JP2013050366A - Fast reactor core - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a fast reactor core capable of reducing the unit cost of power generation.SOLUTION: A fast reactor core 1 includes a Minor Actinide fuel region (MA fuel region) 2 and a parent material region 3 positioned above the MA fuel region 2. The MA fuel region 2 is loaded with a fissile Minor Actinide (MA) and the parent material region 3 is loaded with a parent material fuel which is a nuclear fuel material such as natural uranium. During the initial operation of the fast reactor, fission chain reactions of MA in the MA fuel region 2 occur, which generates heat in the MA fuel region 2 and power generation is continued. A portion of the neutrons generated by the nuclear fission in the MA fuel region 2 reaches the parent material region 3 and irradiates the parent material fuel. The parent material fuel irradiated with the neutrons in the parent material region 3 produces a new fissile material, which starts fission chain reactions, and the region with the chain reactions moves up toward the upper end of the parent material region 3.

Description

  The present invention relates to a fast reactor core, and more particularly, to a fast reactor core using a long-life minor actinide (MA) obtained by reprocessing spent nuclear fuel.

  A fuel assembly and a core of a fast reactor (FR) are described in Naohiro Hirakawa and Tomohiko Iwasaki, “Introduction to Reactor Physics”, Tohoku University Press, October 30, 2003, p279-286. That is, in general, a fuel assembly loaded in the core of a fast reactor is composed of a plurality of fuel rods configured by enclosing depleted uranium (U-238) enriched in plutonium (Pu) in a cladding tube. Fuel rod bundle bundled by a spacer, a wrapper tube surrounding the fuel rod side bundle, a coolant outflow portion disposed above the fuel rod bundle, and a neutron shield and coolant inflow portion disposed below the fuel rod bundle (Entrance nozzle). The core of the fast reactor is loaded with the plurality of fuel assemblies described above. A standard homogeneous core in a fast reactor forms a core fuel region having an inner core fuel region and an outer core fuel region surrounding the inner core fuel region, and the Pu in the fuel assembly loaded in the outer core fuel region. The power distribution in the radial direction of the core fuel region is flattened by making the average enrichment higher than the average enrichment of Pu in the fuel assembly loaded in the inner core fuel region.

Metals, nitrides, oxides, and the like have been studied as forms of nuclear fuel materials loaded on the fuel assemblies, but oxide fuels have the most extensive results. In particular, a mixed oxide (MOX fuel) in which Pu oxide and deteriorated U oxide are mixed is used as the oxide fuel, and a plurality of fuel pellets manufactured with the MOX fuel are divided into the inner core fuel region and the outer core. The fuel rods of the respective fuel assemblies loaded in the fuel region are filled. Within these fuel rods, a fuel region filled with a plurality of fuel pellets made of MOX fuel is formed at a height of about 80 to 100 cm at the center, and the deterioration U is located above and below the fuel region. An axial blanket region filled with a plurality of fuel pellets of UO ₂ fuel, which is an oxide, is formed. The core fuel assembly loaded in the core fuel region is formed by bundling a plurality of fuel rods each having an axial blanket region formed above and below the fuel region and the fuel region, respectively, and having a regular hexagonal cross section. It is housed in a cylindrical trumpet tube.

A radial blanket region surrounds the outer core fuel region. A plurality of blanket fuel assemblies having fuel rods filled with a plurality of UO ₂ fuel pellets produced in UO ₂ fuel is an oxide of depleted uranium is loaded radially blanket region. The blanket fuel assembly also houses a plurality of bundled fuel rods in a cylindrical trumpet tube having a regular hexagonal cross section. In the radial blanket region and the axial blanket region, of the neutrons generated by the fission reaction of fissile material (for example, Pu-239) in the core fuel region, neutrons leaking from the core fuel region are absorbed by U-238. As a result, Pu-239, which is a fissile nuclide, is generated and contributes to the proliferation of Pu in the entire core (growth ratio> 1.0).

Control rods are used when the reactor power is adjusted during startup and shutdown of the fast reactor and even during operation of the fast reactor. The control rod is a bundle of a plurality of neutron absorber rods made by sealing pellets made of boron carbide (B ₄ C) into a stainless steel cladding tube, and the cross section of the control rod is regular, similar to the core fuel assembly. It is stored in a square trumpet tube. The control rod used in the fast reactor has a configuration of two independent systems of a main furnace stop system control rod and a post-furnace stop system control rod. The fast reactor can be stopped urgently by either one of the main furnace stop system control rod and the after-furnace stop system control rod.

  Usually, a radial blanket fuel assembly configured using deteriorated uranium (U) fuel is loaded at the boundary between the core fuel assembly and the radiation shield on the outermost periphery of the core. In this region, a part of the neutrons generated by the nuclear fission in the core fuel region leaks out, and Pu is generated by the (n, γ) reaction in which U-238 occupying most of the degradation U absorbs the neutrons. That is, the main function of the blanket region is fuel growth.

On the other hand, only a small amount of enriched uranium, which is the initial type of fire, generates and continues the fission chain reaction of Na-cooled fast reactors using degraded U or natural U as nuclear fuel, without generating fuel changes during the life of the core. the idea of continuing the TWR (Traveling Wave Reactor) is, Kevan D. Weaver et. al. , "a ONCE-THROUGH FUEL CYCLE fOR FAST REACTORS" Kevan D. Weaver et. al. Proceedings of the 17 th International Conference on Nuclear It is described in Engineering, ICONE17, July 12-16, 2009, paper number 75381. As shown in FIG. 2 of this document, TWR fuel rods and fuel assemblies are described in Naohiro Hirakawa and Tomohiko Iwasaki, “Introduction to Reactor Physics”, Tohoku University Press, pp. 279-286, similar to the fast reactor described on October 30, 2003, the distance between the fuel rods to the 8.8 mm diameter of the fuel rods is as small as 0.8 mm, and the volume ratio of the nuclear fuel to the coolant is large and dense. It is characterized by the fact that a grid is used. The core of TWR is Kevan D. Weaver et. Al., “A ONCE-THROUGH FUEL CYCLE FOR FAST REACTORS” Kevan D. Weaver et. Al. Proceedings of the 17 ^th International Conference on Nuclear Engineering, ICONE17, July 12-16 , 2009, paper number 75381, as shown in FIG. 3, the combustion region in which hexagonal fuel assemblies are arranged in a rectangular parallelepiped shape and the fission chain reaction occurs is shown in FIG. Proceed from one end to the other.

  A reflector-controlled fast reactor is described in JP-A-2005-337898. In this fast reactor, a core is disposed in a reactor vessel filled with a liquid metal as a coolant, and a neutron shield is disposed around the core. The neutron shield is installed in the reactor vessel. A neutron reflector that is moved up and down by the drive device moves between the core and the neutron shield in the axial direction of the core. A neutron absorber (or neutron transmitting material) having a neutron reflection capability inferior to that of the coolant is disposed in the upper region of the neutron shield. A reactor core that is divided into two regions in the axial direction, the proportion of Np-238, which is a minor actinide in the fuel, in the lower region is 3.0% by weight, and the proportion is zero in the upper region is disclosed in JP-A-2005-337898 It is described in FIG. 12 (sixth embodiment) of the publication. Furthermore, Japanese Patent Application Laid-Open No. 2005-337898 has the same effect as that of the fifth embodiment with respect to the sixth embodiment, in which the neutron capture cross section of the minor actinide is large and the ratio of fissile material is reduced. It describes that it has the same effect.

JP 2005-337898 A

Naohiro Hirakawa and Tomohiko Iwasaki, “Introduction to Reactor Physics”, Tohoku University Press, pp. 279-286, October 30, 2003 Kevan D. Weaver et. Al., “A ONCE-THROUGH FUEL CYCLE FOR FAST REACTORS” Kevan D. Weaver et. Al. Proceedings of the 17th International Conference on Nuclear Engineering, ICONE17, July 12-16, 2009, paper number 75381

  The fast reactors and TWRs that have been constructed and proposed as well as light water reactors, both in the maintenance of the fission chain reaction, that is, in the continuation of power generation, contain concentrated fissile material such as U-235 or Pu-239. is necessary. The core described in FIG. 12 (sixth embodiment) of Japanese Patent Application Laid-Open No. 2005-337898 is a lower region in which the ratio of Np-238, which is a minor actinide in the fuel, is 3.0 wt%, and the ratio It has an upper region where is zero. In this sixth embodiment, it is not clear what the fuel in the lower region contains in addition to the minor actinides, but JP-A-2005-337898 states that "the mixing ratio of minor actinides contained in the fuel" Based on the description that the minor actinide has a large neutron capture cross section and has the same effect as reducing the ratio of fissile material, the ratio of Np-238 is 3.0% by weight. It is clear that fissile material (U-235, Pu-239, etc.) is contained in the lower region and the upper region where the ratio is zero.

  If such a fissile material is unnecessary, a reactor that can maintain the fission chain reaction during the life of the nuclear reactor, that is, continue power generation, using the deteriorated U as a nuclear fuel, a uranium enrichment facility and uranium mining will become unnecessary, and as a weapon Further, it is not necessary to continuously supply usable Pu.

  The objective of this invention is providing the core of a fast reactor which can reduce a power generation unit price.

  A feature of the present invention that achieves the above-described object is that a first fuel region containing only a minor actinide as a nuclear fuel material, and a second fuel region disposed adjacent to the first region and containing a parent material fuel that is a nuclear fuel material. It is in having.

  The first fuel region containing only the minor actinide as the nuclear fuel material is disposed adjacent to the second fuel region containing the parent material fuel, which is the nuclear fuel material, so that it occurred due to the nuclear fission of the minor actinide in the first fuel region. A part of the neutron is irradiated to the parent material fuel in the second fuel region, and a new fissile material is generated from the parent material fuel. Even if fission by this fissile material occurs in the second fuel region and the minor actinide in the first fuel region disappears, the fission in the second fuel region is continued, and this fission region is The second fuel region moves toward the end opposite to the end on the first fuel region side. Since the fissionable material generated from the parent material fuel in the second fuel region is not filled with the fissile material other than the minor actinide in the first fuel region, the fissionable material generated from the parent material fuel in the second fuel region is continued. Power generation can be continued without supplying a long-term supply of nuclear fuel material. For this reason, the power generation unit price in the fast reactor can be reduced.

  In the present invention, a nuclear reactor capable of supplying power for a long time without the need for fissile nuclides (U-235, Pu-239, etc.) can be realized, so that extraction of Pu by reprocessing, uranium enrichment facility, and uranium mining are not required. Since the entire fuel cycle can be greatly simplified, the unit price of nuclear power generation can be reduced.

It is a longitudinal cross-sectional view of the core of the fast reactor of Example 1, which is a preferred embodiment of the present invention. It is a cross-sectional view of the core of the fast reactor shown in FIG. FIG. 3 is a cross-sectional view of the fuel assembly loaded in the core fuel region shown in FIG. 2. It is a characteristic view which shows the relationship between the neutron energy and neutron cross section of americium 241 (Am-241). FIG. 2 is a characteristic diagram showing a combustion change at a neutron infinite multiplication factor in a parent material fuel region in the core of the fast reactor shown in FIG. 1. It is a longitudinal cross-sectional view of the core of the fast reactor of Example 2 which is another Example of this invention. It is a cross-sectional view of the core of the fast reactor shown in FIG. It is a cross-sectional view of the MA fuel assembly loaded in the MA fuel region shown in FIG.

  The core of the fast reactor according to embodiment 1, which is a preferred embodiment of the present invention, will be described below with reference to FIGS.

  The core 1 of the fast reactor of this embodiment has a minor actinide region (hereinafter referred to as MA fuel region) 2 and a parent material region 3. The parent material region 3 is disposed above the MA fuel region 2 and is adjacent to the MA fuel region 2. The cross section of the MA fuel region 2 (cross section AA in FIG. 1) and the cross section of the parent material region 3 (cross section BB of FIG. 1) have the same configuration shown in FIG.

  In the MA fuel region 2 and the parent material region 3 of the core 1, the neutron reflector region 5 surrounds the core fuel region 4, and the neutron shield region 6 surrounds the neutron reflector region 5. A plurality of fuel assemblies 7 are loaded in the core fuel region 4, a plurality of neutron reflector assemblies 10 are loaded in the neutron reflector region 5, and a plurality of neutron shield assemblies 11 are loaded in the neutron shield region 6. It is loaded. Each of the fuel assembly 7, the neutron reflector assembly 10, and the neutron shield assembly 11 has a regular hexagonal cross section.

  As shown in FIG. 3, the fuel assembly 7 includes a plurality of fuel rods 8 and a trumpet tube 9. The plurality of fuel rods 8 are bundled with a plurality of fuel spacers (not shown) arranged in the axial direction of the fuel assembly 7 and arranged in the trumpet tube 9. Although not shown in the figure, the fuel assembly 7 is a coolant outlet located above the fuel rod bundle bundled by the fuel spacer, and a neutron shield and coolant inlet located below the fuel rod bundle. Part (entrance nozzle). Each fuel rod 8 of the fuel assembly 7 is filled with a metal of a minor actinide (hereinafter referred to as MA), which is a long-lived nuclear fuel material obtained by reprocessing spent nuclear fuel, in a lower region corresponding to the MA fuel region 2. The upper region corresponding to the parent material region 3 is filled with the metal of natural uranium, degraded uranium or recovered uranium, which is the parent material, as the nuclear fuel material. The MA metal filled in the fuel rods 8 includes fissile MA. There are Np-237, Am-241, Am-243, and Cm-137 as minor actinides filled in the fuel rod 8. As MA, Np-237, Am-241, Am-243, Cm-137 and the like may be used alone, or at least two of these MAs may be mixed and used. In this embodiment, MA-10% Zr is filled in the MA fuel region 2 as MA. MA filled in the MA fuel region 2 does not contain fissile Pu. By loading a plurality of fuel assemblies 7 into the core fuel region 4, the MA fuel region 2 containing MA at the bottom is formed in the core fuel region 4, and the parent material region 3 containing the parent fuel material is formed at the top. The

  The neutron reflector assembly 10 is made of stainless steel and has a regular hexagonal cross section as described above. The neutron shield assembly 11 is a boron carbide shield assembly having a regular hexagonal cross section made of stainless steel and boron carbide. The neutron reflector assembly 10 and the neutron shield assembly 11 are disposed in the MA fuel region 2 and the parent material region 3, respectively, and have a length from the lower end of the MA fuel region 2 to the upper end of the parent material region 3. .

The fuel volume ratio in each of the MA fuel region 2 and the parent material region 3 is 50%. In the MA fuel region 2, by setting the ratio of the volume of the nuclear fuel material existing in the MA fuel region 2 to the volume of the MA fuel region 2 (fuel volume ratio) to 50% or more, as described later, the MA fuel region 2 Becomes critical. Also in the parent material region 3, the ratio of the volume of the parent fuel material existing in the parent material region 3 with respect to the volume of the parent material region 3 (fuel volume ratio) is set to 50% or more. Since the fuel material is irradiated with neutrons generated by the fission of the nuclear fuel material in the MA fuel region 2 and newly generates a fissile material, the criticality is maintained in the region where the new nuclear fuel material is generated in the parent material region 3 can do.
Although not shown, a control rod for adjusting the reactor power is inserted into the core fuel region 4 during operation of the fast reactor.

  FIG. 4 shows the neutron energy (eV) dependence of the micro cross section σf of the fission reaction and the micro cross section σc of the neutron capture reaction. In FIG. 4, the solid line indicates the micro cross section σf of the fission reaction, and the alternate long and short dash line indicates the micro cross section σc of the neutron capture reaction. Furthermore, in FIG. 4, a broken line is an elastic scattering cross section, and a dotted line is an inelastic scattering cross section.

  As shown in FIG. 4, the fission reaction of Am-241 is a threshold reaction that increases as the neutron energy increases. On the other hand, the neutron capture reaction of Am-241 decreases with increasing energy. When the energy is about 700 keV or more, σf / σc> 1, and Am-241 becomes critical. All major MAs such as Np-237 and Am-243 can be made critical with almost the same neutron energy.

  In a standard sodium cooled fast reactor, the average energy of neutrons is several hundred keV. Since this average energy is lower than the above 700 keV, it is necessary to reduce as much as possible Na of the coolant and Fe which is a main element of the structural member of the fast reactor, which cause energy reduction due to elastic scattering of neutrons. The average fuel volume ratio Vf in the core of a standard fast reactor is about 30% to 40%. When the average fuel volume ratio Vf is 50% or more, the volume ratio of the coolant and the structural member is decreased, so that the neutron moderating ability due to the elastic scattering described above is reduced, so the average energy becomes 700 keV or more, and the MA fuel region Even if only MA is loaded as the nuclear fuel material in the core fuel region 4 of No. 2, the fast reactor becomes critical.

  During operation of the fast reactor, a liquid metal (for example, liquid sodium) that is a coolant pressurized by a pump (not shown) is supplied into the core 1 from below. This liquid metal flows into the MA fuel region 2 from the lower end of the MA fuel region 2, rises in the MA fuel region 2, and flows into the parent material region 3. The liquid metal rising in the parent material region 3 is discharged from the parent material region 3, that is, the upper end of the core 1. In the MA fuel region 2 and the parent material region 3, the liquid metal flows through each fuel assembly 7. The liquid metal flowing into the fuel assembly 7 from the entrance nozzle rises in a passage formed in the neutron shielding body provided in the fuel assembly 7, and further, a coolant formed between the fuel rods 8. The passage rises and is discharged out of the fuel assembly 7, that is, out of the core 1 from the coolant outlet. The liquid metal rising between the annual rods 8 removes heat generated by the fission of MA in the fuel rods 8 (decay heat generated in the fuel rods 7 when the fast reactor is stopped). That is, the liquid metal cools the fuel rod 8. The liquid metal heated in the fuel assembly 7 is discharged from the fuel assembly 7.

  In the early stage of operation of the fast reactor having the core 1, a fission chain reaction by MA occurs only in the MA fuel region 2, and this fission chain reaction continues. For this reason, in the initial stage of operation of the fast reactor, heat is generated by fission only in the MA fuel region 2. A part of neutrons generated by the fission of MA in the MA fuel region 2 leaks from the MA fuel region 2 to the adjacent parent material region 3. Neutrons that have reached the parent material region 3 are irradiated to the parent material fuel (for example, U-238) in the parent material region 3, and the parent material fuel, for example, Pu that is a new fissile material from U-238. -239 is generated. As a result, the burnup of the parent material region 3 improves with time.

  FIG. 5 shows the burnup dependence of the infinite neutron multiplication factor k∞ in the parent material region 3. The neutron infinite multiplication factor k∞ increases as the burnup increases, and exceeds 1.0 when the burnup is constant b1. When the neutron infinite multiplication factor k∞ of the parent material region 3 exceeds 1.0, the parent material region 3 has a parent material region 3 in the parent material region 3 even if no neutrons are incident on the parent material region 3 from the MA fuel region 2. The fission chain reaction by the newly generated fissile material is maintained and the thermal energy necessary for power generation is generated.

  When the parent material fuel in the parent material region 3 is irradiated with neutrons from the MA fuel region 2 at a position in the parent material region 3 near the MA fuel region 2, the infinite neutron multiplication factor k∞ at that position Increases as time passes, as described above. When the burnup at that position in the parent material region 3 increases to about 300 GWd / t, the infinite neutron multiplication factor k∞ exceeds 1.0, and the fission chain reaction can be maintained. At this time, even if MA fission is completed in the MA fuel region 2, the rated operation of the fast reactor can be continued with an electric output of 1000 MWe.

  The fission chain reaction in the parent material region 3 occurs near the MA fuel region 2 in the parent fuel material region 3 where neutrons are likely to enter from the MA fuel region 2, and the end of the parent material region 3 on the MA fuel region 2 side. It gradually proceeds from the portion toward the other end portion of the parent substance region 3 located on the opposite side of the end portion. The progress of the fission chain reaction in the parent material region 3 is caused by the fact that neutrons newly generated in the parent material region 3 by the incidence of neutrons from the MA fuel region 2 are the MA fuel in the parent material region 3. A new fissile material is generated by being absorbed by the parent material fuel existing on the side opposite to the end on the region 2 side. As described above, in the parent material region 3, another end of the parent material region 3 on the side of the MA fuel region 2 is located on the side opposite to the end of the MA fuel region 2 side. A new fissile material is generated toward the end, and the region where the fission chain reaction occurs is also automatically moved toward the other end of the parent material region 3. As a result, for about 60 years until the core 1 reaches the end of its life, the refueling of the fast reactor having the core 1 without supplying new fissile materials such as enriched uranium and fissionable Pu to the core 1 by refueling. The rated output operation can be continued.

  According to this embodiment, the rated power operation of the fast reactor can be continued without supplying new nuclear fuel material into the core 1 by fuel exchange over a long period of time. Can be reduced.

  The core of the fast reactor according to embodiment 2, which is another embodiment of the present invention, will be described below with reference to FIGS.

  The fast reactor core 1A of the present embodiment includes an MA fuel region 2A and a parent material region 3A arranged in the horizontal direction in the MA fuel region 2 and the parent material region 3 arranged in the axial direction in the core 1 of the first embodiment. It has the structure replaced with. A plurality of MA fuel assemblies 15 are loaded into the MA fuel region 2A, and a plurality of parent material fuel assemblies 17 are loaded into the parent material region 3A (see FIG. 7).

  As shown in FIG. 8, the MA fuel assembly 15 has a plurality of fuel rods 16 and a trumpet tube 9. The plurality of fuel rods 16 are bundled by a plurality of fuel spacers (not shown) arranged in the axial direction of the MA fuel assembly 15 and arranged in the trumpet tube 9. Although not shown in the drawing, the MA fuel assembly 15 is similar to the fuel assembly 7 and has a coolant outlet located above the fuel rod bundle bundled by the fuel spacer, and below the fuel rod bundle. It has a neutron shield located and a coolant inlet (entrance nozzle). Each fuel rod 16 of the MA fuel assembly 15 is filled with MA metal obtained by reprocessing spent nuclear fuel over the entire length of the effective fuel length of each fuel rod 16.

  The parent material fuel assembly 17 has the same hardware configuration as the MA fuel assembly 15, and each of the fuel rods installed in the trumpet pipe has a metal of natural uranium, degraded uranium, or recovered uranium as a parent material as a nuclear fuel material. Filled. Although not shown in FIG. 7, the neutron reflector region 5 loaded with a plurality of neutron reflector assemblies 10 includes the MA region 2A and the parent material region except for the contact portion between the MA region 2A and the parent material region 3A. A neutron shield flow region 6 that is disposed surrounding 3A and loaded with a plurality of neutron shield assemblies 11 surrounds the neutron reflector region 5.

  In this embodiment, control rods are inserted into the MA fuel region 2A and the parent material region 3A, respectively. These control rods are operated to control the reactor power of the fast reactor having the core 1A.

  During operation of the fast reactor, liquid metal, which is a coolant whose pressure is increased by a pump (not shown), is supplied to the MA fuel assembly 15 and the parent material fuel assembly 17, respectively.

  At the initial stage of operation of the fast reactor having the core 1A, the fission chain reaction by MA occurs only in the MA fuel assembly 15 in the MA fuel region 2A, and this fission chain reaction continues. For this reason, in the initial operation of the fast reactor, heat is generated by fission only in the MA fuel region 2A. Part of the neutrons generated by the fission of MA in the MA fuel region 2A leaks from the MA fuel region 2A to the parent material region 3A adjacent to the MA fuel region 2A in the horizontal direction. The parent material fuel (for example, U-238) in the parent material fuel assembly 17 existing in the parent material region 3A is irradiated with neutrons to generate a new fissile material (Pu-239). For this reason, even if the fissionable MA in the MA fuel region 2A disappears, the region where the fission chain reaction generated by the newly generated fissionable material occurs in the parent material region 3A is indicated by the white arrow. The power generation continues.

  The other end of the parent material region 3A in which the fission chain reaction of the newly generated fissile material in the parent material region 3A is opposite to the end of the parent material region 3A on the MA fuel region 2A side. When the Pu generated by neutron capture of U-238 is accumulated at the other end of the parent material region 3A, the operation of the fast reactor is stopped. Each of the MA fuel assembly 15 in the MA fuel region 2A and the parent material fuel assembly 17 in the parent material region 3A, where the combustion of the fission material is finished, is replaced with a new MA fuel assembly 15 having a burnup of 0 GWd / t and Replace with parent material fuel assembly 17. Thereafter, when the operation of the fast reactor is resumed, the fission chain reaction by the fissionable Pu accumulated at the right end of the parent material region 3A starts, and this time, in the direction opposite to the white arrow, that is, To the left, the zone where the fission chain reaction takes place moves.

  In the present embodiment, the effects produced in the first embodiment can be obtained. Furthermore, this embodiment has an advantage that propagation in the combustion region and continuation of power generation can be controlled by exchanging the fuel assemblies in the MA fuel region 2A and the parent material region 3A.

  In each of the embodiments described above, the nuclear fuel loaded in the core of the fast reactor is made of metal, but the same effect can be obtained by using nitride. Furthermore, although liquid Na was used as the coolant for the fast reactor, the same effect can be obtained with liquid lead or lead-bismuth.

  DESCRIPTION OF SYMBOLS 1,1A ... Core, 2, 2A ... Minor actinide fuel area | region, 3, 3A ... Parent material area | region, 4 Core fuel area | region, 5 ... Neutron reflector area | region, 6 ... Neutron shield body area | region, 7 ... Fuel assembly, 8, 16 ... Fuel rod, 10 ... Neutron reflector assembly, 11 ... Neutron shield assembly, 15 ... MA fuel assembly, 16 ... Parent material fuel assembly.

Claims (6)

  A fast reactor core comprising: a first fuel region containing only a minor actinide as a nuclear fuel material; and a second fuel region disposed adjacent to the first region and containing a parent material fuel as a nuclear fuel material. .   The fast reactor core according to claim 1, wherein the second fuel region is disposed above the first fuel region.   A plurality of fuel assemblies having a plurality of fuel rods filled with the minor actinide in a lower region and filled with the parent material fuel in an upper region disposed above the lower region and adjacent to the lower region To the core fuel region formed in the first fuel region and the second fuel region, the first fuel region is formed in the lower region, and the second fuel region is formed in the upper region. The core of the fast reactor according to claim 2.   The core of the fast reactor according to claim 1, wherein the first fuel region and the second fuel region are arranged in a horizontal direction.   A plurality of first fuel assemblies having a plurality of first fuel rods filled with the minor actinides therein are loaded in the first fuel region, and a plurality of second fuel rods filled with the parent material fuel therein. The core of a fast reactor according to claim 4, wherein a plurality of second fuel assemblies having the following are loaded in the second fuel region.   The ratio of the volume of the nuclear fuel material in the first fuel region to the volume of the first fuel region in the first fuel region is 50% or more, and the volume of the second fuel region in the second fuel region is greater than the volume of the second fuel region. The core of the fast reactor according to any one of claims 1 to 5, wherein a volume ratio of the nuclear fuel material in the second fuel region is 50% or more.

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