JP2007163245A - Nuclear reactor loaded with spontaneous neutron emission nuclear fuel - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To vanish at low cost Pu 240 or Pu 242 emitting spontaneous neutrons which may produce a nuclear bomb. <P>SOLUTION: In a core loaded with a spontaneous neutron type nuclear fuel assembly designed based on a simplified equation of a diffusion equation including spontaneous neutrons, a cooling system in a conventional BWR is changed from a double-phase flow of water into gaseous steam. <P>COPYRIGHT: (C)2007,JPO&INPIT

Description

The present invention relates to a nuclear reactor in which spontaneous neutron emission nuclear fuel is loaded.

A nuclear reactor is a device in which nuclear fuel such as uranium (U), neptunium (Np), and plutonium (Pu), coolant, and others are arranged so that a controlled fission chain reaction can be sustained for a long time.
Although there are various types of nuclear reactors, light water reactors (pressurized water reactor (PWR) and boiling water reactors) that use micro-enriched uranium or plutonium mixed with micro-enriched uranium as a heat source and light water as a coolant (BWR)) is the mainstream.
FIG. 1 is a schematic view of a conventional nuclear fuel rod (31) of a BWR. Zircaloy-coated tube (41), upper end plug (42) and lower end plug (43) for hermetically closing the upper and lower opening ends of the coated tube (41), and a length of about 370 cm in the coated tube (41) And consists of a number of nuclear fuel pellets (44) and springs (45) in an upper plenum (48) that accumulates gaseous fission products. The nuclear fuel pellet (44) is composed of an enriched uranium oxide enriched with uranium 235 which is easily fissioned.
FIG. 2 is a schematic perspective view of a conventional nuclear fuel assembly (30) containing nuclear fuel material loaded on a BWR (Patent Document 1). The nuclear fuel assembly (30) includes a cylindrical nuclear fuel rod (31) enclosing a nuclear fuel material arranged in a square lattice, and an upper coupling plate (32) for supporting the upper end and the lower end thereof. ) And the lower coupling plate (33), several spacers (34) which are located in the middle of the height of the nuclear fuel rod (31) and regulate the interval between the nuclear fuel rods (31), and these are arranged on four surfaces A channel box (35) having one side of a cross-section covered with about 14 cm.
A cross-sectional view of the nuclear fuel assembly (30) at a height where the spacer (34) is not located is shown in FIG. A gap is secured between the nuclear fuel rods (31) by the spacer (34). During the power operation, most of the control rods (36) are pulled out to the lower part of the nuclear reactor, so that there is a leakage water passage (51) between the adjacent nuclear fuel assemblies (30) as shown in FIG. FIG. 5 shows an example of the core arrangement of the nuclear fuel assembly (30). 1 is the first unloaded nuclear fuel assembly (30), 2 is the nuclear fuel assembly (30) burned for about 1 year, 3 is the nuclear fuel assembly (30) burned for about 1 year, 4 is 3 Is a nuclear fuel assembly (30) burned for about one year. The cross-shaped dotted line indicates the arrangement of the control rod (36).
FIG. 6 shows an overview of a pressure vessel (60) of a conventional boiling water reactor (1) (Non-patent Document 1). The water that has finished work in the turbine is mixed into the shroud water (66) between the pressure vessel (60) wall and the shroud (39) through the water supply pipe (67). The water is accelerated by a coolant circulation pump (37) rotated by a pump motor (38), and enters a conventional nuclear fuel assembly (30) in which nuclear fuel rods containing nuclear fuel material are bundled in the direction of the arrow from the lower end of the shroud (39). Unsaturated water flows in, absorbs heat, and part of the liquid water becomes saturated vapor. It flows into the upper part as a two-phase flow in which liquid water and gas saturated vapor coexist. The proportion of saturated steam in the two-phase flow section is called the void fraction. The void ratio is zero in the lower part of the nuclear fuel assembly (30), the middle void ratio is about 40% in the middle, and the high void ratio is about 70% in the upper part.
The two-phase flow in which the two-phase flow in the direction of the dotted arrow containing a large amount of saturated steam from the upper part of the nuclear fuel assembly (30) and the water in the direction of the arrow from the leakage material passage (51) are mixed is an air-water separator. By being swirled into (65), it is separated into saturated steam rising in the direction of the open arrow and water falling down in the direction of the arrow. Since the rising saturated steam contains some moisture, it is separated by the steam dryer (62) into dried saturated steam rising in the direction of the open arrow and water falling in the direction of the arrow. The dried saturated steam passes from the steam dome (61) through the wall of the pressure vessel (60) and the steam dryer body (63), and then the steam exits from the saturated steam pipe (64) to the turbine.
The saturated steam in the steam dryer (62) is indicated by a broken line. The saturated steam temperature of the saturated steam separated from the two-phase flow is about 286 degrees Celsius at the saturated steam temperature at an operating pressure of about 70 atm.
Control of reactor power is achieved by a control rod (36) that moves up and down by a control rod drive mechanism.
: Sho 61-37591, “Nuclear Fuel Assembly”. : Corona, author Toko “Atomic power” 117, 120 pages.

Spent nuclear fuel from light water reactors contains plutonium (Pu). Pu emits radiation over a long period of time, making it difficult to dispose or store. Among them, plutonium 240 (Pu240) and plutonium 242 (Pu242) are generated in large quantities, and since they generate a lot of spontaneous neutrons, it is difficult to predict the output, and they are being accumulated. If taken by a terrorist, it can be a nuclear bomb. A Pu242 of about 70kg or a Pu240 of about 20kg can be a nuclear bomb.
The conventional light water reactor is slightly improved to reduce the proportion of coolant water, which is also a moderator, and to reduce the generation of Pu242. It is insignificant to extinguish the fire.
Strictly speaking, conventional nuclear reactors contained spontaneous neutrons. That is, uranium 238 (U238) generates spontaneous neutrons although it is weak. When nuclear fuel burns, Pu242 is produced and spontaneous neutrons exist. However, since there was little Pu242, spontaneous neutrons could be ignored. The effective multiplication factor was about 0.99999, which was close to the critical value of 1.0. Prediction in design etc. solved the diffusion equation as an eigenvalue problem without an external neutron source. However, when a large amount of spontaneous neutrons is actively introduced as an external neutron source as in the present invention, the diffusion equation must be solved as an external neutron source problem.

In the present invention, Pu in conventional light water reactor removal fuel or degraded plutonium (DPu) with a higher ratio of Pu242 and Pu240 compared to this has a content of uranium 235 (U235) more than natural uranium (NU) or NU. Effective with control rod operation in a reactor core consisting of a nuclear fuel with an infinite multiplication factor adjusted by mixing less depleted uranium (DU) or slightly enriched uranium (SEU) enriched with U235 and a coolant of water vapor, helium or carbon dioxide. Operate at a multiplication factor of 1.0 or less. Even in nuclear reactors that use liquid metal as a coolant, effective burning and extinction of Pu242, etc. is possible if this nuclear fuel is used.
In Pu242, the fission energy threshold is 1Mev or more. Therefore, if the energy of the external neutron source is 1 Mev or more, the effect is high. Since the spontaneous neutron energy of Pu242 is 1Mev or more, the effect can be expected. In the thermal neutron reactor, Pu242 becomes a neutron absorber and suppresses the fission chain reaction. Therefore, the nuclear fission chain reaction can be sustained with a nuclear reactor that uses liquid sodium or gas, which has a low decelerating action, or helium or water vapor as a coolant.

If the coolant is water vapor, helium, or carbon dioxide, the neutron moderation action is weak, so even with Pu242 or Pu240 with a high fission energy threshold, fission can be continued sufficiently by fast neutrons. If the nuclear reaction is limited by mixing U238, a stable output can be obtained by operating the control rod even in a nuclear reactor loaded with nuclear fuel that releases a large amount of spontaneous neutrons. The presence of U238 keeps the Doppler reactivity coefficient at a small negative value, so there is no loss of safety. If it is gas cooling, the void reactivity coefficient is kept at a small negative value so that safety is not impaired.

  We were able to provide a reactor that burns and extinguishes Pu242 and Pu240 that emit spontaneous neutrons without sacrificing safety and without significantly increasing power generation costs.

Conventional nuclear reactors also contained spontaneous neutrons. U238 emits a small amount of spontaneous neutrons. Since there were few, the effective multiplication factor was about 0.99999, and it was able to operate at a value close to the criticality. The eigenvalue problem without an external neutron source can be predicted by the diffusion equation and designed. In a nuclear reactor having a large amount of spontaneous neutrons uniformly in the reactor as in the present invention, it is considered that the prediction accuracy is poor by using the diffusion equation as an eigenvalue problem without an external neutron source. We have invented a simplified method for properly predicting the power distribution in the reactor with an external neutron source.
An approximate solution of the two-group diffusion equation considering an external neutron source consisting of a high-speed group and a low-speed group in a three-dimensional system with one side length of X0, Y0, and Z0 is described in detail below. “*” Indicates multiplication, “/” indicates division, and “≈” indicates approximately equal.
g = 1 is a fast group (including fission neutrons and spontaneous neutron spectrum S (E)).
g = 2 is the low speed group.
Φg is the g group neutron flux.
S1 is the number of spontaneous neutrons in the high-speed group per unit volume (sum of S (E)).
Σn2n is a macroscopic nuclear cross section that absorbs one neutron and emits two neutrons.
Σsl1 is the macroscopic deceleration nuclear cross section.
Σcg is the g-group macroscopic capture nucleus cross section.
Σfg is the macroscopic fission cross section of the g group.
νg is the number of neutrons generated by g group fission.
Dg is the macroscopic diffusion coefficient of the g group.
q is the energy generated by fission.
Σsl1, Σc1, ν1, Σf1, Σn2n, D1, ν2, Σf2, Σa2, and D2 are values averaged by the detailed group neutron flux solving the problem with external neutron sources. The above-mentioned nuclear cross section must solve the transport equation as an external neutron source problem, but in the case of near criticality, the value obtained by solving the transport equation as an eigenvalue calculation problem without an external neutron source is an approximation or practical.
ΣR1 = (Σsl1 + Σc1 + Σf1 + Σn2n)
Σa2 = Σc2 + Σf2
ke1 = (ν1 * Σf1 + 2 * Σn2n) / ΣR1
ke2 = (ν2 * Σf2 / Σa2) * (Σsl1 / ΣR1)
kinf = ke1 + ke2
Then, it is generally known that the two-group diffusion equation of neutrons is expressed by the following formulas 1 and 2 [Non-patent Document 2]. Formula 1 is the diffusion equation for the fast group related to neutrons generated by fission and neutrons emitted from the external neutron source, and Formula 2 is the diffusion equation for the slow group related to neutrons decelerated from the fast group. . We call kinf the infinite multiplication factor with spontaneous neutrons.
[Formula 1]
-D1 * ∇ ² Φ1 + ΣR1 * Φ1 = (ν1 * Σf1 + 2 * Σn2n) * Φ1 + ν2 * Σf2 * Φ2 + S1
[Formula 2]
-D2 * ∇ ² Φ2 + Σa2 * Φ2 = Σsl1 * Φ1
Next, B1 ² = B2 ² = (3.1416 / X0) ² + (3.1416 / Y0) ² + (3.1416 / Z0) ²
As
(D1 / ΣR1) = M1 ²
-∇ ² Φ1 = B1 ² * Φ1
(D2 / Σa2) = M2 ²
-∇ ² Φ2 = B2 ² * Φ2
L1 = 1 / (1 + M1 ² * B1 ² )
L2 = 1 / (1 + M2 ² * B2 ² )
keff = (ke1 + ke2 * L2) * L1 = ((ke1 / L2) + ke2) * L1 * L2
When Lg is the neutron leakage rate from the core of the g group and keff is the effective multiplication factor with spontaneous neutrons considering the neutron leakage from the core, the average neutron flux of the g group per unit volume of the core is obtained from Equation 1 and Equation 2. The approximate solution of aveΦg is expressed by Equation 3 and Equation 4. An approximate solution of the average power avePW per unit volume of the core is expressed by Equation 5.
[Formula 3]
aveΦ1 = S1 * L1 / ((1 keff) * ΣR1)
[Formula 4]
aveΦ2 = (Σsl1 / Σa2) * L1 * L2 * S1 / ((1 keff) * ΣR1)
[Formula 5]
avePW = q * (Σf1 * aveΦ1 + Σf2 * aveΦ2)
= q * (Σf1 + Σf2 * (Σsl1 / Σa2) * L2) * S1 * L1 / ((1 keff) * ΣR1)
The output level can be determined by controlling Σa2 and keff with a control rod operation made of a material with large Σa2.
When L2≈1.0, Σsl1 << Σa2, and Σn2n << ΣR1, the approximate expression is as shown in Equation 6.
[Formula 6]
avePW = q * (ke1 / ν1) * S1 * L1 / (1 keff)
Since keff is constant everywhere in the core, if `` (i) '' is a value in region i, the output PW (i) at the position (x, y, z) from the center is -X0 ≤ x ≤ X0, Approximation with a cosine (cos) where leakage increases as the region boundary is reached in the range of −Y0 ≦ y ≦ Y0 and −Z0 ≦ z ≦ Z0 is expressed as Equation 7 as π = 3.1416.
[Formula 7]
PW (i) = (π / 2) ³ * q (i) * (Σf1 (i) + Σf2 (i) * (Σsl1 (i) / Σa2 (i)) * L2) *
S1 (i) * L1 * cos (π * x / X0) * cos (π * y / Y0) * cos (π * z / Z0) / ((1 keff) * ΣR1 (i))
Below, the approximate upper limit of kinf (i) with keff <1.0 in the presence of an external neutron source is obtained.
keff = ((ke1 (i) / L2) + ke2 (i)) * L1 * L2
Therefore, when L2≈1.0, keff≈kinf (i) * L1 * L2. Therefore,
kinf (i) ≒ keff / (L1 * L2)
If keff is 1.0 or less, PW (i) will not be extremely large even if there is spontaneous neutrons.
[Formula 8]
kinf (i) <1 / (L1 * L2)
If kinf (j) near the core boundary where neutrons are likely to leak is small, kinf (i) larger than Equation 8 is possible.
The value of kinf close to 1.0 can be obtained by solving the transport equation as an eigenvalue problem without considering spontaneous neutrons, and keff close to 1.0 can be obtained by solving the diffusion equation as an eigenvalue problem without considering spontaneous neutrons. Is an approximation.
: Hyundai Engineering, 1983, Mikami, “Method and Analysis of Nuclear Fuel Management”

FIG. 7 shows the spontaneous neutron type nuclear fuel of the present invention in which DU, NU or SEU is mixed with DPu containing a large amount of Pu242 or Pu240 so that the infinite multiplication factor (kinf) with spontaneous neutrons is within the range of Equation 8. It is a stick (131). Since Pu contains plutonium 241 (Pu241), which has a relatively short half-life, the composition of DPu changes with time, making it difficult to control kinf. In the present invention, the composition of DPu in the nuclear fuel initially loaded into the reactor is the same, the number of spontaneous neutron emissions in the reactor is kept constant, and kinf is adjusted by the ratio of U235 in DU, NU, or SEU. The output distribution flattening is performed based on Expression 7. The upper and lower parts are loaded with NU-containing spontaneous neutron emission nuclear fuel pellets (145) using DPu and NU mixed oxides as nuclear fuel, and the DU-containing spontaneous neutrons using DPu and DU mixed oxides as nuclear fuel in the center. Emission nuclear fuel pellets (144) were loaded. In the center of the pellet, the solid moderator pellet (101) such as carbon or silicon carbide is used so that the nuclear reaction does not become excessively active even if the coolant becomes too hot and the density decreases and the neutron moderating action decreases extremely. Boron carbide enriched with boron 11 is provided. The cladding tube is a stainless steel cladding tube (141) that is resistant to high temperatures. Zircaloy can be used instead of stainless steel.
FIG. 8 shows a spontaneous neutron nuclear fuel assembly (130) of the present invention composed of spontaneous neutron nuclear fuel rods (131). The channel box (35) is made of stainless steel. The gas coolant flows in the gas coolant passage (149) between the spontaneous neutron fuel rods (131). A gas coolant flows in the leaked gas passageway (52) between the channel boxes (35).
FIG. 9 shows a core of a spontaneous neutron reactor in which the spontaneous neutron nuclear fuel assembly (130) of the present invention is loaded. 21 is a spontaneous neutron type nuclear fuel assembly (130) in which kinf is reduced by DU and NU in which the ratio of DPu and U235 is small. 22 is a spontaneous neutron nuclear fuel assembly (130) in which 21 burns for one year. Reference numeral 23 denotes a spontaneous neutron nuclear fuel assembly (130) in which 22 burns for one year. 24 is a spontaneous neutron nuclear fuel assembly (130) in which 23 burns for one year. 11 is a large kinf spontaneous neutron type nuclear fuel in which the DU of the spontaneous neutron type nuclear fuel assembly (130) is set to NU, NU is set to SEU, and the proportion of U235 is increased to increase the kinf to flatten the power distribution. It is an aggregate (232). 12 is a large kinf spontaneous neutron type nuclear fuel assembly (232) in which 11 burns for one year. 13 is a large kinf spontaneous neutron type nuclear fuel assembly (232) in which 12 is burned for one year. 14 is a large kinf spontaneous neutron nuclear fuel assembly (232) in which 13 burned for one year. The power distribution was flattened by loading the large kinf spontaneous neutron type nuclear fuel assembly (232) with a larger kinf around the core.
FIG. 10 shows an overview of the spontaneous neutron reactor of the present invention with an improved BWR. The shroud (39) was a sealed shroud (54). The cooling of the nuclear fuel rods was changed to a two-phase flow and turned into steam. The low-temperature water vapor that has finished work in the turbine is discharged to the supply air suction nozzle (83) through the supply pipe (81), and enters the spontaneous neutron emission nuclear fuel assembly (130) from the low-temperature gas dome (86) and is heated. After exiting the hot gas dome (55), it exits through the hot gas pipe (56) to the turbine. The outlet of the air supply pipe (81) is in the middle of the air supply suction nozzle (83), and the suction port of the air supply suction nozzle (83) is covered with the air supply pipe skirt (82). Even if the supply air from the supply pipe (81) is interrupted, the gas outside the shroud gas coolant (53) enters the spontaneous neutron emission nuclear fuel assembly (130), and further, always at the bottom cooling fin (84). The cooled bottom cryogenic coolant (85) cools the spontaneous neutron emitting nuclear fuel assembly (130).
For emergency cooling in the event of an accident, liquid is injected from the emergency water injection pipe (87). Liquid is also injected into the leaking gas passageway (52) from the emergency spray pipe (88) to cool the spontaneous neutron emission nuclear fuel assembly (130) through the channel box (35). Since liquid water slows down fast neutrons, Pu242 fission is suppressed.
The reactor power is adjusted by operating the control rod (36) so that the effective multiplication factor (keff) with spontaneous neutrons does not exceed 1.0.
The gas coolant can be helium or carbon dioxide in addition to water vapor.
If reprocessed uranium containing a large amount of Np, transuranium element, uranium 234 (U234), or uranium 236 (U236) is used as DU, it will be useful for annihilation of Np, transuranium element, U234, and U236.
Even when a conventional BWR is loaded with the spontaneous neutron type nuclear fuel assembly of the present invention to eliminate water as much as possible and operate at a high void rate, the output can be obtained while burning out the Pu242. For example, DPu and U mixed nuclear fuel obtained by reprocessing spent MOX fuel from a light water reactor is equivalent to NU-containing spontaneous neutron emission nuclear fuel pellet (145). It is possible to construct a core with a spontaneous neutron type nuclear fuel assembly in which spontaneous neutron type nuclear fuel rods made of nuclear fuel pellets (44) are bundled. This can be realized at an early stage if the core is loaded with a nuclear fuel assembly designed based on the simplified equation of the diffusion equation containing spontaneous neutrons according to the present invention. If the cooling method in the conventional BWR is changed from a two-phase flow of water to gaseous water vapor, a further effect can be expected.

According to the present invention, DPu, Np, transuranium elements, U234, and U236, which are difficult to handle in disposal of spent nuclear fuel, can be effectively utilized and extinguished. The problem of the final disposal site is also reduced. If this is the case, power generation costs will be reduced.
It can be applied to existing gas-cooled reactors, sodium-cooled reactors and PWRs.

Fig. 2 is a schematic view of a conventional nuclear fuel rod (31) of BWR. FIG. 3 is a schematic perspective view of a conventional nuclear fuel assembly (30) containing nuclear fuel material loaded on a BWR. FIG. 3 is a cross-sectional view of the nuclear fuel assembly (30) at a height where the spacer (34) is not located. Fig. 4 is a partial view of the core when the control rod (36) is pulled out during power operation. Shows an example of the core arrangement of the nuclear fuel assembly (30). These are the general-view figures in the pressure vessel (60) of the conventional boiling water reactor. FIG. 3 is an overview of a spontaneous neutron nuclear fuel rod (131) in which DU and NU are mixed with DPu of the present invention. These are sectional drawings of the spontaneous neutron type nuclear fuel assembly (130) of this invention which consists of a spontaneous neutron type nuclear fuel rod (131). Is a core for loading the spontaneous neutron type nuclear fuel assembly (130) of the present invention and the large kinf spontaneous neutron type nuclear fuel assembly (232) having a larger kinf. FIG. 2 is an overview of the spontaneous neutron reactor of the present invention.

Explanation of symbols

Reference numeral 1 denotes a conventional unloaded nuclear fuel assembly (30) with initial loading.
2 is a nuclear fuel assembly (30) in which 1 burned for 1 year.
3 is a nuclear fuel assembly (30) in which 2 burned for 1 year.
4 is a nuclear fuel assembly (30) in which 3 burned for one year.
11 is a large kinf spontaneous neutron type nuclear fuel assembly (232) having a large unkind kinf in the initial loading.
12 is a large kinf spontaneous neutron type nuclear fuel assembly (232) in which 11 burned for 1 year.
13 is a large kinf spontaneous neutron type nuclear fuel assembly (232) in which 12 burned for 1 year.
14 is a large kinf spontaneous neutron fuel assembly (232) in which 13 burned for 1 year.
21 is a spontaneous neutron type nuclear fuel assembly (130) with a small unloaded kinf at the initial loading.
22 is a spontaneous neutron type nuclear fuel assembly (130) in which 21 burned for 1 year.
23 is a spontaneous neutron fuel assembly (130) in which 22 burned for one year.
24 is a spontaneous neutron nuclear fuel assembly (130) in which 23 burned for one year.
30 is a conventional nuclear fuel assembly.
31 is a conventional nuclear fuel rod.
32 is an upper coupling plate.
33 is a lower coupling plate.
34 is a spacer.
35 is a channel box.
36 is a control rod.
37 is a coolant circulation pump.
38 is a pump motor.
39 is a shroud.
41 is a cladding tube.
42 is an upper end plug.
43 is a lower end plug.
44 is a nuclear fuel pellet.
45 is a spring.
48 is the upper plenum.
49 is a cooling water passage 51 and a leakage water passage.
52 is a leak gas passage.
53 is a gas coolant outside the shroud.
54 is a sealing shroud.
55 is a hot gas dome.
56 is a hot gas pipe.
60 is a pressure vessel.
61 is a steam dome.
62 is a steam dryer.
63 is a steam dryer trunk.
64 is a saturated steam pipe.
65 is a steam separator.
66 is water outside the shroud.
67 is a water supply pipe.
81 is a supply pipe.
82 is a supply pipe skirt.
83 is a supply air suction nozzle.
84 is a bottom cooling fin.
85 is a low temperature coolant at the bottom.
86 is a low temperature gas dome.
87 is an emergency water injection pipe.
88 is an emergency spray tube.
101 is a solid moderator pellet.
130 is a spontaneous neutron type nuclear fuel assembly.
131 is a spontaneous neutron type nuclear fuel rod.
141 is a stainless clad tube.
144 is a DU-added spontaneous neutron emission nuclear fuel pellet.
145 is an NU-added spontaneous neutron emission nuclear fuel pellet.
Reference numeral 149 denotes a gas coolant passage.
232 is a large kinf spontaneous neutron nuclear fuel assembly with a larger kinf.

Claims (4)

A DU-containing spontaneous neutron emitting nuclear fuel pellet (144) made of a mixed oxide nuclear fuel in which DU is mixed with DPu that emits a large amount of spontaneous neutrons with a solid moderator (101) as a core is placed in the center of the stainless steel cladding tube (141). The NU-containing spontaneous neutron emission nuclear fuel pellet (145) made of a mixed oxide nuclear fuel in which NU is mixed with DPu having the same composition as the DPu with the solid moderator (101) as the core is loaded into the stainless steel cladding tube (141). Spontaneous neutron type nuclear fuel assembly (130) comprising a large number of spontaneous neutron type nuclear fuel rods (131) loaded on the upper and lower sides and flattening the power distribution in the height direction. In the above nuclear fuel assembly, DPu has the same composition, but DU is replaced with NU and NU is replaced with SEU. A large kinf spontaneous neutron fuel assembly (232) is loaded around the core and the spontaneous neutron type A core characterized in that a nuclear fuel assembly (130) is loaded in the center of the core to flatten the power distribution. Formula 7
PW (i) = (π / 2) ³ * q (i) * (Σf1 (i) + Σf2 (i) * (Σsl1 (i) / Σa2 (i)) * L2) *
S1 (i) * L1 * cos (π * x / X0) * cos (π * y / Y0) * cos (π * z / Z0) / ((1 keff) * ΣR1 (i))
And Equation 8
kinf (i) <1 / (L1 * L2)
The nuclear fuel rod and the nuclear fuel assembly according to claim 1 and the core according to claim 2, wherein the power distribution is flattened by the above. 3. A spontaneous neutron reactor characterized in that the BWR is modified in the core according to claim 2 and steam is used as a coolant.

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