JPH07191171A - Small-sized atomic reactor using liquid nuclear fuel - Google Patents

Abstract

PURPOSE:To accelerate combustion by filling the void in a reactor core where a decelerating material is arranged, with a liquid nuclear fuel consisting of Tr and Pu, separating any time as applicable the <233>U produced from Tr using a collecting device, and thereby adding Pu instead. CONSTITUTION:A salt consisting of molten fluoride (<7>LiF-BeF2-TrF4-<239>PuF3) which is stable chemically and has a normal pressure, intrudes into a reactor core from inlets 4, 4... and passes upward through reactor voids leading to decelerating materials 10, 13, 14 to flow out from the outlet 5. If a control rod 8 is inserted into the central region I of the reactor core using a driving mechanism 9, absorption of neutrons lessens, and the density of neutrons heightens to accelerate the reaction. The reaction is such that <239>Pu makes fission to generate an energy and also produce neutrons, and part of the neutrons are absorbed by <232>Tr so that conversion of <233>U takes place. The <233>U generated by absorption into Tr of a part of the produced neutrons undergoes a fluorinastion process at two-year intervals approximately by a lower situated drain tank 20 and fluorination device 22, and is separated from the fuel salt and collected back.

Description

Detailed Description of the Invention

[0001]

FIELD OF THE INVENTION The present invention relates to a small nuclear reactor with liquid nuclear fuel.

[0002]

The inventor of the present application has already invented Japanese Patent Application No. 60-272165 as a small nuclear reactor using such liquid nuclear fuel. This small molten salt power reactor invented earlier has a simple structure, operation and maintenance that does not require replacement of moderator graphite and does not require continuous chemical treatment. Plutonium in the nuclear fuel is fissionable. To generate energy and to convert thorium to uranium by neutrons generated at that time,
The reaction is continued.

[0003]

At present, however, plutonium obtained by reprocessing from waste nuclear warheads or spent nuclear fuel is a big problem. The present invention is to change the fuel cycle form of the conventional small-sized nuclear reactor using the above liquid nuclear fuel and, if necessary, slightly change the internal structure of the nuclear reactor so that the plutonium can be effectively burned out as nuclear fuel.

[0004]

According to the present invention, the inside of a core cavity in which a moderator is arranged is communicated with a uranium recovery device, and the inside of the cavity is filled with a liquid nuclear fuel composed of thorium Th and plutonium Pu. Uranium ²³³ U produced from thorium Th is characterized by being separated at any time by a recovery device, and instead of which plutonium Pu is added to promote combustion. The uranium recovery device comprises a drain tank 20 and a fluorination device 22.

[0005]

The salt containing thorium Th and plutonium Pu nuclear fuel flows into the void of the core, and the plutonium in the nuclear fuel undergoes fission to generate energy and neutrons generated at that time convert thorium into uranium ²³³ U. Uranium ²³³ U produced from this thorium Th is fluorinated by the recovery device and separated at any time, and instead of adding plutonium Pu, its combustion is promoted.

[0006]

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will be described in detail below with reference to the drawings. The one shown is for a 15,000 kW power reactor. As shown in FIG. 1, a flat, cylindrical reactor vessel 3 made of a Ni—Mo—Cr alloy is arranged below the thick wall shields 1 and 2 made of concrete. The dimensions of each part are as compared with the scale line 2m in FIG. The lower part of the container 3 is provided with salt inlets 4 and 4, and the upper part is provided with salt outlets 5 and 5. The composition of this salt is ⁷ LiF-BeF ₂ -ThF ₄
- ²³⁹ in PuF _3, ⁷ LiF of mol% is 71.7, Be
F ₂ is 16, ThF ₄ is 12, and ²³⁹ PuF ₃ is 0.3. A graphite reflector 6 is arranged around the container 3, and control rods 8, 8 ... Made of graphite are inserted in a central region I of a core 7 inside the container 3 so as to be vertically moved by a drive mechanism 9,
As shown in FIGS. 1 and 2, moderators 10 made of fixed graphite having a length of 2 m are arranged around the moderator 10. The moderator 10 has a length of about 2 m and support portions at the upper and lower ends.

The moderator 10 in the central region I is shown in FIG.
(A) Six rhombic elongated plates 11, 11 ... Having dimensions shown in the horizontal cross section shown in the figure are joined in a hexagonal shape so that a constant void 12 is formed between them by a projection or the like. The porosity is 6 to 8%, preferably 7%. Therefore, the volume ratio of graphite is 94 to 92%, preferably 93%. Moderators 13, 13 ... Made of graphite are also arranged in the peripheral region II outside the central region I. 2 are shown in white in FIG. 2, and the horizontal cross section thereof is substantially the same as that in FIG. 3A, but the porosity is 8 to 12
%, Preferably 10%. Therefore, the volume ratio of graphite is 92 to 88%, preferably 90%. The peripheral area
Also in the blanket region III outside II, the moderators 14, 14 ... Of similar graphite are arranged with a thickness of 30 to 50 cm. The moderator 14 is nine rhombus having dimensions as shown in the horizontal cross section of FIG. 3B, and is a hexagon so that a constant void 16 is formed between the elongated plates 15, 15 ... It is bonded to the shape and has a porosity of 30 to 34%, preferably 32%. Therefore, the volume ratio of graphite is 70 to 6
6%, preferably 68%. The dimensions of the graphite reflector 6, the central region I of the core 7, the peripheral region II, and the blanket region III in the container 3 are as shown in FIG.

The container 3 is composed of graphite moderators 10, 13, 1
After filling 4 inside, it will be sealed. Therefore, the movable portion is only the drive mechanism 9 for the central control rod 8. The total amount of fuel salt is 12.1 m ³ including the outside of the core, and
It becomes 0.5 ton. Among ^{2 39} Pu is 530 kg, Th is 1.75Ton. The center of the reactor core 7 communicates with the lower portion of the lower drain tank 20 via a pipe 19, and the lower portion of the drain tank 20 is connected via a conduit 21 to the fluorination apparatus 2
2, a helium and fluorine gas supply pipe 23 is connected to the lower part of the fluorination device 22, and an outlet 24 for helium and fluorine gas and UF ₆ is provided on the upper part thereof.

Next, the operation of this device will be described. Chemically stable and normal pressure molten fluoride ⁽⁷ LiF-BeF ₂ -T
hF ₄ - ²³⁹ salt consisting PuF ₃₎ the inlet 4,4 ...
It further enters the core 7, passes through the voids 12, 12, 16 to the moderators 10, 13, 14 from the bottom to the top and flows out from the outlet 5. When the control rod 8 is inserted into the central region I of the reactor core 7 by the drive mechanism 9, the absorption of neutrons is reduced, the neutron density is increased, and the reaction is accelerated contrary to the conventional reactor. In this reaction, plutonium ²³⁹ Pu undergoes fission to generate energy and neutrons, and part of the neutrons is absorbed by thorium ²³² Th to convert it to uranium ²³³ U. The conversion rate is about 90%.
During this operation, rare gas elements (Kr, Xe) of fission products
Does not dissolve in salt, so about 99% of the cover gas is separated outside the furnace.

The fuel is replenished by adding ⁷ LiF- ²³⁹ PuF ₃ salt to the dump tank of the salt at any time.
At that time, a little dirty fuel salt is removed to keep the capacity constant.

On the other hand, the high temperature fuel salt discharged from the furnace flows through the two salt loop pipes and flows into the secondary heat molten salt [N] in the first heat exchanger.
aBF ₄ -NaF (92-8 mol%)], and then heat is transferred to water by the second heat exchanger to generate steam to generate turbine power. About 43% of efficiency can be secured under supercritical conditions. ²³³ U generated by absorbing a part of the generated neutrons to Th is fluorinated by the lower drain tank 20 and the fluorination device 22 every one or two years, separated and recovered from the fuel salt, Used as fuel for other general molten salt power reactors. The processed fuel is returned to the original furnace, Pu is added again, and it is restarted, and it is used for Pu extinction and ²³³ U production. ^{233 On} the occasion of U separation work, part of some fission / reaction products was subjected to a distillation method in addition to the above fluorination method.
It can be used for neutron utilization efficiency improvement, corrosion prevention, etc. by separating by oxidative precipitation method, liquid metal contact extraction method, etc.
The degree of separation is determined by the economic efficiency of the entire system.

(Example 1) 15,000 kW (heat 350,000 kW)
In a small output molten salt power generation reactor (see the example of Japanese Patent Application No. 60-272165), ⁷ LiF-BeF ₂ was used as fuel salt.
-ThF ₄ - ²³⁹ PuF ₃ (71.7-16-12-
0.3 mol%) was used. The average power density of this furnace is as high as about 10 kW th / liter. The initial loading of ²³⁹ Pu is about 530 kg. Approximately 250 kg in 2 years while adding 146 kg / year Th and approximately 26 kg / year Pu
Was burned and incinerated. Two years later, the burned salt was transferred to a treatment tank and a mixed gas of fluorine and helium was blown into it, and about 1
70 kg of ²³³ U were separated as UF ₆ gas. The salt was returned to the furnace again, and about 220 kg of Pu was added to restart the operation. In about 20 years, about 2.3 tons of Pu could be extinguished, and about 1.5 tons of ²³³ U could be produced and obtained.

(Example 2) P from spent fuel of a light water reactor
u ( ²³⁹ Pu: ²⁴⁰ Pu: ²⁴¹ Pu: ²⁴² Pu = 56.
5: 25.3: 13.2: 5.0) to extinguish the 100,000 kW (in a small molten salt power reactor heat 250,000 kW) Output, ⁷ to the fuel salt LiF-BeF ₂ -ThF ₄ - Pu
F ₃ (71.81-16-12-0.19 mol%) was used. The power density of this furnace is about 4 kW th / liter on average. The initial loading amount of Pu was about 390 kg, and further about 320 kg of Pu and 220 kg of Th were added for 1000 days and the operation was continued to burn about 240 kg of Pu. After 1000 days, the fuel salt was transferred to a treatment tank and fluorinated, and about 190 kg of ²³³ U was separated as UF ₆ . After that, the salt was returned to the furnace again, and about 280 kg of Pu was added to restart the operation. In this reactor, about 1.7 tons of Pu from a light water reactor can be extinguished in 20 years, and about 1.4 tons of ²³³
I was able to produce U.

[0014]

The effects of the present invention are summarized as follows. (1) More Pu can be burned. ²³³ U produced from Th is separated from time to time, and averages about 1 compared to Pu.
Since it is maintained at 0% or less, the burning rate of Pu becomes about 85% to 95% of the total fissionable material. (2) ²³³ U is produced from Th not only for Pu combustion but also for effective use of neutrons generated by nuclear fission, but since it is not placed in the reactor too much, ²³³ U that burns is small and most of it is recovered. Available for different furnaces. Since this furnace has a high conversion rate of about 0.8 or more, the amount of Pu consumed is almost the same (7
It was converted to ²³³ U (0 to 90%). (3) Moreover, at that time, power generation is performed. Separation work is 1
Since it can be set to more than a year, or every 3 to 5 years in some cases, the power generation function is not so disturbed. Further, it is also possible to separate continuously by a bypass circuit while operating the furnace. (4) Most of the fissile material consumed during operation is ²³³ U
Therefore, the reactivity does not change so much. That is, operation and maintenance work is easy.

[Brief description of drawings]

FIG. 1 is a vertical sectional view of an embodiment of the present invention.

FIG. 2 is a sectional view taken along line AA in FIG.

3A and 3B are plan views of a moderator in two regions of a core.

FIG. 4 is an explanatory diagram showing the dimensions of each region of the core.

[Explanation of symbols]

 I Central region II Peripheral region III Blanket region 7 Core 10, 13, 14 Moderator 12, 12, 16 Void

Claims (2)

[Claims] 1. A reactor core in which a moderator is arranged communicates with a uranium recovery device, and thorium T is contained in this chamber.
Filling a liquid nuclear fuel consisting of h and plutonium Pu,
Uranium ²³³ U produced from this thorium Th is separated at any time by a recovery device, and plutonium Pu is added instead to promote combustion, and is a small nuclear reactor using liquid nuclear fuel. 2. The small nuclear reactor with liquid nuclear fuel according to claim 1, wherein said uranium recovery device comprises a drain tank 20 and a fluorination device 22.