JPH0376875B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION [Field of Application of the Invention] The present invention relates to a fuel assembly and a reactor core for a nuclear reactor.

[Background of the invention]

In conventional nuclear reactor fuel assemblies, the fuel enrichment in the low neutron flux region (hereinafter referred to as the low neutron region) is generally higher than the fuel enrichment in the high neutron flux region (hereinafter referred to as the high neutron region). Fuel assemblies with increased height and flattened local power are used.
However, such fuel assemblies place a large amount of fissile material in the low neutron region, which hinders the effective combustion of nuclear fuel.

Therefore, for example, in the fuel assembly disclosed in Japanese Unexamined Patent Publication No. 58-26292, it is possible to flatten the power distribution by creating a fuel concentration difference in the axial direction within the possible range of power distribution within the fuel assembly. The structure is such that the fissile material is moved to the high neutron region in the radial direction of the fuel assembly.

However, in this fuel assembly, the concentration of nuclear fuel material in the axial direction is differentiated, increasing the concentration of nuclear fuel material in the upper half region, and reducing the concentration of nuclear fuel material in the lower region where thermal neutron flux is high within the allowable range of the power distribution. As the fuel was lowered, there was a loss of burnup in the lower part in the axial direction.

Therefore, it was desired not only to shift the fuel enrichment to the high neutron region in the radial direction of the fuel assembly, but also to increase the fuel enrichment in the axial center, which is the higher neutron region, in the axial direction of the fuel assembly. . However, with such a configuration, there is a problem of increasing local output, so a method of increasing the number of fuel assemblies in the core was considered, but increasing the number of fuel assemblies would mean increasing the size of the core. There was a serious drawback.

As a result, there is a strong need for fuel assemblies that can place higher fuel enrichments in the high neutron region without increasing the local maximum power of the fuel rods.

[Purpose of the invention]

The present invention was made in view of the above circumstances, and an object of the present invention is to provide a fuel assembly and a reactor core that can effectively burn nuclear fuel and increase burnup.

[Summary of the invention]

A fuel assembly for a nuclear reactor according to a first aspect of the present invention is a fuel assembly for a nuclear reactor constructed by bundling a large number of long fuel rods, and a region of high thermal neutron flux in a radial cross section of the fuel assembly. The weight fraction of fissile material in the fuel in the fuel rod belonging to is lower in the axially central part and higher in the axially upper and lower parts than in the axially central part,
The weight fraction of fissile material in the fuel in the fuel rods in the low thermal neutron flux region of the radial cross section of the nuclear fuel assembly is high in the axial center and lower in the axial upper and lower parts than in the axial center. The fuel assembly is characterized in that the radially average weight ratio of fissile material in the axially central portion is less than the average value in the axially upper and lower portions, The reactor core has a reactor fuel assembly composed of a large number of long fuel rods bundled together.
Among the fuel assemblies, at least the fuel assembly at the outermost periphery of the core belongs to a region of high thermal neutron flux in the radial cross section of the fuel assembly. is lower, higher in the upper and lower parts of the axis than in the middle part,
The weight fraction of fissile material in the fuel in the fuel rods in the low thermal neutron flux region of the radial cross section of the fuel assembly is high in the axial center, and is lower in the axial upper and lower parts than in the axial center. The average fissile material weight ratio in the radial direction of the fuel assembly is determined by the axial center portion and the axial upper and lower portions, where the average value at the axial center portion is less than or equal to the values at the axially upper and lower portions. It is constructed by arranging a plurality of types of fuel assemblies having different fissile material weight percentages of fuel, and the average fissile material weight percentage of the fuel assemblies at the outermost periphery of the core is:
It is characterized by a lower average content of fissile material than the average content of fissile material in the fuel assemblies on the inner side of the outermost periphery of the core.

The present invention was made based on the following study results.

In other words, in conventional fuel assemblies for nuclear reactors, attention has been focused only on the portion that produces the maximum output peak, and studies have been made to reduce the output of that portion. , focusing on the lower part, taking into consideration that increasing the local power peaking coefficient in the axial upper and lower parts will not cause any problems, and placing high-concentration fuel in the radial periphery of the fuel assembly where the neutron flux is high in the upper and lower parts. However, since the upper and lower axial regions are originally regions with low neutron flux,
By setting the average concentration at the top and bottom to be less than the center in the axial direction, the average concentration is increased in the upper and lower parts and lowered in the center. In addition to flattening the directional power distribution, it also enables effective use of fuel, achieving the intended purpose.

[Embodiments of the invention]

Examples will be described below.

FIG. 1 is an explanatory diagram of the configuration of a concentric cluster fuel assembly for a pressure tube nuclear reactor, which is an embodiment of the present invention, where a is a vertical cross section, and b and c are X-X cross sections of a, respectively. 1 is a fuel rod arranged on the outer periphery, 2 is a fuel rod arranged on the inner side, and one fuel assembly consists of fuel rods 1 and 2.
It consists of 36 rods in total, and a fuel support rod 3 containing a coolant is placed in the center.

A fuel rod is composed of fuel pellets, a cladding tube covering the pellets, and end plugs for sealing the upper and lower ends, and plutonium-uranium mixed oxide is used as the fuel.

The fuel rod 1 is filled with fuel pellets P in the upper part 1a and the lower part 1c, and the fuel pellets Q in the middle part 1b.The fuel rod 2 is filled with fuel pellets Q in the upper part 2a and the lower part 2c, and the middle part 2b is fuel pellet P
is filled. The enrichment of fuel pellets P and Q is 3.2 wt% and 1.6 wt%, respectively.

That is, in the pressure tube reactor fuel assembly of this embodiment, the plutonium enrichment of the fuel rod 1 in the radially outermost layer of the assembly with high thermal neutron flux is 1.6 wt%, whereas the plutonium enrichment in the axial center portion 1b is 1.6 wt%. , axial upper and lower parts 1a,
In 1c, the plutonium enrichment is 3.2wt%, which is higher than in the central part 1b. On the other hand, in the fuel rods 2 in the middle and inner layers, where the thermal neutron flux is low, the plutonium enrichment in the axial center part 2b is 3.2wt%, while the plutonium enrichment in the axially upper and lower parts 2a and 2c is 3.2wt%. It is set at 1.6wt%, which is a lower value than that of the axial center portion 2b. As a result, in this example, the relationship between the plutonium enrichment of the 18 fuel rods in the outer layer and the 18 fuel rods in the middle and inner layers is reversed between the axially upper and lower parts and the axially central part. ing.

In addition, in conventional fuel assemblies for pressure tube reactors, fuel rods with a constant plutonium enrichment in the axial direction are used, and the outer layer fuel rods have a plutonium enrichment of 3.2 wt%. The plutonium enrichment is 1.6wt% for the fuel rods in the middle and inner layers, which reduces the maximum output of the fuel rods near the axial center of the fuel. However, the average extraction burnup is approximately
It was 30000MWd/t.

In this way, the average extraction burnup of conventional fuel assemblies with the same average plutonium enrichment is approximately
30,000 MWd/t, whereas the average burnup of the fuel assembly in this example was 31,500 MWd/t.
It is possible to measure an increase of 1500 MWd/t. The axial output peaking coefficient has also been reduced by approximately 6%,
The maximum linear power density is reduced by 6% and the health of the fuel is further improved.

In addition, this example can be implemented by simply changing the arrangement of the fuel pellets in the outer layer, inner layer, and middle layer, without changing the type or amount of plutonium enrichment in the conventional fuel. It is a simple construction that does not require the use of granular fuel pellets.

Fig. 2 shows a comparison of the output distribution of the fuel assembly of the embodiment shown in Fig. 1 and the conventional fuel assembly, where a indicates the embodiment (the present invention) and b indicates the conventional fuel assembly. In this case, the vertical axis shows the axial position, and the horizontal axis shows the average plutonium enrichment (wt%), axial output peaking coefficient, local output peaking coefficient, and axial X local output peaking coefficient. , the axial average plutonium enrichment distribution, the axial power distribution, the intra-assembly local power peaking coefficient distribution, and the axial X local power peaking coefficient distribution. As is clear from FIG. 2, the axial power peaking coefficient was approximately 1.5 in the case of the conventional fuel assembly, even when the axial average plutonium enrichment was 2.4 wt%, but in the present invention In the case of , it was approximately 1.4, which enabled a reduction of approximately 6%. This is because the local output peaking coefficient is increased by replacing the same number of outer layer fuel with the middle layer and outer layer fuel in the upper and lower parts of the axis in the axial direction where there is no problem in terms of power distribution. The output peaking coefficient was also approximately 1.8 for conventional fuel assemblies, but
In the present invention, it is approximately 1.68, which is a reduction of approximately 6%. In addition, at the upper and lower fuel boundaries in the axial direction,
The output is slightly higher than before, but it is smaller than the axially central portion, so there is no problem.

3 and 4 are explanatory diagrams of a core of a pressure tube reactor which is an embodiment of the core of the present invention, in which the fuel assembly shown in FIG. 1 is used, and the fuel assembly 120
It is made up of the body. FIG. 3a shows the fuel arrangement in the radial direction of the core, and FIG. 3b is a radial power distribution diagram of the Z-Z cross section of FIG. 3a, with the fuel output (relative value) plotted on the vertical axis. A, B, and C indicate different types of fuel assemblies used in the outer circumferential portion, intermediate portion, and inner circumferential portion, respectively. Figures 4a, b, and c show the axial configurations of fuel assemblies A, B, and C, respectively, and Figures 4d and e show the XX cross sections of fuel assemblies A, B, and C, and Y- A Y cross section is shown.

The reactor core of this example uses uranium-plutonium mixed oxide fuel as the fuel material,
Highly enriched fuel P, which is natural uranium enriched with 3.2wt% fissile plutonium, and natural uranium enriched with 1.6wt% fissile plutonium.
Low enrichment fuel Q enriched with fissile plutonium
Two types of fuel rods are used. fuel assembly A,
As shown in FIG. 4, B and C have different fuel enrichment arrangements within the assembly.

In the fuel assembly A, the low enrichment fuel Q is used in all the axial directions of the inner layer fuel and the middle layer fuel, and the highly enriched fuel P is used in all the axial directions of the outer layer fuel. In the fuel assembly B, a range of approximately 1/3 of the total length of the inner layer fuel and the middle width layer fuel in the axial direction is the highly enriched fuel P, and the other upper part,
The lower part is the low enrichment fuel Q. Further, the outer layer fuel has low enrichment fuel Q in a range of about 1/3 of the total length in the axial center, and high enrichment fuel P in the other upper and lower parts.

In the fuel assembly C, approximately 1/2 of the total length of the inner layer fuel and the middle layer fuel in the axial direction is made into highly enriched fuel P.
The other upper and lower parts are low enriched fuel Q. Moreover, the outer layer fuel has a low enrichment fuel Q in about 1/2 of the axial center area, and a high enrichment fuel P in the other area.

A nuclear reactor has a general property that the thermal neutron flux is lower at the radial periphery or at the axial upper and lower parts than at the center of the core. Also, pressure tube reactors,
Alternatively, a light water reactor has a characteristic that the thermal neutron flux is lower at the center than at the radial periphery of the fuel assembly. The fuel assembly C in FIG. 3 is located in the central region of the reactor core and has a large thermal neutron flux overall, but the thermal neutron flux decreases at the upper and lower parts in the axial direction and at the center in the radial direction of the fuel assembly. Therefore, only the outer layer fuel in the fuel assembly, which is effective for flattening the axial power distribution, has approximately 1/4 of the upper and lower axial regions as highly enriched fuel P, and the central region as low enriched fuel Q. . On the other hand, the inner layer and the middle layer are made of highly enriched fuel P in order to flatten the local output within the assembly in the axial center where the thermal neutron flux is high, and the axis where there is no need to flatten the local output and the effect of axial flattening is small. Direction top, bottom about 1/4
In the range of , low enrichment fuel Q is considered.

Fuel assembly B is located outside fuel assembly C in FIG. 3, and is loaded at a position where the average thermal neutron flux is lower than the center of the core. Therefore, if the fuel assembly C is used at this position, the output will be lower than that at the center. Therefore, a fuel assembly B is used in which the position where the fuel arrangement in the axial direction changes is moved closer to the center than in the fuel assembly C.

In addition, since the fuel assembly A in Figure 3 is placed in the periphery of the reactor where the thermal neutron flux is lowest in the radial direction of the core, the outer layer fuel with the highest thermal neutron flux in the fuel assembly is used as the highly enriched fuel P. The plan is to increase the output and flatten the power distribution in the radial direction of the core, and to create a low-enriched fuel Q that is wasteful because not much fuel is burned in the inner and middle layers, where thermal neutron flux is the lowest in the core.

In the core of this embodiment, it is possible to flatten the power distribution in the radial direction of the core, as shown in FIG. 3b. In addition, in the core of this example, the fuel plutonium enrichment in each axial cross section is all the same.
Although it is 2.4wt%, it is possible to flatten the axial power distribution by about 5%. When the core of this example is adopted, the average fuel extraction burnup will be approximately
30,000MWd/t was 32,500MWd/
t, resulting in an improvement in burnup of approximately 2500 MWd/t.

The reactor cores of the embodiments shown in Figures 3 and 4 use two types of plutonium-enriched fuel, but some or all of the inner layer fuel with low thermal neutron flux may contain natural uranium or plutonium that is not needed from the enrichment plant. The same result can be obtained even if low-fissile oxide fuel such as depleted uranium produced as a material or depleted uranium produced by reprocessing spent fuel is used, and plutonium-enriched fuel is used as the middle and outer layer fuel. can form a reactor core.

Further, by arranging fuel in the core using more fuel assemblies with three or more types of fuel concentrations, it is possible to form a core with a more optimal burnup power distribution.

Further, the cores of the embodiments shown in FIGS. 3 and 4 can be formed from enriched uranium oxide fuel, and fuel mixed with flammable neutron poisons such as gadolinia can also be used.

Furthermore, in the embodiments shown in Figures 3 and 4, the fuel concentration in the inner layer, intermediate layer, and outer layer of the fuel assembly is the same for each layer, but it is also possible to vary the concentration for each fuel rod within the layer. It is.

FIG. 5 shows another embodiment of the fuel assembly of the present invention, and is an explanatory diagram of the configuration of the fuel assembly that can also be used in the cores of FIGS. 3 and 4, where a is a vertical cross section;
b, c, d and e are I-I and J- of a, respectively
J, K-K, and L-L cross sections are shown, and 4 shows the fuel rods arranged in the outer layer, 5 shows the fuel rods arranged in the middle layer, and 6 shows the fuel rods arranged in the inner layer. .

In the I-I cross section, since these are the uppermost and lowermost ends in the axial direction, the low concentration fuel Q is used for all of the fuel rods 4, 5, and 6 in the outer layer, middle layer, and inner layer in order to avoid wasteful combustion efficiency. In the J-J cross section, the outer layer fuel rod 4 is used as a high concentration fuel P because of the relatively low thermal neutron flux region in the axial direction, and the inner layer and intermediate layer fuel rods 4 and 5 are used as a low concentration fuel Q. Thermal neutron flux is relatively high in the K-K cross section, but slightly lower than in the central part in the axial direction, so the fuel rods 6 and 4 in the inner and outer layers are filled with low-concentration fuel Q, and the fuel rods 5 in the middle layer are filled with high-concentration fuel P. , the local output peaking coefficient is made slightly larger than that of the LL cross section.
In addition, since the thermal neutron flux is highest at the L-L cross section,
The fuel rods 6 and 5 in the inner and intermediate layers are made of high concentration fuel P, and the fuel rods in the outer layer 4 are made of low concentration fuel Q to flatten the local power distribution.

The fuel rods in the embodiments shown in Figs. 1 to 5 are vertically symmetrical with respect to the central part in the axial direction, but in the fuel section for a boiling water-cooled nuclear reactor where voids occur in the coolant, the fuel rods in the axial direction above the central part are vertically symmetrical. It is also possible to move the boundary surface upwards to make it thermally easier.

FIG. 6 is an explanatory diagram of a fuel assembly for a light water reactor according to yet another embodiment, in which FIG. 6a shows the fuel arrangement in the radial direction of the core, and A, C, and D indicate fuel rods in areas with high thermal neutron flux; B, E, F, G, and H indicate fuel rods in the region of low thermal neutron flux, and fuel rods without symbols indicate water rods. Figure 6b shows fuel assembly A~
In the fuel assembly disclosed in JP-A-58-26292, which shows the axial configuration of H and the overall average configuration of the fuel assembly in the axial direction, the fuel rods in the high neutron region The fuel concentration in the upper and lower parts is higher than that in the center, and the fuel concentration in the upper and lower parts of the low neutron region fuel rod is lower than that in the axial center.
Figure c shows fuel assemblies A' to H' (A
The axial configuration of the fuel assembly (used corresponding to H), the total average configuration of the fuel assembly in the axial direction, and the thermal neutron flux distribution are determined. The numbers shown in Fig. 6b and c are the fuel concentrations (wt%), and the average of the high neutron regions A, C, and D in this example is 3.45wt%,
The average of low neutron regions B, E, F, G, and H is 2.7wt
%, and the average of high neutron regions A′, C′, D′ is
3.3wt%, and the average in the low neutron regions B', E', F', G', and H' is 2.85wt%. In this example, the fuel rods A, C, and D in the high thermal neutron flux region are Lower the center density and make the upper and lower density higher than the center. Among the fuel rods in the low thermal neutron flux region, B, E, and H, other than the fuel rods containing gadolinia, have a high concentration at the center in the axial direction, and a lower concentration at the upper and lower portions than at the center.

In other words, the conventional fuel assembly described in JP-A-58-26292 flattens the axial power distribution by increasing the fuel concentration in the upper part of the axis than in the lower part, and reduces the thermal neutron flux in the radial direction of the assembly. This is an excellent product that increases the fuel usage efficiency by increasing the fuel concentration in the outer periphery than in the central part, but in order to increase the fuel usage efficiency, which is the focus of the present invention, axial flattening is applied to the aggregate. It should be taken into consideration that flattening should be carried out in the region with the highest thermal neutron flux in the radial direction, and conversely in the fuel rods in the region with the lowest thermal neutron flux, the concentration should be lowered above and below the center in the axial direction without carrying out flattening in the axial direction. It had not been done.

Compared to such a conventional fuel assembly, the fuel assembly according to the embodiment of the present invention shown in FIG. 6 can improve burnup by 5% while maintaining the same concentration on average. As shown in Figure 6, this increases the average concentration of fuel rods in the hot neutron region from 3.3wt% to 3.45wt% by approximately 5%.
This is based on the fact that the fuel concentration has been increased from the center to the bottom in the axial direction where thermal neutrons are high in the axial concentration distribution compared to the conventional one.

As is clear from the description of the embodiments above, the axial power distribution can be flattened without increasing the average fuel concentration in the axial direction in the upper and lower parts of the reactor core. In addition, in the upper and lower parts of the axial direction, the fuel concentration is increased in the radial periphery of the fuel assembly where thermal neutron flux is high, and the fuel concentration is lowered in the radial center than in the axial center. Since the axially central portion can have a concentration higher than the average concentration of the axially upper and lower portions, a high burnup can be achieved with a low fuel concentration.

If two or more types of fuel enrichment are used within a fuel assembly, if the reactor has a neutron flux distribution within the fuel assembly, the new type of fuel enrichment can be used without adding a new type of fuel enrichment. It is possible to flatten the axial power distribution and improve burnup simply by changing the fuel enrichment arrangement.

〔Effect of the invention〕

The present invention makes it possible to provide a fuel assembly that can effectively burn nuclear fuel and increase burnup, and has great industrial effects.

[Brief explanation of drawings]

FIG. 1 is an explanatory diagram of the configuration of an embodiment of the fuel assembly of the present invention, FIG. 2 is an explanatory diagram of the output distribution of the embodiment of FIG. 1 and the conventional fuel assembly, and FIGS. The figure is an explanatory diagram of one embodiment of the reactor core of the present invention, FIG. 5 is an explanatory diagram of the configuration of another embodiment of the fuel assembly of the present invention,
FIG. 6 is an explanatory diagram showing the structure of another embodiment in comparison with the structure of a conventional fuel assembly. 1...Fuel rod (arranged on the outer periphery), 2...
Fuel rods (located on the inside), 1a, 1b, 1
c... (of fuel rod 1) upper part, center part, lower part, 2
a, 2b, 2c...upper, center, lower part (of fuel rod 2), P...highly enriched fuel, Q...lowly enriched fuel.

Claims (1)

[Scope of Claims] 1. In a fuel assembly for a nuclear reactor constructed by bundling a large number of long fuel rods, the fuel rods within the fuel rods that belong to a region of high thermal neutron flux in the radial cross section of the fuel assembly The weight fraction of fissile material in the fuel is lower in the axial center, and higher in the upper and lower axial regions than in the axial center, and is higher in the fuel rods in the low thermal neutron flux regions of the radial cross section of the fuel assembly. The weight percentage of fissile material in the fuel is higher in the axial center, and lower in the upper and lower axial regions than in the axial center.
The fuel assembly is characterized in that the radially average weight ratio of fissile material in the fuel assembly is such that the average value at the axial center portion is less than the average value at the axially upper and lower portions. 2. The region of high thermal neutron flux in the radial cross section of the fuel assembly is the region consisting of the fuel rods at the outermost periphery of the fuel assembly, and the region of low thermal neutron flux is the region inside the outermost periphery of the fuel assembly. The fuel assembly according to claim 1, which is a region consisting of fuel rods. 3 The weight proportion of the fissile material used in the fuel assembly as a whole is the weight proportion of the fissile material used in the fuel rods in the radial cross section of the central part in the axial direction; 3. The fuel assembly according to claim 1, wherein the number of types of material weight proportions is the number of types of fissile material weight proportions used within the radial cross section of the axially central portion. 4. In a core having a fuel assembly for a nuclear reactor configured by bundling a large number of long fuel rods, a radial cross section of the fuel assembly excluding at least the fuel assembly at the outermost part of the core among the fuel assemblies. The weight fraction of fissile material in the fuel in the fuel rod, which belongs to the region of high thermal neutron flux, is lower in the axial center, higher in the axial upper and lower parts than in the axial center, and is higher in the radial direction of the nuclear fuel assembly. The weight percentage of fissile material in the fuel in the fuel rod in the low thermal neutron flux region of the cross section is high in the axial center, and lower in the axial upper and lower parts than in the axial center. The radial average fissile material weight ratio is the fissile material weight of the fuel in the axial center, upper and lower axial areas where the average value at the axial center is less than the average value at the axial upper and lower parts. It is constructed by arranging a plurality of types of fuel assemblies with different proportions, and the average fissile material weight proportion of the fuel assemblies at the outermost part of the core is
A reactor core characterized in that the ratio of fissile material is lower than the average content of fissile material in fuel assemblies on the inner side of the core than at the outermost periphery of the core. 5. The ratio of the fissile material weight percentage of the fuel in the axially central portion of the outermost fuel rod of each of the plurality of types of fuel assemblies to the fissile material weight percentage of the fuel in the axially upper and lower portions is 5. The reactor core according to claim 4, wherein the core becomes smaller from the center of the core toward the outer periphery. 6. The region of high thermal neutron flux in the radial cross section of the fuel assembly is the region consisting of fuel rods at the outermost periphery of the fuel assembly, and similarly the region of low thermal neutron flux is the region inside the outermost periphery. 6. The reactor core according to claim 4 or 5, which is a region consisting of fuel rods. 7 The weight proportion of the fissile material used in the fuel assembly as a whole is the weight proportion of the fissile material used in the fuel rods in the radial cross section of the central part in the axial direction; The reactor core according to claim 4, 5, or 6, wherein the number of types of material weight percentages is the number of types of fissile material weight percentages used within the radial cross section of the axially central portion. .