JP2003222694A - Light water reactor core, fuel assembly, and control rod - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a light water reactor core directed to a Pu multirecycle in which a breeding ratio is in the vicinity of 1.0 or 1.0 or more while maintaining a negative void coefficient with approximately the same cost effectiveness and safety as those of a BWR operating at present. <P>SOLUTION: The light water reactor core comprises fuels in which depleted uranium is enriched with plutonium or plutonium and nuclides. In the light water reactor core, the void coefficient is made negative by making part of the horizontal cross sections of the fuel assembly, in the direction of the height of the reactor core except blanket parts at both upper and lower ends, in which the average degree of enrichment of fissionable plutonium is 6 wt.% or more, between 40 cm and 140 cm and making the degree of fissionable plutonium enrichment at the upper part of the reactor core higher than that of fissionable plutonium at the lower part of the reactor core, and the breeding ratio is made in the vicinity of 1.0 or 1.0 or more by making the average void coefficient of the reactor core between 45-70% when the reactor is operated at rated output of 50% or more. <P>COPYRIGHT: (C)2003,JPO

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a light water reactor core, a fuel assembly and a control rod constituting the core, and particularly in a BWR, the economical efficiency and safety are the same as those of a currently operating BWR, that is, in the reactor. The present invention relates to a light water reactor core, a fuel assembly, and a control rod which are intended for Pu multi-recycling in which the change in structure is minimized and the negative void coefficient is maintained, and the breeding ratio is around 1.0 or slightly above 1.0.

[0002]

2. Description of the Related Art A nuclear fission reaction inside a nuclear reactor causes exhaustion of fissile materials such as uranium-235 and plutonium-239, and uranium-238 and plutonium-2.
The conversion of fuel parent substances such as 40 into fissile material is occurring. The ratio of the amount of fissile material contained in the fuel extracted from the core to the amount of fissile material contained in the fuel loaded in the core is called the breeding ratio. In conventional light water cooled nuclear reactors, the breeding ratio is about 0.5. Is. As a method of effectively utilizing uranium resources, increasing the growth ratio is considered.

JP-A-55-10591 and 1982
212-227 of 59 volumes of Nuclear Technology magazine
The page shows that in a pressurized water reactor, fuel rods can be densely arranged in a triangular lattice to reduce the water-to-fuel volume ratio to improve the breeding ratio. However,
The breeding ratio is about 0.9 at most, and it is necessary to supplement the fissile material in order to continue the operation without reducing the output. To further increase the breeding ratio, it is conceivable to narrow the fuel rod gap and further reduce the water to fuel volume ratio, but there are limits in terms of manufacturing fuel assemblies and securing thermal margin, etc.
Realization is difficult.

On the other hand, Japanese Patent Laid-Open No. 1-227993 discloses a method of effectively reducing the volume ratio of water to fuel by utilizing steam voids generated in the core, which is a feature of boiling water reactors. Has been done. However, even in the conventional example, the plutonium multiplication ratio (the ratio of the amount of fissionable plutonium contained in the fuel extracted from the core to the amount of fissionable plutonium contained in the fuel loaded into the core; the multiplication ratio to the fissionable plutonium) is close to 1 Although it has been shown that the growth ratio (when plutonium is enriched in natural uranium, it is about 4-5% smaller than the plutonium multiplication ratio)
Is not shown to be near 1 or above 1. When the plutonium multiplication ratio is close to 1, it is necessary to enrich plutonium with natural uranium in order to continue the operation without reducing the output, and it is not possible to use all uranium resources. In the present invention, a growth ratio of around 1 is 0.98.
It means the above.

[0005]

The first object of the present invention is to maintain the power generation cost, thermal margin and safety at the same level as the currently operating light water reactor, and to contribute to long-term stable supply of energy and fuel assembly. To provide the body.

A second object of the present invention is to achieve a breeding ratio of 1.0 with a fuel enriched with Pu in depleted uranium by reducing the water-to-fuel volume ratio to contribute to a long-term stable supply of energy. To provide a core, a fuel assembly and a control rod.

A third object of the present invention is to contribute to a long-term stable supply of energy, so that the required Pu inventory per unit output can be reduced, and a core capable of operating as many reactors for power generation as possible with a constant amount of Pu. , To provide a fuel assembly.

A fourth object of the present invention is to make the power generation cost comparable to that of the present light water reactor, so that the same material, the same size of the pressure vessel, the same output and burnup as the currently operating reactor are used. Is to provide a core and a fuel assembly that can achieve the above with a similar thermal margin.

A fifth object of the present invention is to increase neutron leakage in the height direction of the core and power distribution swing in the height direction of the core at the time of power increase in order to make the safety comparable to that of the present light water reactor. , To provide a core and a fuel assembly that can realize a negative void coefficient.

A sixth object of the present invention is to maintain the distillation function by boiling and to confine the radioactive material existing in the reactor in the pressure vessel in order to make the safety comparable to that of the present light water reactor. It is to provide a core that can.

A seventh object of the present invention is to provide a core and a fuel assembly in which Pu alone extraction is abolished and Pu and U are integrally recycled to cope with nuclear non-proliferation.

An eighth object of the present invention is to provide a core and a fuel assembly in which an actinide nuclide is allowed to stay in a reactor together with uranium and Pu to be recycled in order not to leave a long-lived radioactive waste for future generations. It is to be.

[0013]

In order to achieve the above first object, according to the present invention, in a core having Pu or Pu and a fuel enriched with actinide nuclide in depleted uranium, the upper and lower ends are 40 cm to 14 cm in the horizontal section of the fuel assembly where the average enrichment of fissile plutonium is 6 wt% or more excluding the blanket part
By setting the void ratio to 0 cm, the void coefficient is made negative, and the void ratio of the core average when operating at 50% or more of the rated output is set to 45 to 70%, so that the breeding ratio is near 1.0 or 1. A zero or more core is provided.

In order to achieve the above first object,
According to the present invention, uranium containing at least one of natural uranium, depleted uranium, and low enriched uranium is added with Pu or Pu.
In a core having a fuel enriched with and actinide nuclides, in the height direction of the core excluding the blanket parts at both upper and lower ends, the portion where the average enrichment of fissile plutonium in the horizontal cross section of the fuel assembly is 6 wt% or more is 40 cm to 140 cm.
By setting the void ratio to a negative value, the void ratio of the core average when operating at 50% or more of the rated output is set to 45 to 70%, so that the breeding ratio is close to 1.0 or 1.0. The core described above is provided.

In order to achieve the above first object, according to the present invention, depleted uranium, natural uranium, depleted uranium,
In a fuel assembly having Pu or a fuel enriched with Pu and an actinide nuclide in uranium containing at least one of low-enriched uranium, an effective water-fuel volume ratio of 0.1
By setting the ratio between 1 and 0.6, a fuel assembly having a growth ratio of around 1.0 or at least 1.0 is provided.

In order to achieve the second object,
According to the present invention, in a hexagonal close-packed fuel assembly in which fuel rods are arranged in an equilateral triangular lattice, the fuel rods are characterized in that the gap between the fuel rods is 0.7 to 2.0 mm. ,
A core composed of the fuel assembly is provided.

In order to achieve the second object,
According to the present invention there is provided a core and fuel assembly characterized by an effective water to fuel volume ratio of between 0.1 and 0.6.

Further, in order to achieve the above second object,
According to the present invention, there is provided a light water reactor core having an average enrichment of fissionable Pu of 6 to 20% in the core portion excluding the outer peripheral portion and the blanket portions at the upper and lower ends.

In order to achieve the second object,
According to the present invention, there is provided a fuel assembly characterized in that the average fissile Pu enrichment in the region excluding the blanket portions at the upper and lower ends is 6 to 20%.

In order to achieve the second object,
According to the present invention, there is provided a boiling water type light water reactor core characterized in that the average void fraction of the core when operated at 50% or more of the rated output is 45 to 70%.

Further, in order to achieve the above second object,
According to the present invention, there is provided a light water reactor core including a hexagonal fuel assembly and a cluster-type control rod inserted therein.

In order to achieve the second object,
According to the present invention, there is provided a light water reactor core composed of a hexagonal fuel assembly and a Y-shaped control rod having three blades each having an interval of 120 degrees inserted between them.

Further, in order to achieve the above second object,
According to the present invention, the Y adjacent to the regular hexagonal fuel assembly integrally
A light water reactor characterized in that each of the V-shaped control rods has two or less blades, and the gap between the fuel assemblies in which the blades are not inserted between the fuel assemblies is narrower than the gap between the fuel assemblies in which the blades are inserted. A core is provided.

Further, in order to achieve the above second object,
According to the present invention, it is composed of square fuel assemblies densely arranged in an equilateral triangular lattice and cross-shaped control rods having four blades, each of which has a blade interval of 90 degrees. A light water reactor core is provided.

In order to achieve the above second object,
According to the present invention, in a hexagonal fuel assembly in which fuel rods are arranged in an equilateral triangular lattice, two sets of fuel rods among three sets of fuel rods parallel to the facing outermost layer fuel rods are arranged. And the number of the fuel rods is one more than the number of the remaining set of fuel rods, and a regular hexagonal fuel assembly lattice is constituted by one blade of the Y-shaped control rod. A light water reactor core is provided.

In order to achieve the second object,
According to the present invention, in order to eliminate the moderator at the upper end of the control rod, a substance having a moderator capacity smaller than that of light water, such as carbon, deuterium,
Provided is a light water reactor core including cluster type, Y-type or cross type control rods in which follower parts made of beryllium, Zr alloy, stainless steel, etc. are installed.

In order to achieve the above second object,
According to the present invention, in a regular hexagonal and eclipse type hexagonal fuel assembly, at least two regions from a region close to the Y-shaped control rod to a region apart from the Y-shaped control rod are fissile. There is provided a hexagonal fuel assembly and an erosion-type hexagonal fuel assembly characterized by being constituted by a plurality of types of fuel rods having different Pu enrichment levels, particularly 2 to 5 types of fuel rods.

In order to achieve the above second object,
According to the present invention, in a square fuel assembly, a plurality of types having different enrichment levels of fissile Pu for at least two regions ranging from a region close to the cross control rod to a region separated from the cross control rod. In particular, a square fuel assembly is provided, which is characterized by being composed of 2 to 5 types of fuel rods.

In order to achieve the above third, fourth and fifth objects, according to the present invention, the average power density of the core portion excluding the outer peripheral portion and the blanket portions of the upper and lower ends is 100 k.
There is provided a light water reactor core characterized in that it is W / l to 300 kW / l.

In order to achieve the above third, fourth and fifth objects, according to the present invention, in a core having Pu or Pu and a fuel enriched with actinide nuclide in depleted uranium, upper and lower end portions of the core In the height direction excluding the blanket part of the above, there is provided a light water reactor core characterized in that a portion of the horizontal cross section of the fuel assembly having an average fissile Pu enrichment of 6 wt% or more is between 40 cm and 140 cm.

In order to achieve the above-mentioned third, fourth and fifth objects, according to the present invention, Pu or Pu and an actinide nuclide are added to uranium containing at least one of natural uranium, depleted uranium and low enriched uranium. In the core having the fuel enriched with the above, in the height direction excluding the blanket parts at the upper and lower ends of the core, the portion where the average fissile Pu enrichment in the horizontal cross section of the fuel assembly is 6 wt% or more is 40 cm to 140 cm.
A light water reactor core is provided.

In order to achieve the above third, fourth, and fifth objects, according to the present invention, a fuel assembly having Pu or Pu and a fuel enriched with actinide nuclide in depleted uranium is provided at both upper and lower ends. 40 cm and 140 cm in the height direction of the fuel assembly excluding the blanket part in the horizontal part, where the average fissile Pu enrichment in the horizontal section is 6 wt% or more
A fuel assembly is provided.

In order to achieve the above-mentioned third, fourth and fifth objects, according to the present invention, Pu or Pu and an actinide nuclide are added to uranium containing at least one of natural uranium, depleted uranium and low enriched uranium. In a fuel assembly having a fuel enriched with hydrogen, the fissionability P of the horizontal cross section in the height direction of the fuel assembly excluding the blanket portions at the upper and lower ends.
u The area where the average enrichment is 6 wt% or more is 40 cm and 1
A fuel assembly is provided that is between 40 cm.

In order to achieve the above-mentioned fourth object,
According to the present invention, the core is divided into two equal areas in the radial direction except for the outermost periphery of the core, so that the average value of the number of fuel assembly core stay cycles in the core outer region is smaller than that in the core inner region. There is provided a light water reactor core characterized by being loaded with a fuel assembly.

In order to achieve the above-mentioned fourth object,
According to the present invention, there is provided a light water reactor core characterized in that the average value of the orifice pressure loss coefficient of the outermost periphery of the core and the fuel assemblies adjacent thereto is larger than the average value of the orifice pressure loss coefficient of the other regions.

In order to achieve the fifth object,
According to the present invention, there is provided a hexagonal fuel assembly characterized in that, except for the blanket portions at both upper and lower ends, the average value of the lower half of the fissile Pu enrichment is lower than the average value of the upper half. It

In order to achieve the fifth object,
According to the present invention, the fissionable Pu enrichment is 6 in the height direction of the fuel assembly excluding the blanket parts at the upper and lower ends.
There is provided a fuel assembly characterized in that there are upper and lower portions of w / o or more, and a fissionable Pu enrichment in a region near the center between them is 6 w / o or less.

In order to achieve the sixth object,
According to the present invention, there is provided a boiling water type light water reactor core characterized in that the steam weight ratio at the outlet of the coolant core when operating at 50% or more of the rated output is between 20% and 40%. Provided.

In order to achieve the seventh object,
According to the present invention, there is provided a light water reactor core and a fuel assembly which are characterized by simultaneously recycling Pu and uranium.

In order to achieve the above eighth object,
According to the present invention, there is provided a light water reactor core and a fuel assembly characterized by simultaneously recycling Pu, uranium and actinides.

[0041]

According to the study by the inventors of the present application, the following has been found.

The amount of natural uranium resources in the world is estimated to be around 15 million tons, which corresponds to the amount of 1,000 existing light water reactors with an electric output of 1 million kW that can be operated for about 100 years. As a result, less than 15 million tons of depleted uranium and 15 million tons of fissile Pu are left. Therefore, with the fuel enriched with Pu in depleted uranium, the electric output 10
The inventory of fissionable Pu, including the inside and outside of the reactor per 10,000 kW, is 10 tons, and the power generation reactor (RB) with a breeding ratio of 1.0
WR) makes it possible to continue fission with depleted uranium using Pu as a catalyst, and uranium generates thermal energy of about 1 MWD per gram, so 1500 RBWRs can be operated for 10,000 years. Since it becomes possible and the total amount of uranium resources can be used up, it contributes to the first purpose of long-term stable supply.

The second object is achieved by the following operation. According to the study by the inventors of the present application, the following has been found regarding the relationship between the breeding ratio and the effective water-to-fuel volume ratio in the LWR core. The effective water-fuel volume ratio [(Vm / Vf) eff] is a geometrical water-fuel volume ratio [(Vm / Vf) eff] considering that steam voids are generated in the core.
m / Vf) geo; water to fuel volume ratio at which steam voids do not occur] is expanded. Assuming that the reduction rate of hydrogen density due to the generation of vapor voids is F, both have the following relationship.

[0044]

## EQU00001 ## (Vm / Vf) eff = F (Vm / Vf) geo (Equation 1) Further, F has the following relationship with the core-average steam void rate V (%).

F = (100-V) / 100 + fV / 100 Here, f: Ratio of saturated vapor density to saturated water density In general, f is a small value of about 1/20, and F can be approximated as follows. .

F.apprxeq. (100-V) / 100 FIG. 2 shows the relationship between the effective water-to-fuel volume ratio and the conversion ratio defined from the neutron balance, and the three factors constituting the conversion ratio.

[0047]

## EQU00002 ## Conversion ratio = .alpha. (1 + .beta.)-(1 + .gamma.) (Equation 2) where .alpha .: number of new neutrons generated when neutron is absorbed by fissile material and one fissile material disappears .beta .: Addition due to nuclear fission in the fast energy region of the fuel parent substance γ: Ratio of wasteful capture of neutrons (including neutron leakage) to neutron absorption by fissile material The effective water-to-fuel volume ratio is , About 2.0, and the growth ratio is about 0.5. In order to achieve a growth ratio close to 1, the conversion ratio must be close to 1. According to a study by the inventors of the present application, fissile Pu within the range described below.
It has been found that by increasing the enrichment of neutrons and increasing the neutron leakage to the blanket, it is possible to achieve a breeding ratio of around 1 at a conversion ratio of 0.85 or more. Therefore, the effective water-to-fuel volume ratio is 0.6 or less. On the other hand, in order to keep the effective water-to-fuel volume ratio below 0.1, the core average steam void fraction must exceed 70%. Cannot be maintained.

The effective water-to-fuel volume ratio of 0.1-0.6 is
This is achieved by arranging the fuel rods densely, utilizing steam voids generated in the core, or inserting a follower at the control rod insertion position to eliminate the moderator when the control rod is not inserted. Alternatively, it is realized by combining the above three. FIG. 3 shows an example of the relationship between the fuel rod gap and the geometrical water-to-fuel volume ratio. In FIG. 3, the fuel rod diameter is set to a range of about 9.5 to 12.3 mm currently used in light water reactors, and an equilateral triangular fuel rod lattice is targeted. When the fuel rod gap is set to 2 mm or less, the (Vm / Vf) geo of the fuel rod lattice becomes about 0.9 or less. In the case of a fuel assembly in which fuel rods are densely arranged in an equilateral triangular lattice, considering the gap area between fuel assemblies and the control rod insertion area, (V
m / Vf) geo is less than (Vm / Vf) geo of the fuel rod lattice.
The value increases from 1 to 0.2. Therefore, in order to realize an effective water-to-fuel volume ratio of 0.6 or less under this geometrical water-to-fuel volume ratio, the core average steam void fraction is 45% or more ( From the relationship diagram shown in FIG. 27, the steam weight ratio at the core outlet needs to be 20% or more). On the other hand, when the fuel rod gap is 0.7 (the minimum value of the fuel rod gap from the viewpoint of manufacturing the fuel assembly and securing thermal margin), it is in the range of 1.0 mm (when the fuel rod diameter is thicker than 9.5 mm). Can be 1.0 mm or more), the vapor void ratio is 0%, and (V
m / Vf) geo can be about 0.6 or less.

FIG. 4 shows the relationship between the average fissile Pu enrichment of the fuel assembly and the growth ratio. In order to maintain the core in the critical state throughout the operation period, the fissile Pu enrichment is set to 6
It is necessary to make it wt% or more. On the other hand, the growth ratio is
Although it decreases with the fissile Pu enrichment, it was found that a proliferation ratio of around 1 can be achieved up to 20 wt% by utilizing the increase in excess reactivity and increasing the neutron leakage to the blanket as described above.

At this time, as a means for controlling the reactivity of the furnace, a cluster type control rod is inserted into the fuel assembly, a Y-shaped control rod is provided around the hexagonal fuel assembly, or A method of inserting a cross-shaped control rod around the square fuel assembly can be considered. Proliferation ratio of 1.0 due to the above combination
The core of can be realized.

Further, by the following action, the third, fourth, fifth
The purpose of is achieved. According to the study by the inventors of the present application, by making the output of the fuel assembly per unit horizontal cross section of the core to be about the same as that of the existing boiling water type light water reactor, the height of the fuel assembly can be ensured while ensuring a thermal margin. It was found that (effective core length: length of region where average fissile Pu enrichment in horizontal section is 6 wt% or more) can be reduced. Effective water to fuel volume ratio of 0.6
For the following reasons, as a result of densely arranging the fuel rods, the number of fuel rods per unit horizontal cross section of the core is 3 to 4 times that of the existing boiling water type light water reactor. Therefore, the height of the fuel assembly (effective core length) at which the average linear power densities are the same is about 1/3 to 1/4 times that of the existing boiling water type light water reactor. In addition, compared to the existing boiling water reactor, the moderator is uniformly distributed, so the local power peaking coefficient of the fuel rod (by adopting the enrichment distribution if necessary) can be reduced by about 30% or more. .
In addition, the change in combustion reactivity and the change in void reactivity are small, and in combination with other means described below, the output peaking coefficient can be reduced by about 40% or more. Therefore, the height of the fuel assembly (effective core length) at which the average linear power density is equal to or higher than that is about 1/10 of the current boiling water type light water reactor.
It will be 0 cm or more. On the other hand, by shortening the effective core length and increasing the neutron leakage in the axial direction, the void coefficient reduction effect can be utilized. According to the study by the inventors of the present application, it was found that if the effective core length is 140 cm or less, a negative void coefficient can be realized in combination with other means described below. The power generation ratio in the blanket part increases due to the shortening, but the average power density in the region excluding the blanket part becomes about 100 to 300 kW / l due to the decrease in the effective core length.

As a result, since the RBWR having the same output is contained in the pressure vessel having almost the same diameter as that of the existing reactor, the power generation cost can be maintained at the same level as that of the existing light water reactor, and the safety is also the same as that of the existing light water reactor. Can be kept at the standard of. In addition, this makes it possible to reduce the Pu inventory, and therefore, it is possible to operate a large number of power generation reactors with a fixed amount of Pu, and a stable energy supply can be realized.

Further, the fifth object is achieved by the following operation. According to the study by the inventors of the present application, by making the enrichment degree of fissile Pu in the upper part of the core higher than that in the lower part of the core, the axial power distribution of the core can be flattened, and as a result, the Pu inventory can be reduced. Can be made. Further, when the power rises or the core coolant flow rate decreases, the vapor void fraction in the core rises, but at that time, as shown in FIG. 5, the fissile Pu enrichment is relatively low. , The power distribution swings to the lower part of the core where the neutron importance is small, and the reactivity of the core can be reduced (with a negative void coefficient).

The second object is achieved by the following action. According to the study by the inventors of the present application, by inserting the cluster-type control rods into the fuel assembly, it is possible to realize a core structure in which the fuel rods are densely arranged in a regular triangular lattice.
Further, according to the study by the inventors of the present application, even if a hexagonal fuel assembly and a Y-shaped control rod are combined, a core structure in which the fuel rods are densely arranged in a regular triangular lattice shape can be realized. When combining a Y-shaped control rod and a hexagonal fuel assembly, there is a method in which the fuel assembly shape is a regular hexagon and a method in which one blade of the Y-shaped control rod and the fuel assembly form a regular hexagon. It is possible. The former has an advantage that the fuel assembly structure is simplified, and the latter has an advantage that the center position of the assembly in the core forms an equilateral triangle. Further, according to the study by the inventors of the present application, even if the square fuel assemblies and the cross-shaped control rods are combined, a core structure in which the fuel rods are densely arranged in a regular triangular lattice shape can be realized.

Further, according to the study by the inventors of the present invention, in the combination of the hexagonal fuel assembly and the Y-shaped control rod or the square fuel assembly and the cross-shaped control rod, the control is performed in the fuel rod facing the control rod. When the rod is pulled out, the neutron moderating effect of water increases, the neutron energy decreases, and if the enrichment level of the fissionable Pu of each fuel rod in the assembly is kept the same, the fuel rod is output to the fuel rod facing the control rod. Peaking occurs. Therefore, a fuel assembly having a flat output distribution in the fuel assembly can be realized by changing the enrichment degree of the fissile Pu in the assembly according to the distance from the control rod insertion position.

The fourth object is achieved by the following operation. According to the study by the inventors of the present application, by optimizing the fuel assembly arrangement and the orifice configuration in the core, the output and flow rate of the fuel assembly in the core can be flattened and the thermal margin can be improved. The region excluding the outermost periphery of the core is divided into two equal areas in the radial direction, and the fuel assemblies are loaded so that the average number of core assembly core stay cycles in the outer core region is smaller than that in the inner core region. As a result, the infinite neutron multiplication factor in the outer core region can be made higher than that in the inner core region, and the radial power distribution can be flattened. By loading a fuel assembly having a large stay cycle number in the outermost peripheral region of the core having a low neutron importance, it is possible to reduce the enrichment of the required fissionable Pu. According to the study by the inventors of the present application, the influence of neutron leakage from the core outer periphery is particularly large in the core outermost layer and the fuel assemblies adjacent thereto, and as a result, the fuel assembly output becomes lower than in other regions. The flow rate in the fuel assembly increases. Therefore, the outermost circumference of the core and
By setting the average value of the orifice pressure loss coefficient of the fuel assembly adjacent to it to be larger than the average value of the orifice pressure loss coefficient of the other regions, the flow rate distribution can be flattened. As a result, the flow rate near the outermost periphery of the core can be reduced, and the total core flow rate can be reduced. Further, the vapor void ratio can be increased in the region where the orifice pressure loss coefficient is increased, which can contribute to the improvement of the void coefficient and the increase of the breeding ratio.

The sixth object is achieved by the following operation. According to the study by the present inventors, it is possible to increase the breeding ratio to 1.0 or more by cooling with light steam.
Current BWR due to steam temperature exceeding saturation temperature
It is necessary to develop materials with higher resistance to high temperatures than those used in, and radioactive nuclides such as corrosion products flow out of the core together with steam. In the present invention, the steam weight ratio at the core outlet is suppressed to 40% or less, and even when the output rises due to an abnormal transient change, the coolant maintains the two-phase flow state of the saturation temperature and maintains the saturation temperature. ,
The same structural material as the existing light water reactor can be used, and the distillation function by boiling in the reactor prevents the inclusion of radioactive nuclides such as corrosion products in the steam that goes to the turbine, while maintaining a core with a breeding ratio of 1.0 or more. Can be realized.

Further, the first and the seventh are operated by the following actions.
The purpose of is achieved. According to the study by the present inventors, an example of fuel enriched with Pu in depleted uranium produced as a residue during the production of enriched uranium used in the current light water reactor is examined with the aim of long-term stable energy supply. , When there is still a large amount of natural uranium and depleted uranium recovered from spent fuel as in the present day, natural uranium and depleted uranium and low enriched uranium (0.71 wt% to 2.0 wt) are used instead of depleted uranium. %), The enrichment of fissionable Pu is reduced by about 0.5 wt% or more as compared with the case of using depleted uranium. In comparison, it is possible to realize a core having performance equal to or higher than the breeding ratio and void coefficient.

The eighth object is achieved by the following operation. According to the study by the present inventors, not only is Pu enriched in depleted uranium, but the actinide nuclide is also recycled at the same time, whereby a long-lived radionuclide is in an equilibrium state in the reactor and reaches a certain amount. Therefore, in the reactor of the present invention, the actinide nuclide balances the generation amount and the annihilation amount, the increase amount becomes zero, and the total generation amount of the long half-life actinide nuclide which is particularly problematic in radioactive waste is greatly increased. It is possible to realize a reactor system in which actinide nuclides containing Pu can be confined only in the reactor, the reprocessing facility, and the fuel manufacturing facility.

[0060]

Embodiments of the present invention will now be described in detail with reference to the drawings. In the following examples, the core having an electric output of 1.35 million kW was targeted, but the output scale is not limited to this. It can be applied to other output scales by changing the number of fuel assemblies.

(First Embodiment) A first embodiment of the present invention will be described with reference to FIGS. 1 and 6 to 11.

FIG. 1 shows the electric output of this embodiment of 1356 MW.
The horizontal section of e is shown. 720 fuel assemblies 1 and 223 Y-shaped control rods 2 are shown in an integrated ratio with three fuel assemblies. FIG. 6 shows a cross section of the fuel assembly lattice. Fuel rods 3 having a diameter of 10.1 mm are arranged in an equilateral triangle with a fuel rod gap of 1.3 mm, and a regular hexagonal fuel assembly is formed by forming a regular hexagonal fuel assembly lattice with a channel box 4 and one blade 5 of a Y-shaped control rod. The structure is such that the outermost fuel rod row on one side of the assembly does not exist. That is, in the hexagonal fuel assembly, two of the three sets of fuel rod rows parallel to the outermost fuel rod rows facing each other are equal to 17 rows, and the remaining one set is 16 rows. A stainless steel tube filled with B ₄ C was placed on each of the three blades of the control rod, and the blade intervals were 120 degrees. Three wings are arranged. That is, it is extended to the center side along one side of each wing and three sides facing the same direction around the center of the control rod (eg, the clockwise direction in FIG. 6). Three wings are arranged so that an equilateral triangle is formed by three straight lines. Also, at the tip of the control rod,
It has a follower part made of carbon, which is a substance that has a moderating power smaller than that of light water. Fig. 7 shows the fuel arrangement of the equilibrium core. The number written on the fuel assembly 1 indicates the period of stay in the core by the number of cycles. At the outermost periphery of the core, where the neutron importance is low,
Cycle fuel is loaded. The inner region outside the core is loaded with the fuel with the highest infinite neutron multiplication factor in the first cycle of the in-core stay period in order to flatten the power distribution in the radial direction of the core. The fuel in the second and third cycles of staying in the core is dispersedly loaded in the inner region of the core to flatten the power distribution in the inner region. FIG. 8 shows the state of the orifices in the equilibrium core.
The numbers marked with indicate that the opening and closing degrees of the orifices installed in the fuel support portion are different, and there are three regions. The orifice diameter of the core outer area (numbers 1 and 2) where the fuel assembly output is small is smaller than the orifice diameter of the inner area. FIG. 9 shows the distribution in the height direction of the enrichment of fissile Pu averaged in the horizontal cross section of the fuel assembly for the equilibrium core. The uranium enriched with Pu is depleted uranium. The core height is 55 cm, 8/1 from the bottom of the core
It is divided into 2 areas at 2 and the enrichment at the top is 12w.
t% and the lower part are 10 wt%. Also, 25 cm and 20 cm depleted uranium blankets are attached above and below the core, respectively. FIG. 10 shows the enrichment distribution of Pu in the horizontal cross section of the lower portion of the fuel assembly. Fissile Pu enrichment is 1
0.4wt%, 9.4wt%, 8.4wt%, 7.4w
4 kinds of t%, the average enrichment is 10 wt%. The enrichment distribution of Pu in the horizontal cross section at the top of the fuel assembly is the same as that at the bottom, and the enrichment of fissile Pu is 12.4 wt%,
Four kinds of 11.4 wt%, 10.4 wt% and 9.4 wt%, the average enrichment is 12 wt%. FIG. 11 shows the power distribution in the height direction of the core average and the void fraction distribution. Average core void ratio is 61%, steam weight ratio at core outlet is 32 wt
%.

Next, the operation of this embodiment will be described.

An effective volume ratio of water to fuel of 0.27 was achieved by the combination of a dense hexagonal fuel assembly of an equilateral triangular lattice with a fuel rod gap of 1.3 mm, a core average void fraction of 61%, and a Y-shaped control rod. , Breeding ratio in furnace 0.90, blanket breeding ratio 0.9.
11, a total growth ratio of 1.01 was achieved.

For the above reason, in this embodiment, the effective water-to-fuel volume ratio is reduced from about 2.0 of the current reactor to 0.27, thereby realizing a light water reactor of 1.01 for breeding. .

The output of this core is the same as that of the current ABWR, 1.35 million kWh, and the outer radius of the core is 2.8 m.
It is almost the same as the value of BWR. The core height is 55 cm, and blankets of 25 cm and 20 cm are attached to the top and bottom, respectively, to form a short fuel assembly. However, since the fuel rods are dense, the total length of the fuel rods is not much different from that of ABWR, MCPR is 1.32, the thermal design standard value,
Satisfies 1.24. Since the fuel was 55 cm in length, regardless of the density, the Pu inventory was
It is as small as 4.4 tons when converted to the amount of fissile Pu per 1 million kWh output, and is 10 tons or less per 1 million kWh in consideration of the period during which Pu stays outside the reactor, such as reprocessing.

For the above reasons, in the present example with a proliferation ratio of 1.01, 1 million tons of fissile Pu and 15 million tons of depleted uranium produced from the world's uranium reserves of 15 million tons were used for 1 million tons. 1,500 kWe furnaces can be operated for 10,000 years, realizing a long-term stable energy supply system.

In the present embodiment, a pressure vessel having substantially the same size as the ABWR currently under construction produces the same output, the fuel cladding material is the same zircaloy as the ABWR, and the same burnup is 45G.
Wd / t is achieved.

For the above reasons, in the present embodiment, a BWR capable of coping with a stable long-term energy supply can be realized at a power generation cost similar to that of a light water reactor that is fully fueled at present.

The BWR currently in operation has a core height of about 3
Although it is 70 cm, it is 55 cm in this embodiment. Therefore, the neutron leakage effect that makes the void coefficient negative, which represents the increase in reactivity when the amount of steam generated in the core increases, is large. Also, 1 from the top in the height direction of the fuel assembly
At 8.3 cm, the upper and lower two with different enrichment of fissile Pu
It is a regional fuel, with an upper enrichment of 12 wt% and a lower enrichment of 10 wt%. On the other hand, when the void volume of the core increases, the relative increase in void fraction is about 20% greater in the lower core with a lower void fraction than in the upper core which has already reached a saturated state, and as a result, the neutron importance is increased. Swing of the neutron flux distribution from the upper core with high neutron importance to the lower core with low neutron importance,
Negative void reactivity is injected. Further, in this embodiment, the steam weight ratio at the core outlet is 32%, all the coolant does not become steam even during an abnormal transient change, and always maintains a two-phase flow state, similar to the present BWR. The radioactive substances such as corrosion products accumulated in the core are confined in the core by the distillation action by boiling to prevent the migration to the turbine side.

For the above reasons, in this embodiment, a BWR capable of coping with a stable long-term energy supply is realized with the same level of safety as that of a light water reactor which is currently running fuel.

The BWR currently in operation is 85% of the fission reaction.
% Occurs around 0.6 eV or less in the thermal neutron region, whereas the median energy of the fission reaction of this example is about 1 keV, and the reaction ratio in the resonance region is very large. Therefore, the Doppler coefficient is 1.6 × 10 ⁻⁵ Δk / k / ° C. for the currently operating BWR, while the value of this example is 3.7 × 10 ⁻⁵ Δk / k / ° C. About twice as big.

The void coefficient of the BWR currently in operation is −
It is 7.0 × 10 ⁻⁴ Δk / k /% void, and the value of this embodiment is −0.5 × 10 ⁻⁴ k / k /% void, which is designed to have a small absolute value. As a result, the thermal margin becomes relatively large in the event that the pressure increases or the temperature of the cooling water decreases.

For the above reasons, in this embodiment, a BWR core having a larger safety margin can be realized in a considerable transient event as compared with the BWR currently in operation.

According to the present embodiment, a combination of a dense hexagonal fuel assembly, a Y-shaped control rod, and a core average void fraction of 61% resulted in an average of 10.5 wt% fissile Pu for depleted uranium.
With a fuel enriched with methane, a breeding ratio of 1.01 was achieved, and the core height was set to 55 cm, which also reduced the Pu inventory, and the world's natural uranium reserves of 15 million tons. With a BWR that can operate the base for 10,000 years, long-term stable energy supply can be achieved. Also, the BW currently in operation
By making R and the operating conditions such as the diameter and output of the pressure vessel, and the materials used almost the same, it is possible to suppress the power generation cost to the same level as the current BWR, although the performance is greatly improved. Also, by maintaining a negative void coefficient by the short fuel assemblies and upper and lower two-region fuel assemblies, and by suppressing the steam weight ratio at the core outlet to about 30%, the distillation function by boiling is maintained and the activated substances are removed. Current B such as confined in a pressure vessel
You can keep the same safety margin as WR.

In this example, the composition, action, and effect of the fuel enriched with Pu in the depleted uranium produced as the residue during the production of enriched uranium used in the current light water reactor are described with the aim of long-term stable energy supply. It was But,
Instead of depleted uranium, natural uranium, depleted uranium recovered from spent fuel, fuel enriched with Pu in low-enriched uranium, and the same or better effect can be obtained. In this case, by increasing the weight ratio of uranium-235 contained in the fuel,
The enrichment of fissile Pu can be lowered by 0.5 wt% or more as compared with the case of using depleted uranium. As a result, the multiplication ratio for fissile Pu can be increased by about 3% or more, and the void coefficient can be made more negative.
Also, because Pu inventory can be reduced, RBWR
The number of operating bases can be further increased.

Although the void coefficient is negative in this embodiment, the output coefficient including the Doppler coefficient can be negative even if the void coefficient is 0 or slightly positive. According to the study by the inventors of the present application, it is shown from the safety evaluation result that if the output coefficient is negative, the positive / negative of the void coefficient is essentially no problem. Therefore, the core is
It can be made longer than cm to further increase the thermal margin. It is also possible to increase the breeding ratio by making the fuel rod gap narrower than 1.3 mm.

In this embodiment, the fuel in which only Pu was enriched in uranium was described, but other actinide nuclides can be enriched in addition to Pu. In this case, RB
Since WR has a high average energy of neutrons, it becomes difficult for Pu to move to an actinide nuclide with a high mass number, and
Actinide nuclides can be eliminated by fission reaction.

Further, in the present embodiment, the enrichment of fissile Pu is in the upper and lower two regions at 8/12 from the lower end of the core, but it is not limited to this. FIG. 28
An example of distribution in the height direction of the fissile Pu enrichment averaged in the horizontal cross section of a fuel assembly for an equilibrium core different from that of this example is shown in FIG. The uranium enriched with Pu is the same depleted uranium as in this example, and is 25 cm and 20 cm above and below the core, respectively.
A depleted uranium blanket is attached. The height of the core is 55 cm, which is the same as that of this embodiment, and is 1 / from the lower end of the core.
It is divided into 5 areas at 12, 2/12, 7/12, and 8/12. Fissionable Pu enrichment is 1 from the top
2.5wt%, 10.5wt%, 9.5wt%, 10.5wt
%, 12.5 wt%, the average fissile Pu enrichment in the aggregate is 11 wt%, the average enrichment in the upper half is 11.7 wt%, and the average enrichment in the lower half is 10. 2 wt%
Is. While increasing the fissile Pu enrichment in the region near the lower end, add an intermediate enrichment (10.5 wt%) between the highest enrichment (12.5 wt%) and the lowest enrichment (9.5 wt%). By arranging them, as shown in FIG. 29, the output distribution in the axial direction can be made more flat. In the embodiment of FIG. 28, output peaking can be further reduced by 5% as compared with the present embodiment. Further, in the height direction of the fuel assembly, the enrichment degree of the average fissile Pu in the upper half is higher than the average value in the lower half, and the effect of reducing the void reactivity coefficient is the same as in the present embodiment. Is obtained. Furthermore, by flattening the axial power distribution, the amount of neutron leakage from above and below the core increases. As a result, the required fissile Pu enrichment is increased as compared with the present embodiment, but the effect of further reducing the void reactivity coefficient is obtained. FIG. 30 shows a case where the intermediate enrichment (10.5 wt%) is eliminated by the modification of FIG. 28. The effect of flattening the output distribution is greater in FIG. 28, but it is the same as in this embodiment.
Similar effects are obtained with different types of fissile Pu enrichment.

(Second Embodiment) A second embodiment of the present invention will be described with reference to FIGS.

FIG. 18 shows the electric output of this embodiment 1356M.
The core horizontal cross section of We is shown. 223 Y-shaped control rods 10 are shown in an integrated ratio with 720 fuel assemblies 9 and 3 fuel assemblies. FIG. 19 shows a cross section of the fuel assembly lattice. Fuel rods 3 with a diameter of 10.1 mm have a fuel rod gap of 1.
They are arranged in an equilateral triangle with 3 mm and form a regular hexagonal assembly of 10 fuel rod rows. Then, the Y-shaped control rods are arranged in an integrated ratio with the three fuel assemblies as shown in FIG. 18, and the gap between the fuel assemblies in which the control rods are not inserted is the gap between the inserted fuel assemblies. It is getting narrower. A stainless steel tube filled with B ₄ C is arranged on the blades of the control rod, and the blade intervals are 120 degrees. Further, at the tip of the control rod, there is a follower part made of carbon, which is a substance having a moderating ability smaller than that of light water. The distribution of fuel in the core, the state of the orifices, and the distribution of the fissile Pu enrichment in the height direction averaged in the horizontal cross section of the fuel assembly for the equilibrium core are shown in FIGS. Figure 9
Is the same as. FIG. 20 shows the enrichment distribution of fissile Pu in the lower horizontal cross section in the fuel assembly. The enrichment distribution of fissile Pu is symmetrical with respect to one blade of the Y-shaped control rod that is not adjacent to the regular hexagonal fuel assembly. The enrichment of fissile Pu is 10.4wt%, 9.4wt%, 8.4wt%,
Four kinds of 7.4 wt%, the average enrichment is 10 wt%. The enrichment distribution of Pu in the horizontal cross section at the top of the fuel assembly is the same as that at the bottom, and the enrichment of fissile Pu is 12.
4wt%, 11.4wt%, 10.4wt%, 9.4wt
%, And the average enrichment is 12 wt%.

In this embodiment, the fuel assembly has a regular hexagonal shape, and the number of fuel rods per fuel assembly is 1 as compared with the first embodiment.
The number of lines increases by 0, the average line power density decreases, and the heat transfer area increases, so that the thermal margin is improved. On the other hand, since the space of the Y-shaped control rod is increased outside the fuel assembly, the circumscribed radius of the core is increased as compared with the first embodiment. Also in this embodiment, the effective volume ratio of water to fuel of 0.27 is achieved by the combination of the dense hexagonal fuel assembly, the Y-shaped control rod and the core average void fraction of 61%. As a result, the core characteristics are the same as in Example 1, and the same effects are obtained.

Also in this embodiment, the same or higher effect can be obtained by using natural uranium, depleted uranium recovered from spent fuel, or fuel enriched with Pu in low enriched uranium instead of depleted uranium. In addition, other actinide nuclides can be enriched with Pu.

Further, in the present embodiment, 8 from the lower end of the core.
At / 12, the fissile Pu enrichment is in the upper and lower two regions, but is not limited to this. According to FIG. 28 or FIG. 30 which is a modification of the first embodiment, the same effect as that of the first embodiment can be obtained.

(Third Embodiment) A third embodiment of the present invention will be described with reference to FIGS.

FIG. 15 shows the electric output 1356M of this embodiment.
The horizontal cross section of the core of We is shown. 223 control rod drive mechanisms 7 are shown in which the cluster control rods inserted into 720 regular hexagonal fuel assemblies 6 and 3 fuel assemblies are operated by one control rod drive mechanism. FIG. 16 shows a horizontal cross section of the fuel assembly lattice. The fuel rods 3 having a diameter of 10.1 mm are arranged in an equilateral triangle with a fuel rod gap of 1.3 mm to form a regular hexagonal assembly of 10 fuel rod rows. Among them, guide tubes 8 in which the cluster control rods are inserted are arranged at 12 positions of the fuel rod lattice. The distribution of the fuel in the core, the state of the orifice, and the distribution of the enrichment degree of the fissile Pu in the height direction averaged in the horizontal cross section of the fuel assembly for the equilibrium core are shown in FIGS. , The same as in FIG. FIG. 17 shows the Pu enrichment distribution in the lower horizontal section in the fuel assembly. Since the moderator distribution is more uniform than in Examples 1 and 2, output peaking can be suppressed by two types of fissile Pu enrichment. The fissionable Pu enrichments of the fuel rods 1 and 2 are 9.0 wt% and 10.1 wt%, respectively. The distribution of Pu enrichment in the upper horizontal section in the fuel assembly is the same as that in the bottom, and the fissionable Pu enrichment of fuel rods 1 and 2 is 11.0 wt.
%, 12.1 wt%.

In this embodiment, since the control rods are inserted into the fuel assembly, the number of fuel rods is reduced by 2 as compared with the first embodiment, but a greater reactivity control effect can be obtained and a natural absorbent is used. The required reactivity can be controlled even by using boron. Also in this embodiment, the effective volume ratio of water to fuel of 0.27 is achieved by the combination of the dense hexagonal fuel assembly, the cluster type control rod and the core average void fraction of 61%. as a result,
The core characteristics are the same as in Example 1, and the same effects can be obtained.

Also in this embodiment, the same or higher effect can be obtained by using natural uranium, depleted uranium recovered from spent fuel, or fuel enriched with Pu in low-enriched uranium instead of depleted uranium. In addition, other actinide nuclides can be enriched with Pu.

Further, in the present embodiment, 8 from the lower end of the core.
At / 12, the fissile Pu enrichment is in the upper and lower two regions, but is not limited to this. According to FIG. 28 or FIG. 30 which is a modification of the first embodiment, the same effect as that of the first embodiment can be obtained.

(Fourth Embodiment) A fourth embodiment of the present invention will be described with reference to FIGS. The present embodiment is the same as the first embodiment.
The core performance is advanced based on the configuration of
An equivalent core can be realized based on the configuration of the second or third embodiment.

In this embodiment, the electric output is 1356 MWe,
This is the case of a core characterized by high burnup of fuel. The horizontal cross section of the core, the cross section of the fuel assembly lattice, and the orifice distribution of this embodiment are the same as those of FIG. 1, FIG. 6, and FIG. 8 of the first embodiment.
FIG. 12 shows the fuel arrangement of the equilibrium core. The number written on the fuel assembly indicates the period of stay in the core by the number of cycles. At the outermost periphery of the core, where the neutron importance is low,
The third cycle fuel, which has the longest stay in the reactor, is loaded. The inner region outside the core is loaded with the fuel with the highest infinite neutron multiplication factor in the first cycle of the in-core stay period in order to flatten the power distribution in the radial direction of the core. The fuel in the second and third cycles of staying in the core is dispersedly loaded in the inner region of the core to flatten the power distribution in the inner region. In this embodiment, since the blanket portion is provided at the central portion in the axial direction, the combustion reactivity becomes smaller than that in the first embodiment, and therefore the fuel in the third cycle is large in the central region of the core. FIG. 13 shows the distribution in the height direction of the fissile Pu enrichment averaged in the horizontal section of the equilibrium core fuel assembly. The uranium enriched with Pu is depleted uranium. The core height is 77 cm, and it is divided into three regions 1, 2, and 3 at 23 cm and 52 cm from the lower end of the core. The respective fissionable Pu enrichments are 17 wt%, 0 wt%, and 17 wt%.
%, And the average is 10.6 wt%. In addition, 25 cm and 20 cm depleted uranium blankets are attached to the upper and lower parts of the core, respectively. FIG. 14 shows the power distribution in the height direction of the core average and the void fraction distribution. Core void ratio is 60
%, And the steam weight ratio at the core outlet is 29%.

Next, the operation of this embodiment will be described.

The fuel assembly structure is the same as in Example 1,
The fuel assembly is a close-packed hexagonal fuel assembly with an equilateral triangular lattice with a fuel rod gap of 1.3 mm. The core average void fraction is 60%, and the effective volume ratio of water to fuel is 0.27 due to the combination of Y-shaped control rods.
Was achieved, the breeding ratio in the furnace was 0.87, and the blanket breeding ratio was 0.1
4, a total growth ratio of 1.01 was achieved.

In the present embodiment, in the height direction of the fuel assembly, there are upper and lower portions where the enrichment of fissile Pu is 17 wt%, and the central region between them is depleted uranium that does not contain fissile Pu. There is. When the power rises or the core coolant flow rate decreases, the steam void fraction in the core rises, but at that time, the power distribution in the upper part of the core swings to a region not containing the fissile Pu in the center. As a result, the effect of lowering the reactivity of the core, which is larger than that in Example 1, can be obtained. As a result, even if the fuel has a higher burnup than in Example 1,
The void coefficient is the same as in Example 1, -0.5 × 10 ^-4 k / k.
/% Void was possible.

In the present embodiment, a pressure vessel of approximately the same size as the ABWR currently under construction produces the same output and outputs 70 GW.
d / t is achieved.

Compared to Example 1, the core length was 77 cm, which was slightly longer, but the Pu inventory was as small as 6.2 tons in terms of the amount of fissile Pu per million kilowatt output, and Pu for reprocessing etc. 10 even when considering the period of stay outside the furnace
The amount is 10 tons or less per 100,000 kWh, and the same effect as in Example 1 is obtained.

Also in this embodiment, the same or more effect can be obtained by using natural uranium, depleted uranium recovered from spent fuel, or fuel enriched with Pu in low enriched uranium instead of depleted uranium. In addition, other actinide nuclides can be enriched with Pu. further,
In the present embodiment, in the height direction of the fuel assembly, there are upper and lower portions having the same fissile Pu enrichment, and there is a structure of depleted uranium containing no fissile Pu between them.
However, the upper and lower fissile Pu enrichments do not necessarily have to be equal. Further, in the present embodiment, the depleted uranium region is arranged slightly above the center of the core, but the present invention is not limited to this. Upper and lower fissile P
It is possible to make the axial output peaking equal by combining the enrichment degree of u and the position of the depleted uranium region.

(Fifth Embodiment) A fifth embodiment of the present invention will be described with reference to FIGS. In this embodiment, the core performance is enhanced based on the structure of the first embodiment, but an equivalent core can be realized also based on the structure of the second or third embodiment.

In this embodiment, the electric output is 1356 MWe,
This is the case of a core characterized by having a margin in the minimum limit power ratio and maximum line power density. The configuration of the core of this example viewed from a horizontal section is the same as that of the first example. FIG. 21 shows the distribution in the height direction of the fissile Pu enrichment averaged in the horizontal cross section of the fuel assembly for the fifth embodiment. The uranium enriched with Pu is depleted uranium. Core height is 91 cm
Then, from the bottom of the core, at 33 cm and 53 cm,
The fissionable Pu enrichment is divided into 3 and 2 regions, 11.7 wt%, 0 wt%, 11.7 wt%, and 9.1 wt% on average. In addition, 25 cm and 20 cm depleted uranium blankets are attached to the upper and lower ends of the core, respectively. FIG. 22 shows the power distribution in the height direction of the core average and the void fraction distribution. The average core void ratio is 57%, and the steam weight ratio at the core outlet is 26%.

The structure of the fuel assembly is the same as in Example 1,
The fuel assembly is a close-packed hexagonal fuel assembly with an equilateral triangular lattice with a fuel rod gap of 1.3 mm. The average void ratio of the core is 57%, and the effective volume ratio of water to fuel is 0.28 due to the combination of Y-shaped control rods.
Was achieved, the breeding ratio in the furnace was 0.93, and the blanket breeding ratio was 0.9.
A total growth rate of 08, 1.01, was achieved.

In the present embodiment, there are upper and lower portions where the enrichment of fissile Pu is 11.7 wt% in the height direction of the fuel assembly, and the central region between them is depleted uranium containing no fissile Pu. ing. When the power rises or the core coolant flow rate decreases, the steam void fraction in the core rises, but at that time, the power distribution in the upper part of the core swings to a region not containing the fissile Pu in the center. As a result, the effect of lowering the reactivity of the core, which is larger than that in Example 1, can be obtained. As a result, even if the core height is made higher than in Example 1, the void coefficient is the same as in Example 1, -0.5 × 10 ^-4 k / k /% void.
I was able to Further, the proliferation ratio can be increased by inflowing neutrons from the upper and lower regions containing fissile Pu into the region not containing fissile Pu in the center of the core. Therefore, even if the core flow rate is increased and the average void ratio of the core is made lower than that of the first embodiment, the multiplication ratio equal to or higher than that is obtained. Also, since the core flow rate was increased, MCP
R is 1.45, and a core having an increased thermal margin can be realized as compared with the first embodiment.

Also in this embodiment, natural uranium, depleted uranium recovered from spent fuel, or fuel enriched with Pu in low-enriched uranium, in place of depleted uranium, can achieve the same or higher effect. , Pu as well as other actinide nuclides can be enriched. Further, also in this embodiment, the enrichment levels of the upper and lower fissile Pu do not necessarily have to be equal, and the depleted uranium position in the central region of the core is not limited to this.

Compared with Example 1, the core length was slightly longer at 91 cm, but the Pu inventory was as small as 6.3 tons in terms of the amount of fissile Pu per million kilowatt output, and Pu for reprocessing etc. 10 even when considering the period of stay outside the furnace
The amount is 10 tons or less per 100,000 kWh, and the same effect as in Example 1 is obtained.

(Sixth Embodiment) A sixth embodiment of the present invention will be described with reference to FIGS. In this embodiment, the core performance is enhanced based on the configuration of the first embodiment, but a similar core can be realized based on the second or third embodiment.

In this embodiment, the electric output is 1356 MWe,
This is the case of a core that increases the Pu inventory and has characteristics as a Pu storage reactor. The configuration of the core of this example viewed from a horizontal section is the same as that of the first example. No. 6 in FIG.
Fissile P averaged in the horizontal cross section of the fuel assembly for the example of FIG.
The distribution of u enrichment in the height direction is shown. The uranium enriched with Pu is depleted uranium. The core height is 126 cm,
42 cm, 82 cm from the bottom of the core, 1, 2,
The fissionable Pu enrichment is divided into three regions of 3 and 11.7 wt%, 0 wt%, 11.7 wt%, and 8.0 wt% on average. In addition, 25 cm and 20 cm depleted uranium blankets are attached to the upper and lower ends of the core, respectively. FIG. 24 shows the power distribution in the height direction of the core average and the void fraction distribution. The average core void ratio is 60%, and the steam weight ratio at the core outlet is 31%.

The fuel assembly configuration is the same as in Example 1,
The fuel assembly is a close-packed hexagonal fuel assembly with an equilateral triangular lattice with a fuel rod gap of 1.3 mm. The core average void fraction is 60%, and the effective volume ratio of water to fuel is 0.27 due to the combination of Y-shaped control rods.
Was achieved, the breeding ratio in the furnace was 0.95, and the blanket breeding ratio was 0.9.
A total growth rate of 07, 1.02, was achieved.

In the present embodiment, there are upper and lower portions where the enrichment of fissile Pu is 11.7 wt% in the height direction of the fuel assembly, and the central region between them is depleted uranium containing no fissile Pu. As compared with Examples 4 and 5, the depleted uranium region in the central portion of the core was increased to 40 cm, and the effect of reducing the void coefficient and the effect of increasing the breeding ratio could be increased.
As a result, the area containing fissile Pu can be increased, and the Pu inventory can be increased to 10.3 tons.

Although the Pu inventory is slightly reduced, in the present embodiment, instead of depleted uranium, natural uranium, depleted uranium recovered from spent fuel, and fuel enriched with Pu in low enriched uranium are also used. Can be used. In addition, other actinide nuclides can be enriched with Pu.

Further, in this embodiment, there are upper and lower portions in the height direction of the fuel assembly having the same fissile Pu enrichment, and there is a structure of depleted uranium containing no fissile Pu between them. . However, the upper and lower fissile P
The enrichments of u do not necessarily have to be equal. Also,
In the present embodiment, the region of depleted uranium is arranged slightly above the center of the core, but the present invention is not limited to this.

(Seventh Embodiment) A seventh embodiment of the present invention will be described with reference to FIGS. In this embodiment, the core performance is enhanced based on the configuration of the first embodiment, but a similar core can be realized based on the second or third embodiment.

This example has an electric output of 1356 MWe,
This is the case of a core characterized by a large negative void coefficient. The configuration of the core of this example viewed from a horizontal section is the same as that of the first example. FIG. 25 shows the distribution in the height direction of the fissile Pu enrichment averaged in the horizontal cross section of the fuel assembly for the seventh embodiment. The uranium enriched with Pu is depleted uranium. The core height is 65 cm, and it is 23c from the bottom of the core.
At m, 38 cm, it is divided into three regions 1, 2, and 3, and each of them has a fissile Pu enrichment of 13.5 wt.
%, 0 wt%, 13.5 wt%, and 10.5 wt% on average. The upper and lower ends of the core are 25 cm and 20 cm, respectively.
A depleted uranium blanket is attached. FIG. 26
Figure 2 shows the power distribution in the height direction of the core average and the void fraction distribution. The average core void ratio is 60%, and the steam weight ratio at the core outlet is 29%.

The fuel assembly structure is the same as in Example 1,
The fuel assembly is a close-packed hexagonal fuel assembly with an equilateral triangular lattice with a fuel rod gap of 1.3 mm. The core average void fraction is 60%, and the effective volume ratio of water to fuel is 0.27 due to the combination of Y-shaped control rods.
Was achieved, the breeding ratio in the furnace was 0.90, and the blanket breeding ratio was 0.1
2, a total growth ratio of 1.02 was achieved.

In the present embodiment, the fissile Pu enrichment of 13.5 wt% is at the top and bottom in the height direction of the fuel assembly, and the central region between them is depleted uranium containing no fissile Pu. Compared to Examples 5 and 6, the regions of fissile Pu above and below the core are reduced, and the effect of reducing the void coefficient due to the shorter length is further added. As a result, the void coefficient could be set to −1.8 × 10 ⁻⁴ k / k /% void. This enables output control and reactivity control by flow rate control.

Also in this embodiment, instead of depleted uranium, it is possible to use Pu enriched fuel for natural uranium, depleted uranium recovered from spent fuel, and low enriched uranium. In addition, other actinide nuclides can be enriched with Pu.

Further, in this embodiment, there are upper and lower portions in the height direction of the fuel assembly having the same fissile Pu enrichment, and there is a structure of depleted uranium containing no fissile Pu between them. . However, the upper and lower fissile P
The enrichments of u do not necessarily have to be equal. FIG. 31 shows a modification of this example with respect to the enrichment of fissile Pu and the position of the depleted uranium region. Also in FIG. 31, the output peaking in the axial direction can be made equal to that in this embodiment.

Further, FIG. 32 shows an example of distribution in the height direction of the fissile Pu enrichment averaged in the horizontal cross section of the fuel assembly different from that of this embodiment. The uranium enriched with Pu is depleted uranium as in this example, and the core height is 6
Blankets of depleted uranium of 5 cm and 25 cm and 20 cm, respectively, are attached to the upper and lower ends of the core. At 20 cm and 35 cm from the lower end of the core, the fissionable Pu enrichment was divided into three regions 1, 2, and 3.
5wt%, 0wt%, 13.5wt%, 10.5wt on average
%. When the region of depleted uranium that does not contain fissile Pu moves to the lower part of the core further than in this example, the axial power distribution has an upper peak, as shown in FIG. Therefore, the effect of making the void reactivity coefficient more negative can be obtained.

(Eighth Embodiment) In this embodiment, the actinide nuclide extracted from the spent fuel in the seventh embodiment is recycled at the same time as Pu.

The structure of the core of the present embodiment viewed from the horizontal cross section is the same as that of the first embodiment. Also in the fuel assembly for this example, the distribution in the height direction of the fissile Pu enrichment averaged in the horizontal section is the same as that in example 7. The uranium enriched with Pu is uranium taken out of the spent fuel at the same time as plutonium, and the actinide nuclide taken out of the spent fuel is also added at the same time. Fissionable Pu enrichment is 13.5wt%, 0wt%, 13.5w
t%, and 10.5 wt% on average. In addition, 25 cm and 20 cm depleted uranium blankets are attached to the upper and lower parts of the core, respectively.

The fuel assembly structure is the same as in Example 1,
The fuel assembly is a close-packed hexagonal fuel assembly with an equilateral triangular lattice with a fuel rod gap of 1.3 mm. The core average void fraction is 61%, and the effective volume ratio of water to fuel is 0.27 due to the combination of Y-shaped control rods.
Was achieved and the breeding ratio in the furnace was 0.91, and the blanket breeding ratio was 0.1.
A total growth rate of 0, 1.01, was achieved.

In the present embodiment, the fissile Pu enrichment of 13.5 wt% is located above and below the height direction of the fuel assembly, and the central region between them is depleted uranium containing no fissile Pu. Compared to Examples 5 and 6, the regions of fissile Pu above and below the core are reduced, and the effect of reducing the void coefficient due to the shorter length is further added. As a result, the void coefficient can be made negative even if the actinide nuclide extracted from the spent fuel is recycled at the same time as Pu.

Further, by repeatedly recycling the actinide nuclide extracted from the spent fuel at the same time as Pu, the long-lived radionuclide is in an equilibrium state in the furnace and reaches a certain amount. Therefore, in this example, the amount of actinide nuclide generation and the amount of annihilation are balanced, the increase is zero, and the total amount of generation of long-lived actinide nuclides, which is particularly problematic in radioactive waste, is significantly increased. To reduce the actinide nuclide containing Pu,
It can be locked only in reprocessing facilities and fuel manufacturing facilities.

(Ninth Embodiment) In this embodiment, the present invention will be described.
This is the case when applied to WR.

In the present embodiment, the same cluster type control rod as that of the third embodiment and a diameter larger than the outer diameter of the fuel rod of the current PWR 14.
The core was composed of regular hexagonal fuel assemblies in which 3 mm were densely arranged in a regular triangular lattice with a fuel rod gap of 1.0 mm. Figure 3
4 shows the distribution in the height direction of the fissile Pu enrichment averaged in the horizontal cross section of the fuel assembly for the ninth embodiment. The uranium enriched with Pu is depleted uranium. The core height is 50
In cm, the fissile Pu enrichment is uniform at 10.5 wt%. A 30 cm depleted uranium blanket is attached to each of the upper and lower ends of the core.

In a dense hexagonal fuel assembly with an equilateral triangular lattice having a fuel rod gap of 1.0 mm, a water-fuel volume ratio of 0.44 was achieved by combining a large-diameter fuel rod and a cluster-type control rod. As a result, a breeding ratio of 0.90, a blanket breeding ratio of 0.11, and a total breeding ratio of 1.01 were realized.

In this embodiment, the fissile Pu enrichment is uniform in the height direction of the fuel assembly. However, the fuel composition is not limited to this.
Modifications of this example for the axial enrichment distribution of fissile Pu are shown in FIGS. 35 and 36. In FIG. 35, the fissionable Pu enrichment in the upper and lower ends of the core is higher than that in the central region in order to make the axial power peaking more flat. In FIG. 36, the core height is 65 cm, and 25, 40 cm from the bottom of the core, 1, 2, 3
It is divided into regions, and each fissile Pu enrichment is 1
3.0wt%, 0wt%, 13.0wt%, 10w on average
It is t%. A 30 cm depleted uranium blanket is attached to each of the upper and lower ends of the core.
Since the region of depleted uranium that does not contain fissile Pu is in the central part of the core, the effect of making the input reactivity more negative at the time of void generation can be obtained.

In this embodiment, the outer diameter of the fuel rod is 14.3 mm, which is larger than the outer diameter of the fuel rod of the current PWR.
A WR of 9.5 mm can also be used. In this case, a water-to-fuel volume ratio of 0.58 is achieved by combining cluster-type control rods in a regular hexagonal close-packed hexagonal fuel assembly with a fuel rod gap of 1.0 mm. Due to the increase in the volume ratio of water to fuel, the fissile Pu enrichment was 0.5 w compared to the present example.
Although it is necessary to increase it by about t%, a growth ratio of 1.0 can be realized.

Also in this embodiment, instead of depleted uranium, Pu enriched fuel can be used for natural uranium, depleted uranium recovered from spent fuel, and low enriched uranium. In addition, other actinide nuclides can be enriched with Pu.

(Tenth Embodiment) This embodiment is a case where the present invention is applied to the same square fuel assembly as the current BWR.

The structure of the fuel assembly of this embodiment is shown in FIG. In order to reduce the volume ratio of water to fuel, 13.8 mm, which is larger than the outer diameter of the fuel rod of the current BWR, was densely arranged in a regular triangular lattice with a fuel rod gap of 1.0 mm. The number of fuel rods per fuel assembly is 85. The moderator outside the channel box 11 is a follower part installed at the upper end of the cross-shaped control rod, and a gap water region on the opposite side where the control rod is not inserted is a water-excluding plate made of a substance having a small moderating ability like the follower part. Has been eliminated by. As a result, an effective water-to-fuel volume ratio of 0.53 is achieved, and a breeding ratio of 1.0 can be realized by the fuel composition in the core height direction similar to the other embodiments. In order to make the fuel rod output peaking in the fuel assembly flat, in this embodiment, the enrichment of fissionable Pu in the fuel rod facing the channel box is made lower than that in the fuel rods in other regions.

Also in this embodiment, instead of depleted uranium, Pu enriched fuel can be used for natural uranium, depleted uranium recovered from spent fuel, and low enriched uranium. In addition, other actinide nuclides can be enriched with Pu.

[0131]

According to the present invention, a fuel in which Pu is added to depleted uranium, natural uranium, depleted uranium or low-enriched uranium can achieve a growth ratio of around 1.0 or at least 1.0.
Using Pu as a catalyst, depleted uranium, natural uranium, depleted uranium and low-enriched uranium can be burned, which can contribute to long-term stable energy supply.

Further, a dense hexagonal fuel assembly or a square fuel assembly composed of a fuel obtained by adding Pu to depleted uranium, natural uranium, depleted uranium or low enriched uranium, 45% to 7%
By combining 0% high void fraction coolant and cluster type, Y-shaped or cruciform control rods and setting an effective water to fuel volume ratio of 0.1 to 0.6, the breeding ratio of 1. Near zero or more than 1.0 can be achieved, and it can contribute to long-term stable energy supply.

Further, the same output is produced by a pressure vessel having the same diameter as that of the ABWR currently under construction, and the height of the core is 40 to 1
By making the fuel assembly as short as 40 cm, the inventory of Pu can be reduced, and Pu generated from spent fuel of LWR under the limited natural uranium reserves in the world.
Thus, many furnaces of the present invention can be operated, which can contribute to long-term stable energy supply.

Further, by making the diameter of the BWR, which is currently in operation, the pressure vessel, the operating conditions such as the output, and the materials used almost the same, the power generation cost can be reduced even though the performance is greatly improved. Can be suppressed to the same level as.

Further, the short fuel assembly, the upper and lower two-region fuel assembly, and the axially inhomogeneous fuel assembly increase the neutron leakage in the vertical direction of the core, and utilize the swing of the power distribution in the vertical direction of the core. A void coefficient core can be realized,
It can have the same level of safety as existing fuel-burning light water reactors.

Further, the combination of the dense fuel assembly and the high void fraction coolant increases the proportion of neutrons in the resonance energy region, increases the Doppler effect, and decreases the absolute value of the negative void coefficient. The safety of power increase event, pressurization event, coolant void rate decrease event, etc. is improved.

Further, the steam weight ratio of the core outlet of the coolant is set to 4
Since it was kept to 0% or less, radioactive materials such as corrosion products accumulated in the furnace can be confined in the furnace by maintaining the distillation function by boiling, and the turbine has the same level as the BWR currently in operation. The radiation level on the side can be maintained, and the radiation level can be significantly reduced compared to the conventional steam-cooled fast reactor, which is the concept of the breeder reactor.

Further, by the core composed of the hexagonal fuel assembly and the cluster type control rods inserted therein,
The homogeneity of the core is increased and the thermal margin is increased.

Further, the core of the hexagonal fuel assemblies and the Y-shaped control rods inserted between the assemblies allows the existing BWR technology of inserting from the lower part of the core to be utilized as it is.

The core system of the current BWR can be utilized as it is by the core composed of the square fuel assemblies and the cross-shaped control rods inserted between the assemblies.

Further, in a regular hexagonal, eclipse hexagonal fuel assembly or square fuel assembly, a multi-region ranging from a region near the Y-shaped or cross-shaped control rod to a region distant from the control rod, particularly 2 to Fissile Pu for 5 regions
By using 2 to 5 types of fuel rods with varying enrichment, the output peaking in the fuel assembly is reduced and the thermal margin is increased.

Further, the power density of the core is set to 100 to 300 k.
By increasing to W / l, the amount of Pu inventory per unit output is reduced, the capacity of the power generation equipment of the present invention that can operate for a certain Pu is increased, and it contributes to a long-term stable supply of energy.

Further, since the portion where the average enrichment of fissile Pu in the horizontal section in the height direction of the fuel assembly is 6 wt% or more is between 40 and 140 cm, the Pu inventory amount per unit output is reduced. However, the power generation facility capacity of the present invention that can be operated for a certain Pu is increased, contributing to a long-term stable supply of energy, and the neutron leakage effect in the core height direction when the steam generation amount is increased is increased. , Increase the negative void coefficient and contribute to safety.

Further, since the average value of the lower half of the fuel assembly using the blanket portions at the upper and lower ends is lower than the average value of the fissile Pu enrichment in the upper half of the fuel assembly, the power in the core height direction is reduced. As the distribution is flattened and the thermal margin increases,
When the steam generation amount increases, the swing of the power distribution in the core height direction works, and the negative void reactivity coefficient increases, which contributes to safety.

Further, in the height direction excluding the blanket portions at the upper and lower ends of the fuel assembly, there are upper and lower portions where the enrichment degree of fissionable Pu is 6 wt% or more, and the fissionable Pu enrichment in the region near the center between them. By setting the degree of conversion to 6% or less, the power of the reactor is increased and the negative void coefficient due to the power distribution swing in the vertical direction of the core when the amount of steam in the core is increased becomes large, and the safety is improved. Further, due to the neutron absorption effect in the region near the center in the axial direction of the core, the Pu inventory that can be loaded in the core becomes large, and the function as the Pu storage reactor is improved. Further, the core height becomes relatively high and the overall length of the fuel rod becomes long, so that the thermal margin for the maximum linear power density also becomes large.

Further, by simultaneously recycling Pu and uranium, the effect of preventing nuclear non-proliferation becomes large.

By simultaneously recycling Pu, uranium and actinide nuclides, the amount of actinide nuclides generated and the amount of actinoid nuclides are balanced to make the amount of increase zero, and this is a particular problem in radioactive waste. The long-lived actinide nuclide can be confined only in the reactor, reprocessing facility, and fuel manufacturing facility, which improves the environmental characteristics.

[Brief description of drawings]

FIG. 1 is a horizontal sectional view of a core according to a first embodiment of the present invention.

FIG. 2 is a characteristic diagram showing a relationship between a fuel rod lattice constant necessary to represent a conversion ratio and an effective water-to-fuel volume ratio.

FIG. 3 is a characteristic diagram showing a relationship between a fuel rod gap and a geometrical water-to-fuel volume ratio.

FIG. 4 is a characteristic diagram showing the relationship between the fissile Pu enrichment and the growth ratio.

FIG. 5 is an explanatory diagram showing a change in output distribution when the vapor void ratio increases.

6 is a horizontal cross-sectional view of a fuel assembly loaded in the core of FIG.

FIG. 7 is a fuel assembly layout diagram at the time of the equilibrium core of the embodiment of FIG.

8 is a distribution chart of the orifices in the embodiment of FIG.

9 is an axial enrichment distribution map of the fuel assemblies loaded in the core of FIG.

10 is a fuel rod enrichment distribution map in the fuel assembly loaded in the core of FIG. 1. FIG.

11 is a characteristic diagram showing the axial output of the core and the void fraction distribution in the embodiment of FIG.

FIG. 12 is a fuel assembly layout diagram at the time of an equilibrium core in a fourth embodiment of the present invention.

13 is an axial enrichment distribution map of the fuel assemblies loaded in the core of FIG.

FIG. 14 is an axial output of the core in the embodiment of FIG.
And a characteristic diagram showing a void ratio distribution.

FIG. 15 is a horizontal sectional view of a core according to a third embodiment of the present invention.

16 is a horizontal sectional view of a fuel assembly loaded in the core of FIG.

17 is a fuel rod enrichment distribution map in the fuel assembly loaded in the core of FIG. 15. FIG.

FIG. 18 is a horizontal sectional view of a reactor core according to a second embodiment of the present invention.

FIG. 19 is a horizontal sectional view of a fuel assembly loaded in the core of FIG.

20 is a fuel rod enrichment distribution map in the fuel assembly loaded in the core of FIG. 18. FIG.

FIG. 21 is an axial enrichment distribution map of the fuel assemblies loaded in the core of the fifth embodiment of the present invention.

FIG. 22 is a characteristic diagram showing the axial output and void fraction distribution of the core in the fifth example of the present invention.

FIG. 23 is an axial enrichment distribution map of fuel assemblies loaded in the core of the sixth embodiment of the present invention.

FIG. 24 is a characteristic diagram showing the axial output and void fraction distribution of the core in the sixth example of the present invention.

FIG. 25 is an axial enrichment distribution map of the fuel assemblies loaded in the core of the seventh embodiment of the present invention.

FIG. 26 is a characteristic diagram showing axial power and void fraction distribution of a core in a seventh example of the present invention.

FIG. 27 is a characteristic diagram showing the relationship between the average void ratio of the core and the weight ratio at the core outlet.

28 is a modification of the axial enrichment distribution of the fuel assemblies loaded in the core of FIG. 1. FIG.

29 is a characteristic diagram showing the axial output and void fraction distribution of the core loaded with FIG. 28.

30 is a modification of the axial enrichment distribution of the fuel assemblies loaded in the core of FIG.

FIG. 31 is a modification of the axial enrichment distribution of the fuel assemblies loaded in the core of the seventh embodiment.

FIG. 32 is a modification of the axial enrichment distribution of the fuel assemblies loaded in the core of the seventh embodiment.

33 is a characteristic diagram showing the axial output and void fraction distribution of the core loaded with FIG. 32.

FIG. 34 is an axial enrichment distribution map of the fuel assemblies loaded in the core of the ninth embodiment.

FIG. 35 is a modification of the axial enrichment distribution of the fuel assemblies loaded in the core of the ninth embodiment.

FIG. 36 is a modification of the axial enrichment distribution of the fuel assemblies loaded in the core of the ninth embodiment.

FIG. 37 is a horizontal sectional view of a fuel assembly according to a tenth embodiment of the present invention.

[Explanation of symbols]

1, 6, 9 ... Fuel assembly, 2, 10 ... Y-shaped control rod, 3
... fuel rods, 4, 11 ... channel box, 5 ... 1 blade of Y-shaped control rod, 7 ... control rod drive mechanism, 8 ... guide tube.

   ─────────────────────────────────────────────────── ─── Continued front page    (72) Inventor Moriwaki Honest             2-12-1 Omika-cho, Hitachi-shi, Ibaraki Prefecture             Ceremony Hitachi Energy Research Institute

Claims (24)

[Claims] 1. A fissile plutonium average of a horizontal cross section of a fuel assembly in a core height direction excluding blanket portions at both upper and lower ends of a light water reactor core containing depleted uranium and fuel enriched with plutonium or plutonium and actinide nuclides. The void coefficient is made negative by making the fissile plutonium enrichment in the upper core higher than the fissile plutonium enrichment in the lower core by setting the enrichment of 6 wt% or more between 40 cm and 140 cm. The breeding ratio is set to 1. by setting the core average void fraction to 45 to 70% when operating at 50% or more of the output.
A light water reactor core characterized by being set to near 0 or 1.0 or more. 2. A light water reactor core comprising plutonium or a fuel enriched with plutonium and an actinide nuclide in uranium containing at least one of natural uranium, depleted uranium, and low enriched uranium, and the core height excluding blanket parts at both upper and lower ends. In the vertical direction, the fissionable plutonium enrichment in the upper part of the core is set to be between 40 cm and 140 cm for the part where the average enrichment of fissile plutonium in the horizontal cross section of the fuel assembly is 6 wt% or more. The void ratio is made negative by making the ratio higher than 50 degrees, and the void ratio of the core average when operating at 50% or more of the rated output is set to 45 to 70%.
A light water reactor core characterized by being set to near 0 or 1.0 or more. 3. The growth ratio according to claim 1 or 2, wherein the growth ratio is 1.0.
To 1.15 in the light water reactor core. 4. The light water reactor core according to claim 1, wherein the average fissionable plutonium enrichment in the core portion excluding the outer peripheral portion and the blanket portions at the upper and lower ends is 6 to 20 wt%. 5. The fuel rod according to any one of claims 1 to 4, wherein the fuel rods are composed of hexagonal fuel assemblies densely arranged in an equilateral triangular lattice and cluster type control rods inserted therein. Is a light water reactor core. 6. The hexagonal fuel assembly according to claim 1, wherein the fuel rods are densely arranged in an equilateral triangular lattice, and three fuel rods inserted between them are 120 degrees apart. A light water reactor core comprising a Y-shaped control rod having a blade. 7. The hexagonal fuel assembly according to claim 6, wherein the blades of the Y-shaped control rods adjacent to the hexagonal fuel assembly are two or less, respectively, and the blades are not inserted between the fuel assemblies. , A light water reactor core characterized in that the blades are narrower than the gap between the fuel assemblies into which they are inserted. 8. The square fuel assembly according to claim 1, wherein the fuel rods are densely arranged in an equilateral triangular lattice, and four blades inserted between them are 90 degrees apart. A light water reactor core characterized by being composed of a cross-shaped control rod having a wing. 9. The light water reactor core according to claim 5, wherein the control rod has a follower portion made of a substance having a moderating power smaller than that of light water at a tip portion thereof. 10. The fuel rod according to claim 6, wherein at least a fuel rod having the lowest fissile plutonium enrichment in the fuel assembly is arranged at a position adjacent to the control rod in the fuel assembly. A light water reactor core characterized by 11. A light water reactor core according to any one of claims 1 to 10, characterized in that the average power density of the core portion excluding the outer peripheral portion and the blanket portions of the upper and lower end portions is 100 to 300 kW / l. 12. A light water reactor according to any one of claims 1 to 11, wherein the core outlet steam weight ratio of the coolant when operating at 50% or more of the rated output is 20 wt% to 40 wt%. Core. 13. A region according to any one of claims 1 to 12, wherein a region of the core excluding the outermost periphery is divided into two equal areas in the radial direction,
A light water reactor core, characterized in that the fuel assembly is loaded so that the average number of core stay cycles of the fuel assembly loaded in the outer core area is smaller than that in the inner core area. 14. The average value of the orifice pressure loss coefficient of the outermost periphery of the core and the fuel assemblies adjacent thereto is larger than the average value of the orifice pressure loss coefficients of the other fuel assemblies. A light water reactor core characterized by being set so that 15. In a fuel assembly having depleted uranium and a fuel enriched with plutonium or plutonium and an actinide nuclide, the fission property of the horizontal cross section of the fuel assembly in the height direction of the fuel assembly excluding the blanket parts at the upper and lower ends. Plutonium average enrichment of 6 wt% or more is 40 cm to 140 cm
, The average value of fissile plutonium enrichment in the upper half of the fuel assembly is lower than the average value of fissile plutonium enrichment in the lower half, and the effective water-to-fuel volume ratio is from 0.1 to 0.1. By setting the ratio between 6 and 6, the growth ratio is around 1.0 or 1.
A fuel assembly characterized by being set to 0 or more. 16. A fuel assembly having plutonium or a fuel enriched with plutonium and an actinide nuclide in uranium containing at least one of natural uranium, depleted uranium, and low enriched uranium, except for the blanket parts at the upper and lower ends. 40 to 140 cm in the height direction of the fuel assembly where the average fissile plutonium enrichment in the horizontal cross section of the fuel assembly is 6 wt% or more
The average value of fissile plutonium enrichment in the lower half is lower than the average value of fissile plutonium enrichment in the upper half of the fuel assembly, and the effective water-to-fuel volume ratio is 0.1 to 0.1. The fuel assembly is characterized in that the growth ratio is in the vicinity of 1.0 or at least 1.0 by setting the ratio to be 6 or less. 17. The fuel assembly according to claim 15, wherein the fuel assembly is a hexagonal fuel assembly or a square fuel assembly in which the fuel rods are densely arranged in an equilateral triangular lattice, and the gap between the fuel rods is 0.7 to 2. A fuel assembly characterized by being 0.0 mm. 18. The fuel assembly according to claim 15, wherein the average fissile plutonium enrichment in the fuel region excluding the blanket portions at the upper and lower ends is 6 to 20 wt%. 19. The method according to any one of claims 15 to 18,
The fuel assembly is a hexagonal fuel assembly in which the fuel rods are arranged in an equilateral triangular lattice, and among the three fuel rod arrays parallel to the facing outermost fuel array, two fuel rod arrays are provided. A fuel assembly having the same number and one row more than the number of the remaining set of fuel rods. 20. A Y-shaped control rod having three blades inserted between fuel assemblies loaded in a LWR core, wherein:
The distance between the three blades is 120 degrees, respectively, and one side surface of each of the blades is the center side along the three side surfaces facing in the same direction around the center of the control rod. A control rod in which the three blades are arranged so that an equilateral triangle is formed by three straight lines extended to the. 21. Depleted uranium, natural uranium, depleted uranium,
A light water reactor core comprising a hexagonal fuel assembly having plutonium or a fuel enriched with plutonium and an actinide nuclide in uranium containing at least one of low-enriched uranium, and a Y-shaped control rod, wherein the Y-shaped control rod 21. The control rod according to claim 20, wherein in the hexagonal fuel assembly, the fuel rods are arranged in an equilateral triangle lattice shape, and two of the three fuel rod rows parallel to the facing outermost fuel rod row are arranged. A light water reactor core, characterized in that the number of fuel rod rows in a set is equal and one row is larger than the number in the remaining set of fuel rods. 22. The blades of the Y-shaped control rod according to claim 21, wherein the blades of the Y-shaped control rod are arranged parallel to the remaining set of fuel rods of the hexagonal fuel assembly adjacent to the Y-shaped control rod. A light water reactor core characterized by being 23. The light water reactor core according to claim 21 or 22, wherein the effective water-to-fuel volume ratio is between 0.1 and 0.6. 24. The light water reactor core according to claim 21 or 22, wherein the average void fraction of the core when operating at 50% or more of the rated output is 45 to 70%.