fast reactor core

Abstract

To provide a reactor core of a fast reactor, which flattens an output distribution and increases an outlet temperature of coolant while suppressing the deterioration in reactor performance, so as to be able to achieve a sodium-cooled metal fuel fast reactor with high compatibility with a molten salt thermal storage system.SOLUTION: A reactor core 1 of a fast reactor is such a fuel assembly that fuel rods each formed by storing hollow fuels with a predetermined enrichment degree within a range of 11 to 13 wt.% in cladding tubes are densely arranged in a wrapper tube. A first fuel assembly 2 having the fuel rods with the hollow fuels of a large hollow diameter is loaded at the center of the core and a second fuel assembly 3 having the fuel rods with the hollow fuels having a hollow diameter smaller than that of the hollow fuels of the first fuel assembly 2 is loaded at a periphery of the core.SELECTED DRAWING: Figure 1

Description

[Technical Field]

[0001] The present invention relates to a reactor core for a sodium-cooled metal-fueled fast reactor that increases the coolant outlet temperature of the reactor, thereby increasing its applicability to a heat storage system using molten salt. [Background technology]

[0002] Regarding the fuel assemblies and core of a fast reactor, generally speaking, a fast breeder reactor has the core located inside the reactor vessel, and the reactor vessel is filled with liquid sodium, which is the coolant. The fuel assemblies loaded into the core consist of multiple fuel rods containing plutonium-enriched depleted uranium (U-238), a trumpet tube surrounding the bundled fuel rods, entrance nozzles supporting the lower ends of these fuel rods and the neutron shield located below the fuel rods, and a coolant outflow section located above the fuel rods.

[0003] The core of a fast breeder reactor has a core fuel region comprising an inner core region and an outer core region surrounding the inner core region, a blanket fuel region surrounding the core fuel region, and a shielding region surrounding the blanket region. In the case of a typical homogeneous core, the Pu enrichment of fuel assemblies loaded in the outer core region is higher than that of fuel assemblies loaded in the inner core region. As a result, the power distribution in the radial direction of the core is flattened.

[0004] The nuclear fuel materials housed in each fuel rod of a fuel assembly can take the form of metallic fuel, nitride fuel, or oxide fuel. Of these, oxide fuel has the most extensive track record. Mixed oxide fuel, i.e., MOX fuel pellets, which are a mixture of oxides of Pu and depleted uranium, are packed into the fuel rod to a height of approximately 80-100 cm in the axial center. Furthermore, within the fuel rod, axial blanket regions filled with multiple uranium dioxide pellets made from depleted uranium are positioned above and below the MOX fuel packing region, respectively. The inner core fuel assemblies loaded into the inner core region and the outer core fuel assemblies loaded into the outer core region each have multiple fuel rods filled with multiple MOX fuel pellets in this manner. The Pu enrichment level of the outer core fuel assemblies is higher than that of the inner core fuel assemblies.

[0005] The blanket fuel region surrounding the core fuel region is loaded with blanket fuel assemblies, each containing multiple fuel rods filled with multiple uranium dioxide pellets made from depleted uranium. Of the neutrons generated by the nuclear fission reaction occurring in the fuel assemblies loaded in the core fuel region, those that leak out of the core fuel region are absorbed by U-238 in each fuel rod of the blanket fuel assemblies loaded in the blanket fuel region. As a result, the fissile nuclide Pu-239 is newly generated in each fuel rod of the blanket fuel assembly.

[0006] Furthermore, control rods are used during startup, shutdown, and adjustment of reactor power in fast breeder reactors. Each control rod consists of multiple neutron-absorbing rods, each containing boron carbide (B4C) pellets sealed in a stainless steel cladding tube. These neutron-absorbing rods are housed in a trumpet-shaped tube with a hexagonal cross-section, similar to the inner and outer core fuel assemblies. The control rod system consists of two independent systems: a main reactor shutdown system and a backup reactor shutdown system. This allows for emergency shutdown of the fast breeder reactor using either the main reactor shutdown system or the backup reactor shutdown system alone.

[0007] To achieve carbon neutrality by 2050, nuclear power generation is required to be adaptable to load fluctuations associated with the large-scale introduction of renewable energy. In the United States, a plant has been proposed that can cope with load fluctuations by combining a small sodium-cooled metal-fuel fast reactor with a thermal storage system using molten salt, which has a proven track record in solar power generation. From the perspective of ensuring the integrity of the metal fuel, metal-fuel fast reactors are often designed to have a primary coolant outlet temperature about 50°C lower than oxide-fuel fast reactors, and in the case of a small sodium-cooled metal-fuel fast reactor, which is the main target of this invention, a temperature of about 500°C is assumed. On the other hand, in the case of nitrate-based molten salts used in the thermal storage system with a proven track record in solar power generation mentioned above, it is desirable to set the reactor coolant outlet temperature to about 540°C to 550°C, considering the melting point of the molten salt and the temperature conditions of the high-temperature tank on the thermal storage system.

[0008] To improve the coolant outlet temperature of a sodium-cooled metal-fuel fast reactor, it is necessary to flatten the power distribution, suppress wasted flow, and reduce the coolant flow rate. Patent Document 1 describes a method to flatten the power distribution by making the Pu enrichment level of all core fuels the same, and making the Zr content of the inner core higher than that of the outer core metal fuel, U-Pu-Zr, which has a large neutron leakage. [Prior art documents] [Patent Documents]

[0009] [Patent Document 1] Japanese Patent Publication No. 2005-83966 [Overview of the Initiative] [Problems that the invention aims to solve]

[0010] However, in the metal fuel fast reactor core described in Patent Document 1, where the Pu enrichment level of all core fuels is the same, if the Zr content of the metal fuel in the inner core is made higher than that of the outer core, the amount of heavy metals (U and Pu) loaded into the inner core decreases, which reduces the amount of fuel loaded and leads to problems such as a decrease in core performance, including the breeding ratio and burnup reactivity. Therefore, the present invention provides a fast reactor core that can realize a sodium-cooled metal-fueled fast reactor with high compatibility with molten salt heat storage systems by suppressing deterioration of core performance, flattening the power distribution, and increasing the coolant outlet temperature. [Means for solving the problem]

[0011] To solve the above problems, the reactor core of the fast reactor according to the present invention is Hollow fuel is housed inside the cladding tube. The fuel rods are packed tightly inside the horn tube. to arrangement did fuel assembly The core of a fast reactor loaded with And, Loaded towards the center of the reactor core First fuel assembly The hollow diameter of the hollow fuel contained in the reactor core is loaded onto the outside of the reactor core. Second fuel assembly The hollow diameter of the hollow fuel contained in the first fuel assembly is larger than that of the hollow fuel contained in the second fuel assembly, and the Pu enrichment of the hollow fuel contained in the first fuel assembly and the Pu enrichment of the hollow fuel contained in the second fuel assembly are set to the same value within the range of 11 to 13 wt%, and the burnup dependence of the neutron infinite multiplication factor on the fuel volume ratio of the first fuel assembly and the burnup dependence of the neutron infinite multiplication factor on the fuel volume ratio of the second fuel assembly are made the same, thereby flattening the power distribution in the radial direction of the core throughout the combustion cycle. It is characterized by the following. [Effects of the Invention]

[0012] According to the present invention, it is possible to provide a fast reactor core that can realize a sodium-cooled metal fuel fast reactor with high compatibility with molten salt heat storage systems by suppressing deterioration of core performance, flattening the power distribution, and increasing the coolant outlet temperature. For example, by using hollow fuel with a constant Pu enrichment level of 11-13 wt% in the fuel assemblies of a fast reactor core, loading fuel assemblies with a large hollow diameter towards the center of the core, and loading fuel assemblies with a small hollow diameter towards the periphery of the core, it is possible to realize a sodium-cooled metal-fueled fast reactor core that suppresses spatial and temporal fluctuations in the power distribution without degrading core performance, eliminates wasted flow, raises the reactor coolant outlet temperature, and is highly compatible with molten salt thermal storage systems. Other issues, configurations, and effects not mentioned above will be clarified by the following description of the embodiments. [Brief explanation of the drawing]

[0013] [Figure 1]It is a horizontal cross-sectional view of the core fuel assembly of a fast reactor according to Example 1 of the present invention. (a) is a horizontal cross-section of the inner core fuel assembly, (b) is a horizontal cross-section of the outer core fuel assembly, and (c) is a horizontal cross-sectional view of a 1 / 2 core of a fast reactor loading the inner core fuel assembly and the outer core fuel assembly. [Figure 2] It is a longitudinal cross-sectional view of the inner core fuel assembly and the outer core fuel assembly shown in FIG. 1. [Figure 3] It is a diagram showing the burnup dependence of the infinite multiplication factor of the core fuel assembly of a fast reactor with the Pu enrichment as a parameter. [Figure 4] It is a diagram showing the Pu enrichment dependence of the maximum reactivity change during the burnup period. [Figure 5] It is a diagram showing the burnup dependence of the infinite multiplication factor of a metal fuel assembly with the fuel volume ratio as a parameter. [Figure 6] It is a longitudinal cross-sectional view of the core in a fast reactor according to Example 2 of the present invention. [Figure 7] It is a longitudinal cross-sectional view of the inner core fuel assembly and the outer core fuel assembly shown in FIG. 6. [Figure 8] It is a longitudinal cross-sectional view of the inner core fuel assembly and the outer core fuel assembly according to Example 3 of the present invention. [Figure 9] It is a longitudinal cross-sectional view of the core in a fast reactor loaded with the inner core fuel assembly and the outer core fuel assembly shown in FIG. 8.

Embodiments for Carrying Out the Invention

[0014] Hereinafter, embodiments of the present invention will be described with reference to the drawings.

Examples

[0015] This embodiment will be explained using Figure 1, which shows a horizontal cross-section of the core fuel assembly and the 1 / 2 core in the fast reactor according to this embodiment; Figure 2, which shows a vertical cross-section of the core fuel assembly; Figure 3, which shows the burnup change of the neutron infinite multiplication factor of the core fuel assembly with Pu enrichment as a parameter; Figure 4, which shows the Pu enrichment dependence of the change in maximum reactivity during the burnup period of the core fuel assembly; Figure 5, which compares the burnup changes of the neutron infinite multiplication factor of the inner core fuel assembly and the outer core fuel assembly; and Table 1, which shows the specifications of the core fuel assembly.

[0016] This embodiment focuses on a fuel assembly for a sodium-cooled metal fuel fast reactor, which uses hollow metal fuel to achieve fuel swelling absorption by reducing the fuel smear density to 75% or less of that of conventional metal fuel, while also enabling He bonding by reducing the gap between the fuel alloy and the fuel cladding tube to a level comparable to that of a MOX fuel core, and the fast reactor core into which it is loaded.

[0017] Figure 1(a) shows a horizontal cross-sectional view of the inner core fuel assembly according to this embodiment, Figure 1(b) shows a horizontal cross-sectional view of the outer core fuel assembly, and Figure 1(c) shows a horizontal cross-sectional view of the 1 / 2 core of a fast reactor loaded with these assemblies. As shown in Figure 1, the inner core fuel assembly 2 consists of fuel rods (not shown) containing hollow U-Pu-Zr alloy 7 arranged in a triangular pitch and densely packed inside a hexagonal stainless steel trumpet tube 9. The regions 10 between the fuel rods 7 inside the trumpet tube 9 (regions through which the sodium coolant flows) are filled with sodium coolant flowing from the bottom to the upstream of the fuel assembly. For example, the pitch of the fuel assembly is 157.2 mm, the diameter of the fuel rod is 8.5 mm, and the diameter of the hollow is 2.82 mm. Although simplified in Figure 1, there are 217 fuel rods in one fuel assembly. The volume ratio of fuel to the inner core fuel assembly 2 (including the gaps between fuel assemblies) is 30.0%. On the other hand, the outer core fuel assembly 3 differs from the inner core fuel assembly 2 in that the diameter of the hollow U-Pu-Zr alloy (hollow metallic fuel of the outer core fuel assembly) 8 is narrow at 2.27 mm. As a result, the volume ratio of fuel in the outer core fuel assembly 3 is large at 33.6%.

[0018] The vertical structure of the fuel assembly will be explained. Figure 2 is a longitudinal cross-sectional view of the inner core fuel assembly and the outer core fuel assembly shown in Figure 1. As shown in Figure 2, the fuel rods 110 loaded in the inner core fuel assembly 2 are housed inside a stainless steel circular cladding tube, and consist of cylindrical U-Pu-Zr fuel alloy (metal fuel of the inner core fuel assembly) 113 with a hollow (hollow of the metal fuel of the inner core fuel assembly) 114. These are placed on a metal fuel support member 115 installed above a gas plenum 116 that holds gaseous fission products (FP), and the upper end plug 111 and lower end plug 112 are welded together and sealed with helium (He) gas. The longitudinal length of the U-Pu-Zr alloy is 100 cm. The outer core fuel assembly 3 has a similar structure and dimensions, but as mentioned above, it differs in that the diameter of the hollow 119 of the U-Pu-Zr fuel alloy (metallic fuel of the outer core fuel assembly) 118 is smaller than the diameter of the hollow 114 of the U-Pu-Zr fuel alloy (metallic fuel of the inner core fuel assembly) 113 of the inner core fuel assembly 2.

[0019] In a fuel assembly of a fast reactor loaded with metallic fuel U-Pu-Zr, the infinite neutron multiplication factor (k) is given by the Pu enrichment level. ∞ Figure 3 shows the plot of the dependence curve of ) on burnup (GWd / t) calculated using a fast reactor analysis method. In Figure 3, the horizontal axis is burnup (GWd / t) and the vertical axis is neutron infinite multiplication factor (k) ∞ The curves shown are for the neutron infinite multiplication factor (k) at a Pu enrichment of 18 wt%, 24 at a Pu enrichment of 15 wt%, 25 at a Pu enrichment of 12 wt%, 26 at a Pu enrichment of 9 wt%, and 27 at a Pu enrichment of 6 wt%. From Figure 3, it can be seen that when the Pu enrichment is high, the initial neutron infinite multiplication factor (k) is high. ∞ Although the neutron infinite multiplication factor (k) is large, the conversion ratio is small and the consumption of Pu exceeds the generation, so the neutron infinite multiplication factor (k) is associated with combustion. ∞ It can be seen that the rate of decrease of ) is large. Conversely, when the Pu enrichment level is low, the conversion ratio is large and Pu production exceeds consumption, so the initial infinite neutron multiplication factor (k ∞ ) is small, but the neutron infinite multiplication factor (k) occurs with combustion.∞ ) It can be seen that the increase rate is large. Based on FIG. 3, FIG. 4 shows a diagram that arranges the Pu enrichment dependence of the change in the maximum reactivity during the combustion period of the fuel assembly, that is, a diagram showing the Pu enrichment dependence of the change in the maximum reactivity during the combustion period. From FIG. 4, taking the reactivity 1 $(= effective delayed neutron fraction, defined as about 0.3% in the case of a fast reactor using Pu as fuel) as the reference limit value, the range of Pu enrichment with a small change in the maximum reactivity below it is from 11 wt% to 13 wt%. In this embodiment, within this range, the specifications of the fuel assembly and the number of fuel assemblies loaded in the core are set so that the Pu enrichment is particularly critical at 12 wt%. In addition, it is necessary to make the output of the outer core fuel assembly, where the neutron leakage is large, closer to the output of the inner core fuel assembly on the core center side. In the conventional core design of a fast reactor, by making the Pu enrichment of the outer core fuel assembly higher than that of the inner core fuel assembly, the flattening of the output distribution in the core radius direction is achieved. However, as shown in FIG. 3, when the Pu enrichment changes, the burnup dependence of the infinite multiplication factor (k ∞ ) of the fuel assembly varies greatly, so it is difficult to maintain the radial output flattening throughout the combustion cycle. Therefore, in this embodiment, as shown in FIG. 5, the Pu enrichment of the metal fuel U-Pu-Zr alloy is kept constant at 12 wt%, and by making the fuel volume ratio of the outer core fuel assembly higher than that of the inner core fuel assembly, the infinite multiplication factor (k ∞ )45 for the fuel volume ratio of the inner core fuel and the infinite multiplication factor (k ∞ )43 for the fuel volume ratio of the outer core fuel assembly have the same burnup dependence, and by maintaining the radial output sharing flattening throughout the combustion cycle, the waste flow rate is reduced and an increase in the coolant outlet temperature of the nuclear reactor is realized. The fuel volume ratios of the inner core fuel assembly and the outer core fuel assembly are realized by setting the hollow diameter of the metal fuel U-Pu-Zr alloy to be large for the inner core fuel assembly and small for the outer core fuel assembly, as shown in Table 1. Note that the infinite multiplication factor (k ∞ )44 in FIG. 5 shows the burnup dependence of the average infinite multiplication factor of the core.

[0020] [Table 1]

[0021] In this embodiment, under conditions of a reactor electrical output of 300 MW, a thermal output of 714 MW, and an average burnup of approximately 100 GWd / t of core fuel removal, core calculations confirmed that loading core fuel assemblies with the specifications shown in Table 1 flattens the radial power distribution and minimizes temporal power fluctuations throughout the burnup cycle, thereby reducing wasted flow and raising the outlet temperature of the reactor coolant from approximately 500°C to approximately 550°C.

[0022] As a result, the suitability for heat storage systems using molten salt can be improved, and by raising the outlet temperature of the reactor coolant by approximately 50°C, thermal efficiency can be increased, resulting in improved economic performance.

[0023] As described above, this embodiment makes it possible to provide a fast reactor core that can realize a sodium-cooled metal fuel fast reactor with high compatibility with molten salt heat storage systems by suppressing deterioration of core performance, flattening the power distribution, and increasing the coolant outlet temperature.

[0024] Furthermore, by using hollow fuel with a constant Pu enrichment level of 11-13 wt% in the core fuel assemblies of a fast reactor, loading fuel assemblies with a large hollow diameter towards the center of the core, and loading fuel assemblies with a small hollow diameter towards the periphery of the core, it is possible to realize a sodium-cooled metal-fuel fast reactor core that suppresses spatial and temporal fluctuations in the power distribution without degrading core performance, eliminates wasted flow, raises the reactor coolant outlet temperature, and is highly compatible with molten salt thermal storage systems. [Examples]

[0025] Figure 6 is a longitudinal cross-sectional view of the core of a fast reactor according to Embodiment 2 of the present invention, and Figure 7 is a longitudinal cross-sectional view of the inner core fuel assembly and outer core fuel assembly shown in Figure 6. This embodiment differs from Embodiment 1 in that a sodium plenum, consisting of a trumpet tube and liquid sodium, is installed above the fuel rods containing the hollow metal fuel U-Pu-Zr in the inner core fuel assembly and outer core fuel assembly.

[0026] As shown in Figure 7, the inner core fuel assembly 51 differs from the core fuel assembly structure of Example 1 in that a sodium plenum 601, composed of a trumpet tube 9 and liquid sodium, is installed on top of the fuel rods 62 that house the hollow metallic fuel U-Pu-Zr (metallic fuel of the inner core fuel assembly) 66, which is the same as shown in Figure 2 of Example 1 described above. Furthermore, the vertical length of the hollow metallic fuel U-Pu-Zr alloy (metallic fuel of the outer core fuel assembly) 603 housed in the fuel rods 602 of the outer core fuel assembly 52 is longer than that of the hollow metallic fuel U-Pu-Zr alloy 66 of the inner core fuel assembly, and the height of the sodium plenum 606 is shortened by this length.

[0027] The horizontal cross-sectional layout of the reactor core is the same as in Figure 1(c) of the above-described embodiment 1. The vertical cross-sectional view of the reactor core is as shown in Figure 6, where the height of the inner core region 53 loaded with the inner core fuel assemblies 51 is lower than that of the outer core region 54 loaded with the outer core fuel assemblies 52, and conversely, the sodium plenum 56 is thicker in the inner core region and thinner in the outer core region. Since the sodium plenum 56 functions as a neutron reflector during steady-state operation, the reactor core performance is not impaired, and the same spatial and temporal power leveling effect as in the above-described embodiment 1 is achieved. Therefore, the effect of raising the coolant outlet temperature can also be achieved in this embodiment.

[0028] In an Unticipated Loss of Flow (ULOF) event, which simulates a scram failure in a fast reactor, the coolant temperature at the top of the fuel region of the core fuel assembly rises first, decreasing the sodium density. This increases the amount of neutron leakage into and above the sodium plenum at the top of the core fuel, resulting in a large negative reactivity being applied, thus suppressing increases in reactivity and power. In this embodiment, the core fuel in the inner core region, which contributes greatly to void reactivity, has a low reactivity, and the absolute value of the negative reactivity applied increases. As a result, the net reactivity becomes negative, avoiding boiling of the coolant sodium during ULOF and improving inherent safety.

[0029] As described above, according to this embodiment, in addition to the effects of Example 1, boiling of the sodium coolant during ULOF can be avoided, and the inherent safety has been improved. [Examples]

[0030] Figure 8 is a longitudinal cross-sectional view of the inner core fuel assembly and outer core fuel assembly according to Embodiment 3 of the present invention, and Figure 9 is a longitudinal cross-sectional view of the core of a fast reactor in which the inner core fuel assembly and outer core fuel assembly shown in Figure 8 are loaded. This embodiment differs from Embodiment 1 in that the gap between the hollow metal fuel and the cladding is set to be wide and is immersed in liquid bond Na to improve gap conductance.

[0031] As shown in Figure 8, the fuel rods 71 ​​of the inner core fuel assembly 70 contain hollow metallic fuel U-Pu-Zr alloy 75, similar to that in Example 1, housed in stainless steel cladding tubes 74 and sealed by upper and lower end plugs 72 and 73. The difference from the metallic fuel of Example 1 described above is that the gap between the hollow metallic fuel U-Pu-Zr alloy 75 and the cladding tubes 74 is set to be wider, and it is immersed in liquid bond Na to improve gap conductance. The outer core fuel assembly 78 has a similar structure, but the difference from the inner core fuel assembly 70 is that, similar to Example 1, the diameter of the hollow 702 of the hollow metallic fuel alloy 701 in the outer core fuel assembly is smaller than the diameter of the hollow 76 of the hollow metallic fuel alloy 75 in the inner core fuel assembly. The fuel volume ratios in the inner core fuel assembly 70 and the outer core fuel assembly 78 are the same as those shown in Table 1 of Example 1 described above.

[0032] The longitudinal cross-sectional view of the reactor core is shown in Figure 9, with the inner core fuel assembly 70 shown in Figure 8 loaded into the inner core region 81 and the outer core fuel assembly 78 loaded into the outer core region 82. Unlike Examples 1 and 2, the gas plenum region 83 is located above the core fuel region. Another difference from Example 2 is that the height of the core fuel is the same in the inner core region and the outer core region.

[0033] In this embodiment, the metal fuel is housed in a cladding tube while immersed in liquid Bond Na, which has high thermal conductivity. The temperature of the metal fuel during steady-state operation is lower than in Examples 1 and 2 described above, and it follows the temperature of the coolant during transients. In particular, when the coolant temperature rises during ULOF, a large negative Doppler reaction can be expected to be applied, improving inherent safety.

[0034] As described above, according to this embodiment, in addition to the effects of Example 1, if the coolant temperature rises during ULOF, a large negative Doppler reactivity can be expected to be applied, making it possible to improve inherent safety.

[0035] In Examples 1 to 3 described above, sodium was used as the coolant, but similar effects can be achieved with lead or lead-bismuth. Furthermore, while a metallic fuel U-Pu-Zr alloy was used as the fuel, similar effects can be obtained with MOX fuel or nitride fuel. Moreover, similar effects can be obtained with any combination of the above-mentioned coolants and fuels.

[0036] It should be noted that the present invention is not limited to the embodiments described above, and various modifications are included. For example, the embodiments described above are described in detail to make the present invention easier to understand, and are not necessarily limited to those having all the configurations described. Furthermore, it is possible to replace parts of the configuration of one embodiment with the configuration of another embodiment, and it is also possible to add the configuration of another embodiment to the configuration of one embodiment. [Explanation of symbols]

[0037] 1…Half of a fast breeder reactor core 2…Inner core fuel assembly 3…Outer core fuel assembly 4…Radial blanket fuel assembly 5…Shield aggregate 6…Control rod assembly 7…Hollow metal fuel in the inner core fuel assembly 8…Hollow metal fuel in the outer core fuel assembly 9... Trumpet pipe 10…Area through which the sodium coolant flows 23…Neutron infinite multiplication factor for Pu enrichment of 18 wt% 24…Neutron infinite multiplication factor for Pu enrichment of 15 wt% 25…Neutron infinite multiplication factor for Pu enrichment of 12 wt% 26…Neutron infinite multiplication factor for Pu enrichment of 9 wt% 27…Neutron infinite multiplication factor for Pu enrichment of 6 wt% 43... Neutron infinite multiplication factor relative to the fuel volume ratio of the outer core fuel assembly 44…Neutron infinite multiplication factor relative to the average fuel volume ratio of the core 45... Infinite neutron multiplication factor relative to the fuel volume ratio of the inner core fuel 51, 70… Inner core fuel assemblies 52,78…Outer core fuel assembly 53,81…Inner core region 54,82...Outer core area 55,84...Shield aggregate 56,601,606… Sodium plenum 57, 69, 77, 83, 116, 605… Gasplenum 58…center 62, 71, 110… Fuel rods of the inner core fuel assembly 63, 72, 111… Upper end plug 64, 73, 112… Lower end plug 65,74…cladding tube 66, 75, 113... Metal fuel in the inner core fuel assemblies 67, 76, 114… Hollows in the metal fuel of the inner core fuel assemblies 68,115…Metal fuel support member 79,117,602… Fuel rods of the outer core fuel assembly 118,603,701... Metal fuel of the outer core fuel assembly 119,604,702... Hollow metal fuel in the outer core fuel assembly

Claims

1. A fast reactor core loaded with fuel assemblies in which fuel rods, each containing hollow fuel contained within a cladding tube, are densely arranged within a trumpet tube, A fast reactor core characterized in that the hollow diameter of the hollow fuel contained in a first fuel assembly loaded towards the core center is larger than the hollow diameter of the hollow fuel contained in a second fuel assembly loaded towards the outside of the core, the Pu enrichment of the hollow fuel contained in the first fuel assembly and the Pu enrichment of the hollow fuel contained in the second fuel assembly are set to the same value within the range of 11 to 13 wt%, and the burnup dependence of the neutron infinite multiplication factor with respect to the fuel volume ratio of the first fuel assembly and the burnup dependence of the neutron infinite multiplication factor with respect to the fuel volume ratio of the second fuel assembly are the same, thereby flattening the power distribution in the radial direction of the core throughout the burnup cycle.

2. In the core of the fast reactor according to claim 1, A fast reactor core characterized in that the hollow fuel is a U-Pu-Zr metallic fuel alloy.

3. In the core of the fast reactor according to claim 1, The core of a fast reactor is characterized in that the upper part of the fuel rod has a sodium plenum composed of a trumpet tube and liquid sodium, the hollow fuel of the first fuel assembly is a hollow U-Pu-Zr metallic fuel alloy and the length of the hollow fuel is shorter than the length of the hollow fuel of the second fuel assembly, and the height of the sodium plenum of the first fuel assembly is longer than the height of the sodium plenum of the second fuel assembly.

4. In the core of the fast reactor according to claim 2, The core of a fast reactor is characterized in that it has a sodium plenum composed of a trumpet tube and liquid sodium at the top of the fuel rod, the length of the hollow U-Pu-Zr metallic fuel alloy of the first fuel assembly is shorter than the length of the hollow U-Pu-Zr metallic fuel alloy of the second fuel assembly, and the height of the sodium plenum of the first fuel assembly is longer than the height of the sodium plenum of the second fuel assembly.

5. In the core of the fast reactor according to claim 3, The core of a fast reactor is characterized in that the sum of the length of the hollow U-Pu-Zr metal fuel alloy and the height of the sodium plenum is the same for the first fuel assembly and the second fuel assembly.

6. In the core of the fast reactor according to claim 4, The core of a fast reactor is characterized in that the sum of the length of the hollow U-Pu-Zr metal fuel alloy and the height of the sodium plenum is the same for the first fuel assembly and the second fuel assembly.

7. In the core of the fast reactor according to claim 1, The core of a fast reactor is characterized in that the hollow fuel is a hollow U-Pu-Zr metallic fuel alloy, and the hollow U-Pu-Zr metallic fuel alloy is immersed in bonded sodium as a fuel rod.

8. In the core of the fast reactor according to claim 2, The core of a fast reactor is characterized by being a fuel rod in which the hollow U-Pu-Zr metal fuel alloy is immersed in bonded sodium.

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