JP3828690B2 - Initial loading core of boiling water reactor and its fuel change method - Google Patents

Description

[0001]
BACKGROUND OF THE INVENTION
The present invention relates to an initial loading core of a boiling water reactor (BWR) and a fuel replacement method thereof, and more particularly to an initial loading core suitable for high burnup and high power density and a fuel replacement method thereof.
[0002]
[Prior art]
In recent years, there has been an increasing need to reduce power generation costs for light water reactors, and as part of this, fuel economy has been improved (effective use of fissile materials → reduction of fuel costs) and the amount of waste (fuel bodies) in the initial loading core. Reducing the number of withdrawals) is an important issue. In order to cope with this problem, it is effective to increase the degree of burnout of the fuel assembly (higher burnup).
[0003]
In order to increase the degree of combustion of the fuel assembly, it is necessary to increase the uranium enrichment. However, when the uranium enrichment is increased, the difference in reactivity between cold and power operation increases due to this, and the furnace shutdown margin tends to decrease.
[0004]
It is also effective to reduce the power generation cost by increasing the power density of the core, which is about 50 kW / liter in the current BWR (higher power density), and higher power density of 10% or more is being promoted in Europe and the United States. .
[0005]
Ensuring thermal margin is an important issue to achieve high power density. Further, when the power density is increased, the extraction energy per cycle increases, so that it is necessary to increase the uranium enrichment, and the tendency of reducing the furnace shutdown margin is further expanded.
[0006]
Japanese Patent Laid-Open No. 9-105792 discloses a conventional technique related to high burnup. In this prior art,
(1) Providing a plurality of unit loading patterns consisting of three high enrichment fuel assemblies and one low enrichment fuel assembly near the center of the core;
(2) Construct a control cell with a low enriched fuel assembly;
(3) The number of fuel rods containing gadolinia in the highly enriched fuel assembly should be two or more on the non-control rod side than on the control rod side;
As a result, the thermal margin of the core with increased average enrichment is secured.
[0007]
Japanese Patent Application Laid-Open No. 5-249270 proposes an initially loaded core composed of three types of fuel assemblies having different average enrichments. In this conventional technology, a highly enriched fuel assembly is loaded on the outermost periphery of the core, and a combustible absorbent such as gadolinia is added so that the infinite multiplication factor of the highly enriched fuel assembly when not burned is not excessive. It is a feature.
[0008]
[Problems to be solved by the invention]
In the prior art described in Japanese Patent Application Laid-Open No. 9-105792, the infinite multiplication factor of the high enrichment fuel assembly is reduced by gadolinia so that it is almost equal to that of the low enrichment fuel assembly.
In addition, the side adjacent to the low enrichment fuel assembly is prevented so that the output of the fuel rod adjacent to the low enrichment fuel assembly of the high enrichment fuel assembly does not become excessive due to the difference in the neutron energy spectrum between the fuel assemblies. A large number of gadolinia fuel rods are arranged to suppress the fuel rod output in this region. Due to these effects, under the power density condition (about 50 kW / liter) of the current BWR, it is possible to achieve a high burnup in the initial loading core that is almost equivalent to that of the replacement core.
[0009]
However, in this prior art, the outermost fuel assembly has a low enrichment equivalent to that of the fuel assembly loaded in the control cell, and the output is allowed to decrease at the outer periphery of the core. Does not take into account Therefore, in order to achieve a high output density of about 10 to 15%, further measures for ensuring a thermal margin are required.
[0010]
In the prior art described in Japanese Patent Application Laid-Open No. 5-249270, a combustible absorbent such as gadolinia is added so that the infinite multiplication factor when the highly concentrated fuel assembly is not burned is not excessive. By appropriately setting the concentration of the combustible absorbent such as gadolinia to be added, it is possible to increase the output of the outer periphery of the core and flatten the power distribution in the core radial direction. However, when this highly enriched fuel assembly is loaded on the outermost periphery of the core where the neutron flux distribution is inclined, the combustion of the combustible absorbent becomes non-uniform, and it becomes difficult to flatten the local power distribution in the fuel assembly, and the core management Becomes complicated. In addition, since the outermost periphery of the core has no neutron flux distribution on the outer side, it is difficult to burn, and the highly concentrated fuel assemblies loaded on the outermost periphery of the core are non-uniformly combusted, resulting in variations in excess reactivity. It is difficult to move from the outermost periphery of the core to another location.
[0011]
An object of the present invention is to provide an initially loaded core of a boiling water nuclear reactor and a method for refueling the same, which realizes high power density and high burnup at the same time while ensuring thermal margin and reactor shutdown margin. is there.
[0012]
[Means for Solving the Problems]
(1) In order to achieve the above object, the present invention includes a plurality of fuel assemblies and a plurality of control rods disposed between the plurality of fuel assemblies, and the plurality of control rods are inserted during operation. In a first loading core of a boiling water reactor having a control rod and other control rods, the plurality of fuel assemblies include a first fuel assembly loaded around a control rod inserted during the operation; Including a second fuel assembly loaded on the outermost periphery of the core and a third fuel assembly loaded in the other region; In the other region, the second fuel assembly is not loaded, The first, second, and third fuel assemblies are configured such that the average uranium enrichment has a relationship of first fuel assembly <second fuel assembly <third fuel assembly. The first fuel assembly and the second fuel assembly are not added with a combustible absorbent, and the third fuel assembly is added with a combustible absorbent.
[0013]
1) Effective increase when one control rod is pulled out at low temperature by making the average uranium enrichment of the first fuel assembly loaded around the control rod inserted during operation (control cell) the lowest The magnification does not become excessive, and a furnace shutdown margin is secured. In addition, since the first fuel assembly does not include a combustible absorbent, the output of the first fuel assembly (low enrichment fuel assembly) becomes an appropriate level, and a thermal margin is ensured while being moderate. The combustion proceeds.
[0014]
2) By making the average uranium enrichment of the third fuel assembly loaded in the region excluding the control cell and the outermost periphery of the core highest and adding a combustible absorbent, the third fuel assembly The infinite multiplication factor of the body (highly enriched fuel assembly) can be controlled appropriately, combined with the characteristics of other areas,
・ The power distribution in the radial direction of the core can be flattened
・ It is possible to maintain the excess reactivity of the core at an appropriate level.
It becomes.
[0015]
3-1) By making the average uranium enrichment of the second fuel assembly loaded on the outermost periphery of the core higher than the average uranium enrichment of the first fuel assembly and adding no combustible absorbent, A constant power can be maintained even in the outermost neutron leakage region, and the power density of the core can be increased accordingly. In addition, since a constant power can be maintained at the outermost periphery of the core, the power distribution in the radial direction of the core can be further flattened, and the heat output in the region inside the core other than the outermost periphery can be reduced when the same power density is achieved. The thermal margin can be increased.
[0016]
3-2) In addition, by making the average uranium enrichment of the second fuel assembly loaded on the outermost periphery of the core higher than the average uranium enrichment of the first fuel assembly and adding no combustible absorbent The average uranium enrichment of the core can be increased, and the power density is increased to 51 kW / liter or more, for example, about 55 kW / liter, while maintaining the removal degree of burnout of the initially loaded core at the same level as the replacement core (about 40 GWd / t). be able to.
[0017]
3-3) Further, by making the average uranium enrichment of the second fuel assembly loaded on the outermost periphery of the core higher than the average uranium enrichment of the first fuel assembly and not adding a combustible absorbent. Combustion at the outermost periphery of the core becomes uniform, and the irradiated second fuel assembly burned at the outermost periphery of the core can be loaded into the control cell and further burned at this position. For this reason, the fuel assemblies with higher enrichment are burned in the furnace for a longer period of time, and when compared with the conditions of constant core average enrichment, the taken-out burnup can be increased.
[0018]
4) As described above, high power density and high burn-up can be achieved at the same time while ensuring thermal margin and reactor shutdown margin in the initial loading core. Thus, the power density can be increased, the plant construction cost per unit electrical output can be reduced, and the power generation cost can be reduced. In addition, the amount of spent fuel generated can be reduced.
[0019]
(2) In the above (1), preferably, the average uranium enrichment of the first fuel assembly is 1.8 wt% or less.
[0020]
By setting the average uranium enrichment of the first fuel assembly loaded in the control cell to 1.8 wt% or less, a shutdown margin of 1% Δk or more, which is the design standard in the current BWR reactor installation permission application, is secured. The This point will be described with reference to FIG.
[0021]
The stop margin is defined as the subcriticality when one maximum value control rod (or one set sharing the hydraulic control unit) is fully pulled out.
[0022]
According to the present invention, for the region (cell) loaded with the third fuel assembly (high enrichment fuel assembly), a combustible absorbent is added to the high enrichment fuel assembly, Since the infinite multiplication factor when pulling out the control rod is as small as about 1.0, even if one control rod in this region is pulled out, the effective multiplication factor will not be excessive.
[0023]
Next, for the second fuel assembly (medium enrichment fuel assembly), the infinite multiplication factor when the control rod is pulled out becomes as high as about 1.2. In the case of being composed of two highly enriched fuel assemblies, the infinite multiplication factor when the control rod is pulled out is the highest, averaging about 1.1. However, since the intermediate enrichment fuel assembly is loaded in the outermost peripheral region, the effective multiplication factor does not become excessive due to the effect of neutron leakage.
[0024]
The third is a first fuel assembly (low enrichment fuel assembly) loaded in the control cell. The effective multiplication factor when one control rod is pulled out in this region increases with the enrichment of the low enrichment fuel assembly. Therefore, the stop margin tends to decrease as the enrichment of the low enrichment fuel assembly increases.
[0025]
FIG. 6 shows an embodiment described later (a high-concentration fuel assembly having an enrichment of about 4.4 wt% and a medium enrichment fuel assembly having an enrichment of about 2.4 wt% and a low enrichment having an enrichment of about 1.7 wt%. FIG. 9 shows the analysis result of the stop margin when only the enrichment of the low enrichment fuel assembly is changed in the core in which the fuel assembly is loaded according to the loading pattern shown in FIG. 1. The design margin in the current BWR reactor installation permission application is 1% Δk or more, and the enrichment of the low enrichment fuel assembly is 1.8w% or less as shown in FIG. As a result, a stop margin of 1% Δk or more is secured.
[0026]
(3) In the above (1) or (2), preferably, the average uranium enrichment of the second fuel assembly is 2.5 wt% or less.
[0027]
The average uranium enrichment of the second fuel assembly loaded on the outermost periphery of the core is higher than the average uranium enrichment of the first fuel assembly loaded on the control cell (1.8 w% or more) and 2.5 wt. %, The combustible absorbent is not added, and as described in (1) above, a constant output can be maintained even in the region where there is a lot of neutron leakage at the outermost periphery of the core, and the subcriticality of the fuel storage facility No special consideration is required to secure This point will be described with reference to FIG.
[0028]
FIG. 7 is a characteristic diagram showing the relationship between the average uranium enrichment of the fuel assembly and the infinite multiplication factor.
[0029]
The average uranium enrichment of the second fuel assembly loaded on the outermost periphery of the core is higher than the average uranium enrichment of the first fuel assembly loaded on the control cell (1.8 w% or more) and 2.5 wt. %, The infinite multiplication factor of the second fuel assembly is 1.2 to 1.3 as shown in FIG. 7, as described in (1) above. In addition, a constant output can be maintained even in a region where there is a lot of neutron leakage at the outermost periphery of the core. Further, by setting the average uranium enrichment of the second fuel assembly to 2.5 wt% or less, as shown in FIG. 7, it is equal to or less than the maximum value (1.3) of the infinite multiplication factor of the conventional fuel assembly. Therefore, no special consideration is required to ensure the subcriticality of the fuel storage facility.
[0030]
(4) Further, in any one of the above (1) to (3), preferably, the average uranium enrichment of the core constituted by the first, second, and third fuel assemblies is 3.6 wt% or more. And
[0031]
In this way, by setting the average uranium enrichment of the core (average uranium enrichment of all fuel assemblies loaded in the core) to 3.6 wt% or more, even when the power density is 51 kW or more, the removal of the first loaded core It is possible to increase the burnup almost as much as the replacement core.
[0032]
(5) In any one of the above (1) to (4), preferably, the power density at the rated output operation of the core constituted by the first, second, and third fuel assemblies is 51 kW / Use liters or more.
[0033]
As a result, the power density can be increased as compared with the conventional one while keeping the take-off burn-up degree of the initial loading core at the same level as the replacement core, and the plant construction cost per unit electric power can be reduced as described in (1) above. The power generation cost can be reduced.
[0034]
(6) In any one of the above (1) to (5), preferably, the concentration or number of fuel rods in which the third fuel assembly has the same uranium enrichment and to which a combustible absorbent is added is different. A plurality of types of fuel assemblies are used.
[0035]
In this way, by configuring the third fuel assembly in a plurality of types, a fuel assembly having a fuel rod with a high concentration of the combustible absorbent or a combustible absorbent is added to a position where the thermal margin tends to be small. A fuel assembly with a large number of fuel rods loaded, and a fuel assembly with a fuel rod with a low concentration of the combustible absorbent or a fuel with few fuel rods to which the combustible absorbent is added at a position that does not cause severe thermal conditions By loading the assembly, the thermal margin as a whole can be further increased.
[0036]
(7) Further, in the fuel replacement method for the initial loading core of the boiling water reactor according to any one of (1) to (6), preferably, the first fuel assembly is mounted at the time of the first fuel replacement. The irradiated second fuel assembly is moved around a control rod that is taken out of the furnace and inserted during operation in which the first fuel assembly was loaded, and the second fuel assembly is moved. The third fuel assembly that has been irradiated is transferred to the outermost periphery of the core in which the fuel is loaded, and the fourth fuel assembly that has not been irradiated is newly added to the other region. Load it.
[0037]
By adopting such a fuel exchange method, a fuel assembly having a higher enrichment is burned in the furnace for a longer period of time, so that the removal burn-up can be made higher when compared under the condition of constant core average enrichment. Further, the first fuel assembly with low enrichment and the second fuel assembly with intermediate enrichment are burned in a region (control cell) where the output is relatively high except for the outer periphery of the core for at least one cycle. The fuel assembly having a low enrichment has a high burnup.
[0038]
DETAILED DESCRIPTION OF THE INVENTION
Hereinafter, embodiments of the present invention will be described with reference to the drawings.
[0039]
First, a first embodiment of the present invention will be described with reference to FIGS.
[0040]
FIG. 1 shows the fuel loading pattern (only 1/4 of the entire core) of the initial loading core of a boiling water reactor having an electric output of 1.5 million kW class (power density of about 55 kW / liter). In FIG. 1, the core has a control cell 11 formed around a control rod 51 a (see FIG. 3) inserted during operation, a core outermost periphery 21, and a region 31 other than the control cell 11. 148 low-enrichment fuel assemblies 7, 92 intermediate-enrichment fuel assemblies 8 on the outermost periphery 21 of the core, and 632 high-enrichment fuel assemblies 9 are loaded in the other region 31. .
[0041]
The average uranium enrichment of the low enrichment fuel assembly 7 is about 1.7 wt%, and no gadolinia is added. The intermediate enrichment fuel assembly 8 has an average uranium enrichment of about 2.4 wt%, and no gadolinia is added. The average uranium enrichment of the high enrichment fuel assembly 9 is about 4.4 wt%, and gadolinia is added.
[0042]
As shown in FIG. 2, each of the fuel assemblies 7, 8, 9 has a number of fuel rods 4. The upper portion of the fuel rod 4 is an upper tie plate 5a, the lower portion is a lower tie plate 5b, and the middle portion is a spacer 5c. Each is supported and bundled. Two large-diameter water rods 52 (see FIGS. 3 to 5) are arranged near the center, and the outer periphery is covered with a channel box 53 made of Zircaloy.
[0043]
The detailed enrichment distribution of the fuel assemblies 7, 8, 9 will be described with reference to FIGS.
[0044]
In FIG. 3, 51a is a control rod which is inserted into the core during operation and constitutes the control cell 11, and the fuel assembly 7 is located adjacent to the control rod 51a. The fuel assembly 7 has four types of fuel rods 5 of types 1, 2, 3 and P, and these fuel rods 5 are arranged in a 10 × 10 square lattice. Types 1, 2, and 3 are normal fuel rods having different uranium enrichments except for the natural uranium portion at the upper and lower ends of one node, and no combustible absorbent (gadolinia) is added. Type P is a part-length fuel rod. The partial-length fuel rod P has a shorter effective fuel length (the length of the portion filled with fuel pellets) of about 14/24 than the normal fuel rod, and the core in which the moderator void ratio increases during operation. This contributes to reducing pressure loss in the two-phase flow section at the top. The four types of fuel rods 5 of types 1, 2, 3 and P each have a uranium enrichment as shown in the figure, and the average enrichment of the fuel assembly 7 is about 1.7 wt%.
[0045]
In FIG. 4, 51 b is a control rod that is not inserted into the core during operation, and the fuel assembly 8 is positioned adjacent to the control rod 51 b on the outermost periphery 21 of the core. The fuel assembly 8 has five types of fuel rods 5 of types 1, 2, 3, 4 and P, and these fuel rods 5 are arranged in a 10 × 10 square lattice shape. Types 1, 2, 3 and 4 are normal fuel rods having different uranium enrichments except for the natural uranium portion at the upper and lower ends of one node, and no combustible absorbent (gadolinia) is added. Type P is a part-length fuel rod. The five types of fuel rods 5 of types 1, 2, 3, 4 and P each have uranium enrichment as shown in the figure, and the average enrichment of the fuel assembly 8 is about 2.4 wt%.
[0046]
In FIG. 5, the fuel assembly 9 is positioned around a control rod 51 b that is not inserted into the core during operation in a region 31 other than the control cell 11 and the core outermost periphery 21. The fuel assembly 9 has seven types of fuel rods 5 of types 1, 2, 3, 4, P, G1, and G2, and is configured by arranging these fuel rods 5 in a 10 × 10 square lattice shape. . Types 1, 2, 3 and 4 are normal fuel rods having different uranium enrichments except for the natural uranium portion at the upper and lower ends of one node, and no combustible absorbent (gadolinia) is added. Type P is a part-length fuel rod. Types G1 and G2 are fuel rods to which gadolinia is added as a flammable absorbent, and the concentration and distribution of gadolinia are different between G1 and G2. The smaller the number, the higher the average gadolinia concentration. That is, the density of gadolinia is G1> G2. Here, the fuel rods G1 and G2 added with gadolinia are concentrated in the region opposite to the control rod 51b and are the same as in the prior art. Seven types of fuel rods 5 of types 1, 2, 3, 4, P, G1, and G2 each have a uranium enrichment as illustrated, and the average enrichment of the fuel assembly 8 is about 4.4 wt%. ing.
[0047]
The average enrichment of the core obtained by multiplying the average enrichment of the fuel assemblies 7, 8 and 9 loaded in the core by the number of each assembly and summing them is divided by the total number of assemblies. 0.7 wt%. As a result, as described above, an enrichment capable of achieving a high burn-up level equivalent to that of the prior art while ensuring a high output density of about 55 kW / liter is secured.
[0048]
In the present embodiment configured as described above, the low enrichment fuel assembly 7 has an average uranium enrichment of about 1.7 wt% and no gadolinia is added, and the intermediate enrichment fuel assembly 8 has an average uranium enrichment. When the degree is about 2.4 wt%, no gadolinia is added, and the high enrichment fuel assembly 9 has an average uranium enrichment of about 4.4 wt% and gadolinia is added. By setting the average enrichment of the core composed of the aggregates 7, 8, and 9 to about 3.7 wt% and increasing the power density of the core to about 55 kW / liter, the following effects can be obtained.
[0049]
By making the average uranium enrichment of the low enrichment fuel assembly 7 loaded in the control cell 11 the lowest, as described in 1) of item (1) of “Means for Solving the Problems”, If one control rod is pulled out, this effective multiplication factor will not become excessive, and a furnace shutdown margin will be secured. In particular, by setting the average uranium enrichment of the low enrichment fuel assembly 7 to about 1.7 wt% which is equal to or less than 1.8 wt%, as described with reference to FIG. A stop margin of 1% Δk or more, which is the design standard in the installation permission application, is secured. In addition, since the low enrichment fuel assembly 7 does not include a combustible absorbent (gadolinia), the output of the low enrichment fuel assembly is at an appropriate level, and a thermal margin is ensured and combustion proceeds moderately. .
[0050]
The average uranium enrichment of the high enrichment fuel assembly 9 loaded in the region 31 excluding the control cell 11 and the outermost outer periphery 21 of the core is the highest at about 4.4 wt%, and a combustible absorbent (gadolinia) is added. By adopting the configuration, the infinite multiplication factor of the highly enriched fuel assembly 9 can be appropriately controlled as explained in 2) of the same item (1), and the power distribution in the radial direction of the core along with the characteristics of other regions As a result, the excess reactivity of the core can be maintained at an appropriate level.
[0051]
1. The average uranium enrichment of the intermediate enrichment fuel assembly 8 loaded on the outermost periphery 21 of the core is higher than the average uranium enrichment of the low enrichment fuel assembly 7 and is 1.8 w% or more and 2.5 wt% or less. By setting it to 4 wt% and not adding a flammable absorbent (gadolinia), as explained in 3-1) of the same (1) term and further explained in FIG. 7 in the same (3) term, The infinite multiplication factor of the intermediate enrichment fuel assembly 8 is 1.2 to 1.3, and a constant power can be maintained even in the region where the neutron leakage is large in the outermost periphery 21 of the core, and the power density of the core can be increased accordingly. it can. In addition, since a constant output can be maintained at the core outermost periphery 21, the power distribution in the core radial direction can be further flattened, and the heat output in the region inside the core other than the outermost periphery 21 when the same power density is achieved. Can be reduced, and the thermal margin can be increased. Further, by setting the average uranium enrichment of the intermediate enrichment fuel assembly 8 to 2.5 wt% or less, it is equivalent to the maximum value (1.3) of the infinite multiplication factor of the conventional fuel assembly as shown in FIG. It is limited to the following infinite multiplication factor, and no special consideration is required to ensure the subcriticality of the fuel storage facility.
[0052]
Also, by making the average uranium enrichment of the intermediate enrichment fuel assemblies 8 loaded on the outermost periphery 21 of the core higher than the average uranium enrichment of the low enrichment fuel assemblies 7, 3-2 of the same (1) section As described in, the average uranium enrichment of the core can be increased, and the power density is 55 kW / liter, which is 51 kW / liter or more, while maintaining the removal degree of burnout of the initial loading core at the same level as the replacement core (about 40 GWd / t). Can be raised to a degree.
[0053]
Furthermore, by making the average uranium enrichment of the intermediate enrichment fuel assembly 8 loaded on the outermost periphery 21 of the core higher than the average uranium enrichment of the low enrichment fuel assembly 7 and not adding a combustible absorbent. Combustion at the outermost periphery 21 of the core becomes uniform, and the irradiated intermediate enriched fuel assembly 8 burned at the outermost periphery 21 of the core can be loaded on the control cell 11 and further burned at this position (described later). For this reason, the fuel assemblies with higher enrichment are burned in the furnace for a longer period of time, and when compared with the conditions of constant core average enrichment, the taken-out burnup can be increased.
[0054]
Then, by setting the average uranium enrichment of the core composed of the fuel assemblies 7, 8, and 9 to 3.6 wt% or more, as described in the section (4), the power density of the core is 51 kW. Even in the case described above, it is possible to increase the removal burn-up degree of the initially loaded core almost equal to that of the replacement core. Further, by setting the power density at the rated power operation of the core composed of the fuel assemblies 7, 8, 9 to 51 kW / liter or more, as described in the section (5), the removal combustion of the initial loading core is performed. While maintaining the same level as the replacement core, the power density can be increased more than before, the plant construction cost per unit electrical output can be reduced, and the power generation cost can be reduced.
[0055]
As described above, according to the present embodiment, high power density and high burn-up can be realized at the same time while ensuring a thermal margin and a reactor shutdown margin in the initially loaded core. While maintaining the replacement core level, the power density can be increased, the plant construction cost per unit electrical output can be reduced, and the power generation cost can be reduced. In addition, the amount of spent fuel generated can be reduced.
[0056]
Next, an embodiment of the fuel exchange method of the present invention will be described with reference to FIGS. 1, 8, and 9.
[0057]
FIG. 8 shows a fuel exchange method after the first loaded core of FIG. 1 has completed one or two operation cycles. The low enrichment fuel assembly 7 loaded in the control cell 11 in FIG. To be taken out. In the transition core of FIG. 8, the number of control cells 11 is reduced as compared with the initial loading core of FIG. 1, and the medium enriched fuel 8 loaded on the outermost periphery 21 of the core in FIG. Is done. In the example shown in the figure, the fuel assemblies loaded in the control cell 11 of the transition core in FIG. 8 are a total of 25 as viewed at 1/4 of the entire core, whereas in the initial loading core in FIG. The total number of fuel assemblies 8 loaded on the outermost periphery 21 is 23, and two of the shortage are in the central region of the highly concentrated fuel 9 loaded in the region 31 in FIG. The most burned one located is selected and loaded into the control cell 11.
[0058]
A part of the highly enriched fuel assembly 9 in FIG. 1 is moved to the outermost periphery 21 of the core. As a part of the high enrichment fuel assembly 9, the high enrichment fuel 9 loaded in the region 31 in FIG. 1 and the infinitely enriched fuel 8 loaded in the outermost core 21 in FIG. Select a gain that is relatively close. In the region 31 of FIG. 8, the remaining highly enriched fuel assembly 9 and the unirradiated replacement fuel assembly 6 are loaded. The average enrichment of the replacement fuel assembly 6 is about 4.4 wt%, and has the same burnup distribution as the high enrichment fuel assembly 9 shown in FIG.
[0059]
FIG. 9 shows a fuel change method after one operation cycle is completed from the core of FIG. 8. The medium enrichment fuel assembly 8 and the high enrichment fuel assembly 9 loaded in the control cell 11 are shown in FIG. A part is taken out. A part of the highly enriched fuel assembly 9 is loaded into the control cell 11 and a replacement fuel assembly 6 is newly loaded.
[0060]
By adopting such a fuel exchange method, the fuel assembly 9 having a higher enrichment is burned in the furnace for a longer period of time, so that the extracted burn-up can be made higher when compared under the condition that the core average enrichment is constant. Further, the low enrichment fuel assembly 7 and the intermediate enrichment fuel assembly 8 are burned in the control cell 11 in a region where the output is relatively high except for the outer periphery of the core, that is, the fuel having a low enrichment. The burnup of the aggregate increases.
[0061]
The initial loading core according to the second embodiment of the present invention will be described with reference to FIGS. In the above embodiment, one type of high enrichment fuel assembly 9 is used. However, this embodiment uses a plurality of types having different average uranium enrichments and different gadolinia concentrations and fuel rods containing gadolinia. is there.
[0062]
In the first embodiment, instead of the high enrichment fuel fuel assembly 9 shown in FIG. 5, two high enrichment fuel fuel assemblies 91 and 92 are used as shown in FIGS. An embodiment is conceivable. Here, the high enrichment fuel assemblies 91 and 92 shown in FIGS. 10 and 11 both have an average enrichment of about 4.4 wt%, which is the same as that of FIG. The number of fuel rods is different from 20 and 16, respectively, and 18 in FIG.
[0063]
As the number of fuel rods to which gadolinia is added increases, the infinite multiplication factor decreases and the fuel assembly output decreases, so the thermal margin of the fuel assembly tends to increase. Therefore, as in the loading pattern example shown in FIG. 12, thermal neutrons flow from the control cell 11 with low enrichment, and the thermal margin is likely to be reduced by the fuel rod of the highly enriched fuel assembly facing the control cell 11. A fuel assembly 91 having a large number of gadolinia fuel rods as shown in FIG. 10 is loaded at a position (position adjacent to the control cell 11), and gadolinia fuel is placed at a position (a position other than the above) where thermal conditions are not severe. By loading the fuel assembly 92 having a small number of rods, it is possible to increase the thermal margin of the entire core.
[0064]
Even if the number of gadolinia-added fuel rods is different in the initial stage, the gadolinia has the same characteristics after burning, so in the fuel exchange method of the present invention, the difference in the number of gadolinia-added fuel rods is ignored. Can be handled as a highly enriched fuel assembly.
[0065]
【The invention's effect】
According to the present invention, high power density and high burn-up can be realized at the same time while ensuring a thermal margin and a reactor shutdown margin in the initial loading core, and as a result, the removal burn-up degree of the initial loading core can be changed. While maintaining the same level, the power density can be increased, the plant construction cost per unit electrical output can be reduced, and the power generation cost can be reduced. In addition, the amount of spent fuel generated can be reduced.
[Brief description of the drawings]
FIG. 1 is a diagram (1/4 of all cores) showing a fuel loading pattern of an initial loading core of a boiling water reactor according to a first embodiment of the present invention.
FIG. 2 is a structural diagram of a fuel assembly in the initial loading core of the first embodiment shown in FIG.
FIG. 3 is an enrichment distribution diagram of a low enrichment fuel assembly in the initially loaded core of the first embodiment shown in FIG. 1;
4 is an enrichment distribution diagram of a medium enrichment fuel assembly in the initial loading core of the first embodiment shown in FIG. 1; FIG.
FIG. 5 is an enrichment distribution diagram of a highly enriched fuel assembly in the initial loading core of the first embodiment shown in FIG. 1;
FIG. 6 is a relationship characteristic diagram between an average enrichment of a fuel assembly loaded on a control cell and a furnace shutdown margin showing the operation of the present invention.
FIG. 7 is a characteristic diagram of the relationship between the average enrichment of the fuel assembly and the infinite multiplication factor showing the operation of the present invention.
FIG. 8 is a diagram (1/4 of the entire core) showing a fuel loading pattern for explaining an embodiment of the fuel exchange method of the present invention.
FIG. 9 is a diagram (1/4 of the entire core) showing a fuel loading pattern for explaining an embodiment of the fuel exchange method.
FIG. 10 is an enrichment distribution diagram of one of two types of highly enriched fuel assemblies in an initial loading core of a boiling water reactor according to a second embodiment of the present invention.
FIG. 11 is an enrichment distribution diagram of the other of the two types of high enrichment fuel assemblies in the initial loading core of the boiling water reactor according to the second embodiment.
12 is a diagram (1/4 of the entire core) showing a fuel loading pattern of the initial loading core according to the second embodiment shown in FIGS. 10 and 11. FIG.
[Explanation of symbols]
4 Fuel rod
5a Upper tie plate
5b Lower tie plate
5c Spacer
6 Replacement fuel assembly
7 Low enrichment fuel assembly (first fuel assembly)
8 Medium enrichment fuel assembly (second fuel assembly)
9 High enrichment fuel assembly (third fuel assembly)
11 Control cell
21 outermost core
31 Other areas
51a Control rod (control rod inserted during operation)
51b Control rod (control rod not inserted during operation)
52 Water Rod
53 Channel box
91,92 High enrichment fuel assembly (third fuel assembly)
G1, G2 Fuel rod with gadolinia
P Partial length fuel rod

Claims (7)

A boiling water nuclear reactor comprising a plurality of fuel assemblies and a plurality of control rods disposed between the plurality of fuel assemblies, the control rods inserted by the plurality of control rods during operation and other control rods In the first loading core of
The plurality of fuel assemblies are loaded in a first fuel assembly loaded around a control rod inserted during the operation, a second fuel assembly loaded on the outermost periphery of the core, and other regions. A third fuel assembly, and the second region is not loaded in the other region,
Each of the first, second, and third fuel assemblies has an average uranium enrichment.
The first fuel assembly <the second fuel assembly <the third fuel assembly,
A boiling water atom characterized in that a combustible absorbent is not added to the first fuel assembly and the second fuel assembly, and a combustible absorbent is added to the third fuel assembly. The first loading core of the furnace.   The initial loading core of the boiling water reactor according to claim 1, wherein the average uranium enrichment of the first fuel assembly is 1.8 wt% or less. .   The initial loading core of the boiling water reactor according to claim 1 or 2, wherein the average uranium enrichment of the second fuel assembly is 2.5 wt% or less. Loading core.   The initial loading core of the boiling water reactor according to any one of claims 1 to 3, wherein an average uranium enrichment of the core composed of the first, second, and third fuel assemblies is 3.6 wt. The initial loading core of a boiling water reactor, characterized by being over%.   In the initial loading core of the boiling water reactor according to any one of claims 1 to 4, the power density at the rated output operation of the core composed of the first, second, and third fuel assemblies is set. An initial loading core of a boiling water reactor characterized by being 51 kW / liter or more.   The initial loading core of the boiling water reactor according to any one of claims 1 to 5, wherein the concentration or number of fuel rods in which the third fuel assembly is equivalent in uranium enrichment and to which a combustible absorbent is added. A first loaded core of a boiling water reactor characterized by the fact that the fuel assemblies are of different types. 7. The method for exchanging fuel in an initially loaded core of a boiling water reactor according to claim 1, wherein the first fuel assembly is taken out of the reactor at the time of the first fuel exchange, and the first fuel is removed. The irradiated second fuel assembly is moved around the control rod inserted during the operation in which the assembly was loaded, and the outermost core of the core where the second fuel assembly was loaded A fuel exchange method , wherein the irradiated third fuel assembly is moved and a non-irradiated fourth fuel assembly is newly loaded in the other region .

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