JPS6356514B2 - - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION The present invention relates to a boiling water nuclear reactor installed in a commercial nuclear power plant, and particularly to the arrangement structure of fissile material constituting the reactor core.

Boiling water reactors are operated by controlling the excess reactivity of nuclear fuel, which is fissile material, in the reactor core by inserting and withdrawing control rods, and the neutron absorbing materials such as the control rods are completely removed. Maximum burnup is achieved in the drawn state.

The fuel is then replaced when the reactor has no excess reactivity. That is, old fuel assemblies in which combustion has progressed are removed and new fuel assemblies are loaded, while the old fuel assemblies in which combustion has not progressed are replaced.

In a conventional boiling water reactor (hereinafter abbreviated as BWR), fuel replacement is performed after removing approximately 25% of the fuel assembly with high burnup and advanced combustion, and then replacing it with a new fuel assembly (hereinafter referred to as R). ) are uniformly distributed and loaded as shown in Figure 1 (this is
This is called the uniform distributed loading method. ) On the other hand, when exchanging fuel assembly positions within the core, only the movement near the original fuel assembly position is performed.

In addition, in a conventional BWR, fissile material (nuclear fuel) with an infinite multiplication coefficient k∞ is uniformly distributed in the first core (initial core) as shown in Figure 2.
As shown in Figure 3, a fuel assembly with an enrichment of 1.1wt/o is loaded in the □× position, and a fuel assembly with an enrichment of 2.5wt/o is loaded in the □ position.
Some fuel assemblies are arranged.

Therefore, although the conventional BWR has a simple fuel assembly arrangement and loading method, has a small output peaking factor, and is superior in flattening, the infinite multiplication coefficient k of the fuel in the outermost region of the core is
Since ∞ is the same as the core average (Figure 2) or higher than the core average (Figure 3), there is a lot of neutron leakage from the core, and therefore the obtained burnup is small. (The operating cycle length is short.) Therefore,
This was not necessarily the best way to use fuel effectively. In addition, in conventional nuclear reactors, only the arrangement of fuel assemblies in the initial core was considered, and no consideration was given to the core after the second operation cycle (referred to as a replacement core).

For this reason, in conventional nuclear reactors, it is necessary to keep the power peaking factor small not only in the initial core but also in the replacement core, to level out the reactor output, and to use fuel effectively. The problem was how to configure the layout.

In consideration of the above points, the present invention aims to flatten the reactor power distribution while increasing the acquired burnup, that is, the length of the operating cycle, and making it possible to fully and reliably utilize the fuel effectively. The purpose is to provide

To achieve this objective, the present invention aims to prevent neutron leakage from the BWR core to the outermost region (when viewed from above, the region where the outermost fuel assembly is located). The second position is not included).

EMBODIMENT OF THE INVENTION Hereinafter, one embodiment of the nuclear reactor according to the present invention will be described with reference to the accompanying drawings.

The present invention can be applied to the fuel arrangement in the initial core (core in the first operation cycle) and replacement core (core in the second and subsequent operation cycles) of a nuclear reactor. Of these, FIG. 4 shows an example in which the method is applied to an initial core, and FIG. 5 shows an example in which the method is applied to a replacement core.

The reactor shown in Figure 4 shows 1/4 of the core,
This core section 10 is covered with a reflector (not shown), and the core section 10 has a central region C (a central region occupying about 1/3 of the core radius) and an intermediate region D.
It is divided into three areas: a central area and an area excluding the outermost peripheral area, and an outermost peripheral area B. Outermost area B
A fuel assembly of fissile material with a low enrichment level (e.g. natural uranium with the smallest infinite multiplication coefficient k∞), which is indicated by an ×□ mark, is placed in and a part of the central region C. The volume occupied by the fissile material is set to be approximately 1/4 of the volume of the entire area. This is the case when the core is refueled in 25% increments. Therefore, if the outermost region B occupies 15% of the volume of the entire core, then 10% of the total region (25% minus 15%, that is, the number of fuel assemblies in the outermost region is calculated from the total number of fuel assemblies replaced). fissile material with a low enrichment degree is placed in the central region C of the volume of the remaining volume (remaining after subtracting the amount of The fuel assemblies are arranged like a checkerboard as shown in FIG. In addition, in a part inside the intermediate region D, fuel assemblies containing fissile material with a medium enrichment and fuel assemblies containing fissile material with a high enrichment indicated by □ are alternately arranged. In the outer part only fuel assemblies with highly enriched fissile material are arranged.

By distributing fuel assemblies containing fissile material with low, medium, and high enrichment levels in the reactor core 10 shown in FIG. While the fissile material is distributed so that the multiplication factor increases sequentially, the fuel assembly having the fissile material with the smallest infinite multiplication factor is arranged in the outermost region B.

Therefore, the overall furnace output is increased, and the output peaking factor can be kept small.

This will be explained in detail as follows.

The infinite bridge coefficient k∞ of the fissile material at the radial position of the reactor core 10 has a relationship as shown by the solid line in FIG. From FIG. 6, it can be seen that the infinite multiplication coefficient k∞ of the center C and the outermost region B of the core 10 is smaller than the infinite multiplication coefficient of the uniform distributed loading system shown by the broken line as in the conventional BWR. For this reason, as shown in Figure 4, 92 fuel assemblies with an enrichment level of 0.7% (natural uranium) are placed in the outermost region B and the central region C, and the central region of the inner region (core region excluding the outermost region). 76 medium concentration fuel assemblies of 2.1% in C and middle area D, and 2.6 in the remaining inner area A.
When 200 fuel assemblies with a high enrichment level of 200% are incorporated (368 fuel assemblies in total), the reactor power of the initial core becomes as shown by the solid line in Figure 7, which is similar to that of a conventional BWR. This increases the reactor output by about 20% compared to the case where fuel assemblies with an enrichment level of about 2% are uniformly distributed and loaded (indicated by the broken line). In other words, the burnup in the first operation cycle of a conventional BWR is approximately
8000MWD/T, whereas the present invention
The BWR burnup will be approximately 9500MWD/T. This is because neutron leakage largely depends on the infinite multiplication coefficient of the outermost region B. In other words, as will be explained later, since the size of the cross section of the fuel assembly is large compared to the neutron diffusion distance, the amount of neutron leakage from the entire core is almost determined by k∞ of the fuel assembly in the outermost region. . This is the reason why the fuel assembly in the outermost region has the lowest infinite multiplication coefficient k∞, which is the gist of the present invention.

Moreover, in the case of this embodiment, the power peaking factor in the radial direction of the reactor core 10 is smaller than that of the conventional BWR.
This decreases from 1.35 to approximately 1.32, and the output distribution is flattened (averaged).

In this embodiment, all the fuel assemblies taken out after the end of the first operation cycle are fuels with low enrichment, which is advantageous in terms of fuel cost as will be described later, and the fuel can be used effectively.

Further, FIG. 5 shows a replacement core that has been replaced after completing one operation cycle (operation time from one fuel exchange to the next refueling). This replacement core will replace approximately 25% of the total fuel assemblies with new fuel assemblies R.
It was exchanged with As can be seen from FIG. 5, the new fuel assembly R is not loaded in the outermost area B, but is loaded in the inner area A as shown in the figure. Meanwhile, the old fuel assemblies left in the reactor core 10 are replaced. Among these, the fuel assemblies with the highest burnup (the lowest infinite multiplication coefficient k∞) are arranged in the outermost region B and the center region C indicated by ×□ marks. Next, among the remaining fuel assemblies, a fuel assembly with a high burnup (low infinite multiplication coefficient) is placed in the central region C,
Hereinafter, the fuel assemblies with higher burnup are arranged in order from the center of the center region C toward the intermediate region D, and the outer periphery of the intermediate region D is where combustion has not progressed much (k
high ∞) to distribute the fuel assemblies. The old fuel assembly is taken out along with the loading of the new fuel assembly, which is the fuel with the highest burnup at the end of the previous cycle, as in the past.

As a result, even in the replacement core, the outermost region B
The fuel substances having the lowest infinite multiplication coefficient k∞ and whose infinite multiplication coefficient k∞ gradually increases outward from the central region C are distributed as shown by the solid line in FIG. As a result, the reactor power of the replacement core is expressed as shown by the solid line in Figure 9 (conventional
The infinite multiplication coefficient k∞ and the reactor power of the BWR replacement core are shown by the broken lines in Figs. 8 and 9. ), approximately 110% more reactor power can be obtained compared to conventional replacement cores.

Note that FIG. 10 shows the relationship between the infinite multiplication coefficient k∞ of the fissile material and burnup, and FIG. 11 shows the relationship between the infinite multiplication coefficient k∞ and enrichment. 1st
0 and 11, it is shown that the infinite multiplication coefficient k∞ decreases as combustion progresses, while increasing as the enrichment level increases.

Therefore, it is understood that a fuel assembly with advanced combustion and a fuel assembly with low enrichment are equivalent in that both contain a small amount of fissile material and have a low infinite multiplication coefficient. Therefore, the present invention is applicable not only to the initial core but also to the replacement core.

The reason for limiting fuel loading will be explained in detail below.

(a) Minimizing neutron leakage In light water-moderated thermal neutron reactors such as BWR, the diffusion distance of combustion neutrons is 2 to 3 cm, and even fast neutrons are 7 to 8 cm.

On the other hand, the cross-sectional size of a BWR fuel assembly is approximately 15 x 15 cm, which is larger than the neutron diffusion distance mentioned above. For this reason, neutrons generated in one fuel assembly will have a large effect on only the adjacent fuel assemblies, and
It almost never reaches fuel assemblies that are one or more rows apart. Most of the neutrons that leak out from the core and are lost are generated at the outermost periphery of the core. Neutrons generated in one or more rows of fuel assemblies inside the core rarely leak out of the core.

On the other hand, in a fuel assembly with a small infinite multiplication coefficient, that is, with a small amount of fissile material, the number of neutrons generated is small, so the number of neutrons that go out from this fuel assembly is also small. By arranging such a fuel assembly at the outermost periphery of the core, the effect of reducing neutron leakage from the core can be sufficiently achieved.

That is, the present invention was made based on the above-mentioned physical characteristics of neutrons unique to BWR.

On the other hand, for example, in Japanese Patent Publication No. 50-715,
An example is disclosed in which the core is divided into three equal volumes, that is, the "outer periphery" accounts for 33% of the total core. In this case, the outer periphery naturally has a width of 2 to 3 rows, so compared to the present invention, it is an unnecessary part (i.e., an unnecessary part for the effect of reducing neutron leakage from the core).
Up to this point, fuel assemblies with low infinite multiplication coefficients are arranged. This is not only wasteful, but also inevitably causes drawbacks such as increased output peaking and decreased extraction burn-up, as will be described later.

(b) Flattening of output distribution By configuring the present invention as described above,
It is possible to minimize neutron leakage from the reactor core, maximize the burnup obtained during the operating cycle, and even out the power distribution. On the other hand, in the aforementioned Japanese Patent Publication No. 50-715, fuel with high k∞ is concentrated in the middle second region, which accounts for 33% of the total, so the output peaking is high and the output distribution is flat. It has the disadvantage that it is difficult to convert into a carrier. In addition, according to Japanese Patent Publication No. 50-715, since the outer periphery with low reactivity occupies 33% of the total volume, fuel with high reactivity inevitably remains in the center of the core, worsening power peaking in the center. This is the result.

In addition, in the present invention, by arranging fuel assemblies of two types of enrichment in combination at the center or intermediate portion of the core, it is possible to increase the infinite multiplication factor sequentially from the center of the core. (Tokuko Showa 50
-Not mentioned in Publication No. 715) (c) Maximizing the extraction burnup In a normal boiling water reactor, approximately 1/4 to 1/3 of the fuel is extracted at the end of each operation cycle, so the reaction It is clear that fuel with a low degree of multiplication (infinite multiplication coefficient) needs to be taken out from the center in addition to the outermost 10 to 15%.

Therefore, by placing fuel with low reactivity at the outermost periphery and also at the center, in addition to increasing the operational burnup due to the decrease in reactivity at the outermost periphery, it is possible to increase the reactivity at the center of the core. By arranging a low fuel, the burnup of the fuel in the center of the core that is taken out after the end of the operation cycle is increased, and the fuel economy is improved due to the effect of increasing the takeout burnup.

On the other hand, in the example of Special Publication No. 50-715,
Because the fuel with the lowest reactivity is placed in the outer periphery, which occupies 1/3 of the entire core, the output in this area is low.
Combustion does not proceed either. Therefore, in the current BWR, which extracts 1/4 to 1 to 3 of the fuel at the end of the operating cycle, the fuel is extracted from the portion where combustion has not progressed very much, and the extracted burnup cannot be significantly improved. do not have.

As described above, in the nuclear reactor according to the present invention, fissile material is distributed so that the infinite multiplication factor increases sequentially from the center of the core in the inner region toward the outside (excluding the outermost region), and the fissile material is distributed in the outermost region. Because the fissile material with the smallest infinite multiplication factor is placed in the area, the absolute amount of neutron leakage is small not only in the initial core but also in the replacement core, and the same amount of fissile material is larger and more flat than before. It is possible to obtain a high reactor power, make effective use of fissile material, and save fuel costs sufficiently and efficiently.

Further, since the power peaking factor is kept low and the radial power distribution in the reactor core is flattened, there is an advantage that the operation control of the nuclear reactor is facilitated.

[Brief explanation of the drawing]

Figure 1 is a replacement core diagram showing the fuel arrangement structure incorporated into the 1/4 core of a conventional nuclear reactor, and Figures 2 and 3 show the fuel arrangement structure incorporated into the initial core of a conventional nuclear reactor. FIG. 4 is an initial core diagram showing one embodiment of the nuclear reactor according to the present invention, for a 1/4 core, FIG. 5 is a replacement core diagram of the nuclear reactor according to the present invention, and FIG. A graph showing the core radial distribution of the infinite multiplication factor incorporated in the nuclear reactor core,
Figure 7 is a graph showing the core radial distribution of reactor power for a reactor with the fuel arrangement shown in Figure 6;
The figure is a graph showing the core radial distribution of the infinite multiplication factor incorporated in the replacement core of Figure 5, Figure 9 is a graph showing the reactor power of the replacement core of Figure 5, and Figure 10 is the infinite multiplication FIG. 11 is a graph showing the relationship between the coefficient (k∞) and fuel burnup, and FIG. 11 is a graph showing the relationship between the infinite multiplication coefficient and fuel enrichment. 10...Reactor core, B...Outermost region, C...Central region, D...Intermediate region, R...New fuel assembly.

Claims (1)

[Claims] 1. In a boiling water reactor installed in a commercial nuclear power plant with an electrical output of 500,000 kW or more, in which a fuel assembly containing fissile material is located in the reactor core, the reactor core is divided into a central region, an intermediate region, and an uppermost region. The fuel assembly with the smallest infinite multiplication factor is arranged in a part of the central region and the outermost region, and the central region and the intermediate region are divided into an outer peripheral region and a fuel assembly with an infinite multiplication factor outward from the center of the core. A nuclear reactor characterized in that fuel assemblies are arranged so that the multiplication coefficient increases sequentially.