high degree of effective fuel utilization.

FIG. 10 shows changes in the neutron multiplication factor of fuel loaded in a core with two regions of fuel rod density in comparison with the case of fuel loaded in a conventional one region core with the same enrichment. In a two-zone core, as mentioned above, due to the effects of protonium accumulation in the early stage of combustion and spectral shift in the final stage of combustion, in principle, the extraction burnup increases by about 1.5 times under the same critical conditions (E _a →
E _b ), more than 30% uranium savings will be realized.

FIG. 11 shows an example of a core configuration when the fuel rod density is set to two regions. This example shows the case of a hexagonal fuel assembly. In the diagram, aggregate A is included.
Region 27 includes the region with high fuel rod density, assembly B.
The 28 region has a low fuel rod density.

The above is an outline of the prior invention.

Now, it is difficult with the prior art and the prior invention to flatten the power distribution (linear power density distribution of fuel) of a reactor core having such regions with different fuel rod densities. From the viewpoint of the fuel health of the reactor core, it is desirable to flatten the linear power density distribution rather than the power density distribution. Conventionally, flattening the power density distribution immediately led to flattening the linear power density, but this is not the case if there are regions with different fuel rod densities.
Figure 12 shows the power distribution of the two-zone core. Using technology such as conventional three-area loading (Patent No. 824747),
Even if the power density distribution is flattened as shown in the figure, if the fuel rod densities are different in the two regions of the core, the linear power density distribution will not be flattened. In the illustrated example, the fuel rod density in the 28 region is lower than in the 27 region, so even if the same power density is achieved, the linear power density in the 28 region increases discontinuously.

In order to flatten the linear power density distribution in such a two-zone core, it is necessary to change the power density distribution in stages as shown in FIG. Flattening of the linear power density distribution is achieved only after such a stepwise power density distribution is obtained.

Conventional power distribution flattening methods target a core with one region of fuel rod density, and no technology has been developed to realize stepwise flattening of power density distribution in multiple regions as described above. Therefore, even in the core of the prior invention, if flattening of the core linear power density distribution is not achieved and the distribution becomes as shown in Figure 12, the power density must be relatively reduced due to thermal limitations. As a result, the above-mentioned high burnup and uranium saving effects are lost.

Therefore, the present invention realizes a core configuration in which the radial linear power density distribution of the core is flattened in a core in which the fuel rod density differs in multiple radial regions, and an efficient light water reactor that achieves uranium savings and high burnup. It was made with the purpose of providing.

In the present invention, in a reactor core having a region in which the fuel rod density differs in the radial direction, the hydrogen to uranium atomic ratio is determined as (H/U) ₁ <(H/U) ₂ <...<(H/U) _i <... ...<(H/U) _o ...(2), then the average neutron infinite multiplication factor K _∞ of the partial region of the i region (i=2...n) that is in contact with the j region (j=i-1) is It is set lower than the average K _∞ of i-region and j-region. Alternatively, a partial region of the i region that is in contact with the j region is used as a moderator region that absorbs neutrons. Furthermore, in each region except for partial regions, K _∞ is increased on average toward the outer radial direction (however, in order to reduce neutron leakage at the outermost periphery of the core,
K _∞ may be lowered).

With such a core configuration, the step-type power density distribution described above is obtained, and the linear power density distribution of the core becomes flat as shown in FIG. this
The height of K _∞ is determined by the burnup of the fuel, so
By appropriately arranging fuel according to the burnup, the K _∞ distribution described above can be achieved.

[Embodiments of the invention]

Next, embodiments of the present invention will be described, but prior to that, the structure of a pressurized water reactor to which the present invention is applied will be described.

In FIG. 14, a pressurized water reactor 1 includes a reactor pressure vessel 2 containing a reactor core. The reactor vessel 2 has an inlet nozzle 3 and an outlet nozzle 4 on its wall, and has a sealing lid 5 installed on the top.
The substantially cylindrical core shroud 6 has a sealing lid 5
It is suspended from the canopy of reactor pressure vessel 2 nearby. A buttful 7 is attached to the lower interior of the core shroud 6. A lower core support plate 8 having holes for accommodating the lower ends of fuel assemblies A and B is attached to the lower end of the core shroud 6.
An upper core support plate 9 having holes for accommodating the upper ends of the fuel assemblies A and B is supported by a plurality of struts 10 on an upper support plate 11 having a thick beam. A plurality of guide tubes 26 are arranged between the upper support plate 11 and the upper core support plate 9. The guide tube 26 is provided with an opening 33.

The reactor core 12 is constructed in the lower part of the core shroud 6 and consists of a large number of combustion assemblies A and B and a cylindrical partition member 13. Partition member 13
is made of Zircaloy (zirconium alloy). In this embodiment, the core 12 includes the partition member 1
3, it is divided into two areas: a central area 27 on the inside and a peripheral area 28 on the outside. The fuel assembly A is located in a central region 27 within the partition member 13.
The fuel assemblies B are arranged in the peripheral area 28 outside the partition member 13, respectively.

As shown in FIGS. 15 and 16, the fuel assembly A includes a plurality of fuel rods 15 and a lower tie plate 1.
6, an upper tie plate 17 and a spacer 18. 21 is a handle. The lower tie plate 16 and the upper tie plate 17 are regular hexagons. Both ends of the fuel rod 15 are held by a lower tie plate 16 and an upper tie plate 17, respectively. The lower tie plate 16 has an inner cylindrical portion 16A, and the cylindrical portion 16A is connected to an outer cylindrical portion 16C by a plurality of connecting plates 16B arranged radially. The fuel rod 15 is the first
As shown in Figure 6, they are arranged in a regular hexagonal shape. Some of the fuel rods 15 function as tie rods 15A. Both ends of the tie rod 15A pass through the lower tie plate 16 and the upper tie plate 17. A nut 19 is attached to the lower end of the tie rod 15A.
A tightening nut 20 is attached to the upper end of A. The lower tie plate 16 and the upper tie plate 17 are connected by these tie rods 15A. Fuel rod 15 (including tie rod 15A)
This is a sealed zircaloy cladding tube filled with fuel pellets. The fuel pellets contain enriched uranium-235 as fissile material. In the axial direction of the bundle of fuel rods 15,
A plurality of spacers 18 are arranged. The spacer 18 prevents adjacent fuel rods 15 from coming into contact with each other and ensures a passage for cooling water to flow between the fuel rods 15.

As shown in FIGS. 17 and 18, the fuel assembly B includes a plurality of fuel rods 15 and burnable poison rods 15.
B, lower tie plate 23, upper tie plate 2
4 and a spacer 25. Fuel assembly B
Similarly to the fuel assembly A, the lower tie plate 23 and the upper tie plate 24 are connected to each other by tie rods 15A which are a part of the plurality of fuel rods 15.
are connected. The lower tie plate 23 has a cylindrical portion 23A inside.
A is connected to the outer cylindrical portion 23C by a plurality of connecting plates 23B arranged radially. fuel rod 15
Both ends of the burnable poison rod 15B are held by a lower tie plate 23 and an upper tie plate 24. Fuel rod 15 (including tie rod 15A)
has the same configuration as that of fuel assembly A. The burnable poison rod 15B is a sealed cladding tube filled with a mixture of zirconium hydride, which is a moderator, and gadolinia, which is a burnable poison. The concentration distribution of zirconium hydride and gadolinia in the burnable poison rod 15B is uniform in its axial direction. The spacer 25 has a regular hexagonal shape, and bundles the fuel rods held on each tie plate into a regular hexagonal shape. A plurality of spacers 25 are arranged in the axial direction to prevent contact between the fuel rods. Instead of the burnable poison rod 15B, a UO ₂ pellet of the fuel rod 15 mixed with gadolinia may be placed in the fuel assembly B.

The distance (pitch) between adjacent fuel rods in fuel assembly B is greater than the pitch in fuel assembly A. That is, the number of fuel rods constituting fuel assembly B is smaller than the number of fuel rods in fuel assembly A. Lower tie plate 1
6 and 23 and upper tie plates 17 and 24 have the same horizontal cross-sectional area (cross-sectional area perpendicular to the axis of the fuel assembly).

Cylindrical portion 16A of fuel assembly A and fuel assembly B
The cylindrical portion 23A is inserted into the aforementioned hole of the lower core support plate 8. Then, the cylindrical portion 16C of the fuel assembly A and the cylindrical portion 23C of the fuel assembly B are installed on the lower core support plate 8. The upper part 17A of the upper tie plate 17 of the fuel assembly A and the upper part 24A of the upper tie plate 24 of the fuel assembly B are inserted into the aforementioned holes of the upper core support plate 9. In this way, fuel assemblies A and B are held by the lower core support plate 8 and the upper core support plate 9. In the central region 27, adjacent fuel assemblies A
lower tie plate 16 and upper tie plate 1 of
7 are in contact with each other. The lower tie plate 17 of the fuel assembly A, which is located at the outermost periphery of the central region 27, is in contact with the inner surface of the partition member 13. Also in the peripheral region 28, the lower tie plate 23 and the upper tie plate 24 of adjacent fuel assemblies B are in contact with each other. In the fuel assembly B located innermost in the peripheral region 28, the lower tie plate 23 and the upper tie plate 24 are in contact with the outer surface of the partition member 13.

The pressurized water reactor includes a control rod drive mechanism 29 that operates with fluid pressure. Control rod drive mechanism 29
The structure is the same as that shown in US Pat. No. 3,607,629. Control rod drive mechanism 29
is attached to the upper flange 32 of the adapter tube 31. The adapter tube 31 passes through the hermetic lid 5 and is attached to the hermetic lid 5 by welding. A control rod 34 shown in FIG. 19 is detachably attached to the lower end of the control rod drive mechanism 29. control rod 34
moves up and down within the guide tube 26. The control rod drive mechanism 29 inserts or withdraws the control rods 34 into or from the fuel assemblies A and B. In addition, 35 is a main body, 36 is a support member, and 37 is a neutron absorption rod.

Now, in one embodiment of the present invention shown in FIG.
This is an example of a core in which the fuel rod density is varied in two regions, with the fuel rod density in the inner 27 regions being higher than in the outer 28 regions. Here, a hexagonal lattice type fuel assembly is used. FIG. 2 shows an example in which a square lattice type fuel assembly is used in a similar two-zone core. In the diagram,
The average K _∞ of the fuel assemblies L in the annular region among the 28 regions adjacent to the 27 region is lower than the average K _∞ of the fuel assemblies L in the 27 region and 28 region. In addition, the average K _{∞ of the outermost fuel assembly (marked with a circle) is lower than the average K ∞ of} the 27 and 28 regions in order to reduce neutron leakage. Figure 3 shows the K _∞ distribution and power density in the radial direction of this two-zone core.
Shows the linear power density distribution. In the figure, the region indicated by L corresponds to the fuel assembly with low K _∞ shown in FIGS. 1 and 2. The K _∞ of the innermost annular partial region of the 28 region that touches the 27 region is lower than the other regions, and in other regions, except for the outermost periphery, the K _∞ becomes higher on average as you move toward the outside in the radial direction of the furnace. Arrange the fuel assembly accordingly. As a result, a stepped power density distribution as shown in FIG. 3 is realized, and even if there are regions with different fuel rod densities, the linear power density distribution can be flattened.
The specific fuel arrangement may be made so as to achieve the target K _∞ distribution according to the burnup of the fuel. For example, in the example shown in FIG. 3, the fuel assembly in which combustion has progressed most among the 28 regions may be placed in the region indicated by L and on the outermost periphery.

The linear power density distribution can also be flattened in a multi-region core using a similar method. FIG. 4 shows a core divided into three regions in the radial direction, with the density of fuel rods decreasing from the inside to the outside. In the figure, the average K _∞ of the innermost fuel assembly L in the 28 area that touches the 27 area and the innermost fuel assembly L in the 28A area that touches the 28 area.
are the average of the combustion aggregates in 27, 28, and 28 regions, respectively.
It is lower than K _∞ . Figure 5 shows the K _∞ distribution, power density, and linear power density distribution in the radial direction of this three-zone core. As in the case of FIG. 3, a step-type power density distribution is realized, and even if there are many regions with different fuel rod densities, the linear power density distribution in the core can be flattened.

Basically, by increasing the K _∞ of the outermost fuel in the region with high fuel rod density and lowering the K _∞ of the innermost fuel in the region with low adjacent fuel rod density, the neutron flux at the boundary is reduced. It is possible to make the change in the power density very large so that the power density changes nearly discretely. In principle, this area where fuel with a low K _∞ is loaded can be replaced with a moderator, a structural material, or the like. However, in that case, the economic advantage of being able to burn fuel with a low K _∞ to a higher burnup will be lost.

FIG. 6 shows an example in which the present invention is applied to a so-called island type multi-region core. Here, the fuel rod density in seven island regions 27 provided in the core is higher than in the other regions 28. For such a core, by lowering the average K _∞ of the fuel assemblies L surrounding the island region compared to other regions, a stepped power density as shown in Figures 3 and 5 can be achieved. The distribution is realized locally, and the linear power density distribution of the fuel in the core can be flattened.

〔Effect of the invention〕

According to the present invention, even if the reactor core is divided into multiple regions and the fuel rod densities differ in each region, it is possible to flatten the linear power density distribution of the core, as in the case of a conventional one-region core. It becomes possible. As a result, high power density can be obtained under certain thermal constraints, resulting in a high burnup and uranium-saving reactor.

In the case of a two-zone core, the extraction burnup will be approximately 1.5 times higher, achieving a uranium savings of more than 30%.

[Brief explanation of the drawing]

Figure 1 is a horizontal cross-sectional view of a hexagonal lattice type two-zone reactor core that employs the linear power density distribution flattening method according to the present invention, Figure 2 is a horizontal cross-sectional view of a square lattice type two-zone core, and Figure 3 is a diagram showing the neutron infinite multiplication factor, power density, and linear power density distribution in the radial direction of the two-zone core of the present invention, FIG. 4 is a cross-sectional view in the horizontal direction of the core showing the three-zone core of the present invention, and FIG. Figure 6 is a diagram showing the neutron infinite multiplication factor, power density, and linear power density distribution in the radial direction of the three-region core of the embodiment; Figure 6 is a horizontal sectional view showing the island-type core of the present invention; Figure 7 is a diagram showing the island type core of the present invention; A horizontal cross-sectional view of the reactor core showing the concept of the target core configuration, Figure 8 is a diagram showing the change in the infinite neutron multiplication factor with respect to fuel burnup, and Figure 9 is a diagram showing the burnup with the hydrogen to uranium atomic ratio as a parameter. Figure 10 is a diagram comparing changes in the neutron infinite multiplication factor of a two-zone core and a one-zone core. Figure 11 is a horizontal cross-sectional view of the core showing a specific example of a two-zone core. Figures 12 and 13 are diagrams showing the case where the power density is flattened and the case where the linear power density is flattened for the radial power density distribution of the two-zone core and the linear power density distribution of the fuel, respectively, and Figure 14 1 is a longitudinal sectional view of a pressurized water reactor having the core shown in FIG. 1, which is a preferred embodiment of the present invention, FIG. 15 is a side view of a fuel assembly loaded in the central area of FIG. 1, and FIG. is a cross-sectional view taken along line XX in Fig. 15, and Fig. 17
The figure is a side view of a fuel assembly loaded in the peripheral area of FIG. 1, FIG. 18 is a YY sectional view of FIG. 17, and FIG. 19 is a perspective view of a control rod. 1... Pressurized water reactor, 2... Reactor vessel, 3
...Inlet nozzle, 4...Outlet nozzle, 5...Sealing lid, 6...Core shroud, 7...Bathful, 8
...Lower core support plate, 9...Upper core support plate, 1
0... Support column, 11... Upper support plate, 12... Core, 13... Partition member, 15... Fuel rod, 16
...Lower tie plate, 17...Upper tie plate, 18...Spacer, 19...Nut, 20...
...Tightening nut, 23...Lower tie plate, 24
... Upper tie plate, 25 ... Spacer, 26
... Guide tube, 27 ... Central region, 28 ... Peripheral region, 29 ... Control rod drive mechanism, 31 ... Adapter tube, 32 ... Upper flange, 33 ... Opening, 34
...control rod.

Claims (1)

[Claims] 1. The reactor core is divided into a plurality of regions, and the hydrogen to uranium atomic ratio (H/U) _i of each region is (H/U) ₁ <(H/U) ₂ <...<( In different reactors such as H/U) _i <...<(H/U) _o , the j-th (j=i-
1) A nuclear reactor characterized in that the average infinite neutron multiplication factor of the fuel in the partial region that is in contact with the region is lower than the average neutron infinite multiplication factor of the fuel in the i-th region and the j-th region. 2. Claim 1 is characterized in that each region of the core is an annular region in the radial direction, and the infinite neutron multiplication factor of the fuel is increased toward the outer side in the radial direction in each region except for the partial regions. Reactor. 3. A nuclear reactor according to claim 2, characterized in that the infinite neutron multiplication factor of the fuel in the outermost region of the core is lower than the average infinite neutron multiplication factor of each region. 4. A nuclear reactor according to any one of the above claims, characterized in that a partial region of the i-th region in contact with the j-th region is replaced by a moderator or a structural material, or a moderator and a structural material. 5 Divide the core into multiple island regions and surrounding regions, and set the hydrogen to uranium atomic ratio (H/U) _i of each island region to (H/U) ₁ ≦ (H/U) ₂ ≦ ……≦(H/U) _i ≦……≦(H/U) _o In a nuclear reactor, the average neutron infinite multiplication factor of the fuel in a partial region in contact with each island region in the peripheral region is expressed as the island region and A nuclear reactor characterized by having a lower average neutron infinite multiplication factor of the fuel in the surrounding area. 6. The nuclear reactor according to claim 5, wherein the partial region is replaced by a moderator or a structural material, or a moderator and a structural material.