GB2098788A - Process and apparatus for molten-salt nuclear chemical reaction process for circulating a fuel salt in a combination system of the apparatus with a molten- salt converter reactor - Google Patents
Process and apparatus for molten-salt nuclear chemical reaction process for circulating a fuel salt in a combination system of the apparatus with a molten- salt converter reactor Download PDFInfo
- Publication number
- GB2098788A GB2098788A GB8214599A GB8214599A GB2098788A GB 2098788 A GB2098788 A GB 2098788A GB 8214599 A GB8214599 A GB 8214599A GB 8214599 A GB8214599 A GB 8214599A GB 2098788 A GB2098788 A GB 2098788A
- Authority
- GB
- United Kingdom
- Prior art keywords
- salt
- molten
- target
- amsb
- fissile material
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
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Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C1/00—Reactor types
- G21C1/04—Thermal reactors ; Epithermal reactors
- G21C1/06—Heterogeneous reactors, i.e. in which fuel and moderator are separated
- G21C1/22—Heterogeneous reactors, i.e. in which fuel and moderator are separated using liquid or gaseous fuel
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21G—CONVERSION OF CHEMICAL ELEMENTS; RADIOACTIVE SOURCES
- G21G1/00—Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation or particle bombardment, e.g. producing radioactive isotopes
- G21G1/04—Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation or particle bombardment, e.g. producing radioactive isotopes outside nuclear reactors or particle accelerators
- G21G1/10—Arrangements for converting chemical elements by electromagnetic radiation, corpuscular radiation or particle bombardment, e.g. producing radioactive isotopes outside nuclear reactors or particle accelerators by bombardment with electrically charged particles
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Abstract
A process for molten-salt nuclear chemical reaction characterized by using a single-fluid type accelerator molten-salt breeder and effecting the reaction while containing a fissile material in the molten salt; an apparatus for the same reaction which comprises a single- fluid type accelerator molten-salt breeder comprising a molten-salt nuclear reactor vessel, a heat exchanger 7, a pump 9 for circulating a molten salt between said nuclear reactor vessel and said heat exchanger, a linear accelerator 17 for generating a fast charge particle such as proton set up so that the particle is injected directly onto the liquid surface of molten salt, and a target molten-salt bypass system 20 having a purification means 22, 23 for the salt; and a fuel usage system utilising the combination of high gain type accelerator molten- salt breeder and molten-salt converter reactor. <IMAGE>
Description
SPECIFICATION
Process and apparatus for molten-salt nuclear chemical reaction and process for circulating a fuel salt in a combination system of the apparatus with a molten-salt converter reactor
The present invention relates to a process and apparatus for molten-salt nuclear chemical reaction and a process for circulating a fuel salt in a combination system of the apparatus with a molten-salt converter reactor (hereinafter abbreviated to "MSCR").
The present inventors have previously invented a standard type accelerator molten-salt breeder (hereinafter abbreviated to "s.AMSB") and filed a patent application (8111129). This s.AMSB aimed at producing fissile materials without any fissile material in initial and a process by which 233U (or 239Pu) can be produced in the mere presence of Th (or U) without any fissile material has been provided, however the concentrations of 233U produced was below 0.1 m/o 233UF4 even after the operation of reactor for one year and it was not considered to have a large amount of fissile material inventory. However, if 233U is increasingly accumulated, it will be possible to design and operate a higher gain and more economical AMSB by utilizing the 233U.That is, in the remained fissile material neutrons are increased by about 2.5 times by its own nuclear fission, thereby the production efficiency of fissile material is raised on the whole.
And in proportion to the increase of nuclear fission heat generated, accordingly power increases, thereby the security of self-comsumption power and the supply of surplus power are more ensured.
One object of the present invention is to provide a process for molten-salt nuclear chemical reaction having improved in production efficiency of fissile material and heat generation efficiency and an improved highgain type AMSB (hereinafter abbreviated to "h.g.AMSB") used for the process.
Another object of the present invention is to provide the most rationalized process for circulating a nuclear fuel for supplying fissile material 233U produced in the h.g.AMSB to the MSCR.
As the result of diligent research, the present inventors have found that the former object can be accomplished basically by mixing a considerably large amount of fissile material self-produced or supplied into a target molten salt in the above described AMSB, and have invented a process for molten-salt nuclear chemical reaction using an AMSB characterized by effecting said reaction while containing fissile materials in the molten salt, and an apparatus for molten-salt nuclear chemical reaction characterized by provided with a target molten salt purification bypass system having such as a liquid metal contact apparatus and a nickel filter in an AMSB comprising a molten-salt nuclear reactor vessel, a heat exchanger and a pump for circulating a molten salt from the nuclear reactor vessel via the heat exchanger to the same, which is a cylindrical vessel having closed top and bottom made of Hastelloy N provided with a graphite shield along the inner wall and having an opening on the top, and on which a linear-accelerator for generating a fast charge particle such as proton is set up so that the particle is injected directly onto the liquid surface of molten salt through the opening, and have further invented a process for circulating a fissile material produced by h.g.AMSB which comprises elevating the concentration of 233UF4 in a target salt containing 0.2 to 1 rnol% of 233UF4 from the h.g.AMSB by adding 233UF4 obtained by reprocessing a fuel salt taken out from a MSCR to the target salt when adding the target salt to t"3 fuel salt of MSCR to supply 233UF4 in a combination system of h.g.AMSB and MSCR.
Figure 1 is a schematic view of AMSB used for effecting a nuclear chemical reaction in the present invention.
In the figure, 1 ... an incident opening of proton beam (1 GeV, 300mA); 2. . . an inlet of molten salt; 3. . . an outlet passage; 4... a nuclear @herwi@al chemical reaction apparatus; 5 and 6... a graphited shield; 7. . a piping; 8... a heat exchanger; 9... a pump for circulating a molten salt; 1 0... a flow control valve; 11 ...an overflow piping; 1 2... a storage tank of molten salt; 1 3... an orifice; 1 4... a shutter; 1 5. . a vacuum exhausting system; 16. .. a steam trap; 1 7... a linear accelerator; 1 8. . a conduit; 1 9... a magnet for converging proton beams; 20... a bypass piping for purifying a salt; 21 ... a pump for bypass; 22... a liquid metal contact reaction apparatus; 23. . a nickel wool filter.
Figure 2 is a schematic view showing an embodiment process in the fuel circulation process of the present invention.
In this figure, 1 ... 1 GeV, 300 mA proton accelerator; 2. . h.g.AMSB; 3... a tank; 4... reprocessing facilities; 5... MSCR; 6... a tank.
The structure of h.g.AMSB used in the present invention is basically the same as that of USSN 250,553 but is devised to decrease the total amount of target salt and the absolute quantity of fissile material contained and the fissile material is positively added to the target molten salt. According to an object and plan the addition of 0.2 to 1 mol% 233UF4 is considered. It does not come up to critical point in nuclear reactivity therewith. However, when the concentration becomes more higher, the amount of fissile material required increases and so it becomes uneconomical. This concentration should be selected so as to be sufficiently more concentric than the compo sition of fuel salt for the molten-salt (fission) power reactor projected to be supplied 233U produced and operated, and utilized as it is.
And in the former AMSB a nuclear spallation reaction in a narrow sense and a fission reaction proceed concurrently while in a case such as the present invention only the fission reaction increases and so much nuclear reaction waste (impurity) increases. Therefore, in the process of the present invention a target molten salt bypass system provided with such as, for example a liquid metal contact reaction apparatus and a nickel wool filter is provided.
The molten salt is pumped out from the storage tank in moderation (about 10 to 30% in volume in year) and, after applying a purification operation, supplied to the molten salt power reactor and, instead, Li F-BeF2-Th F4 salt, etc. not containing a fissile material is added thereto while controlling the fluctuation of salt composition.
In the present invention the operation can be started using 235UF4 or 239PuF3 instead of 233UF4. And also the U containing salt as described in co-pending USSN 250,533 can be used instead of Th containing salt to the s.AMSB as a 239Pu production reactor.
The target salt taken out from the AMSB is needed to be subjected to a purification operation including removing operation of oxides, removal of suspended solid material by filtration, etc. before being transported to the
MSCR.
The fuel salt withdrawn for the MSCR is taken into batch chemical reprocess facilities in an area center having AMSB (safeguarded area) and is at first taken out 235UF4 only.
Therefor 233UF6 taken out utilizing 233UF6 vaporization by fluorination treatment by fluorine gas is added to a target salt for supply while reducing to UF4. And 7LiF or BeF2 is suitably added thereto for making equal the fuel salt composition of MSCR.
The ThF4 containing salt cleaned by chemical reprocessing is adjusted to a salt suitable for diluent of target salt having almost the same composition as the target salt by adding
ThF4 and a small amount of BeF2, etc., and added to the AMSB on occasion.
The present invention will be explained more in detail by the following Examples.
Example 1
In the AMSB, a nuclear reaction vessel used has such a shape that the upper half of salt sink is a cylinder of about 4.5 m in diameter and the lower half is a reversed cone, the shape and dimension of which may be somewhat modified. The volume of molten salt in the whole system is about 90 m3. The composition of molten salt used is 7LiF-BeF2-ThF4 293U F4 (64-18-17.5-0.5 mol%) added with fissile material. In this case the total amount of Th is about 1 20 tons and the amount of 233U required is amount 3.4 tons.
In the present invention the 233U production efficiency is increased by 40 to 50% in comparison with the case in which fissile materials are not added. Therefore, the fissile materials are produced at the rate of about 1.2 ton per year and so the doubling time is 2 to 3 years.
This means an increase of about 0.2 mol% per year. Thus, the salt in the storage tank 12, which is equivalent to the target molten salt in composition through the overflow piping, is taken out about once per month and the same amount of salt, 7LiF-BeF2-ThF4 (64-18-18 mol%), not containing fissile materials is added thereto.
Since the amount of impurities produced by nuclear fission is increased than the case of s.AMSB (about 40 kg is increased to about 1 70 kg per year), a bypass system 20 for continuously purifying the molten salt is provided as shown in the figure. This is a system of branching off from the heat exchanger outlet piping 7 and returning to the main pump for circulation of molten salt 9 through the liquid metal contact reaction apparatus 22 and the nickel wool filter 23 by means of the pump 21.
The target salt taken from the storage tank 1 2 is further purified to be utilized as a fissile material additive to a fuel salt for a molten salt power reactor. And 233U may be extracted therefrom to be utilized as a solid fuel. This high gain type AMSB is increased about 330,000 KW (heat) in power, which is treated by increasing the flow of target molten salt by about 1 m3 per second.
Example 2
A fissile material added to the target salt in
Example 1 was 233U which can be replaced with 20% 235U + 80% 238U. In the case about
14 tons of 238U is replaced with 14 tons in about 1 20 tons of 238Th. Therefore, about 12% of 233U produced is converted to 239Pu.
However, the concentration of 239Pu decreases gradually since 238U (and 235U) is not added thereafter. And in the molten salt power reactor supplied with this fissile material, 239Pu becomes extinct more quickly.
Example 3
Even in case of using 239PuF3 as a fissile material added early, almost the same or higher efficiency can be expected.
As understood from the above, a fissile material is needed in the initial stage of reaction although an excellent doubling time of 2 to 4 years can be expected. And the production efficiency of fissile material is increased
by above 50%.
Further, the heat generation increases and the security of necessary electric power is ensure so that the sale of power is possible.
In the present invention the salt taken out from the target salt can be charged as it is
into the fuel salt of molten salt (fission) power reactor to serve for addition of fissile material.
But the removal of impurity is necessary.
The apparatus of the present invention is scarcely required to change new R and D even if the design is changed. Therefore, the optimum operation in the composition of molten salt at the point of time is possible while considering the specification of power reactor, the condition of nuclear energy and economy.
Example 4
The h.g.AMSB having 1 GeV, 300 mA proton accelerator can produce about 1.33 tons of 233U fuel for one year when using 7LiF
BeF2-ThF4-233UF4 (64-18-17.5-0.7 mol%) as a target salt. Therefore, fissile materials necessary for supply of about 1 7 each of 1,000,000 KWe MSCR (conversion ration 0.9) can be supplied thereby.
Then, the fuel cycle will be explained for one MSCR. 650 kg of salt taken out from
MSCR is 7LiF-BeF2-ThF4-233UF4 (71.75-16-12-0.25 mol%) in composition and about 7.8 kg of 233UF4 (5.7 kg of 233U) is separated by fluoride vaporization at the chemical reprocessing facilities in the area center. When adding to 560 kg of target salt taken out from AMSB, it becomes 7LiF-BeF2
ThF4-233UF4 (63.8-17.9-17.4-0.85 mol%).
This means addition of about 8 kg of 233U if deducting the amount taken out from MSCR.
Such operation is carried out about 10 times per year.
And, when adding to MSCR, about 75 kg of 7LiF and about 1 8 kg of BeF2 are added simultaneously for adjusting the composition.
Example 5
Using 7LiF-BeF2-ThF4-233UF4 (71.5-16-12-0.5 mol%) similar to the salt taken out from MSCR in composition as a target salt the production and circulation of fuel can be done in the same manner as in
Example 4.
According to the present invention, the most economical fuel cycle can be constituted by the minimum operation. Thus, the present invention has such an effect that the fwml salt used is a fluoride which is chemically safe, vaporizable (or escapable) wastes has been already treated in AMSB or MSCR and it gets off with the minimum material transportation.
Furthermore, in the process of the present invention, since the fissile material is transported as a material having solidified molten salt containing a considerably large amount of radioactive material outside of safe guarded area center, it has a high y radioactivity so that the fissile material cannot be simply separated. And the fissile material 233U has a complicated isotope consitution containing 232U, 234U(235U) and 235U, and is high in critical mass, and so it has such an effect that it cannot be easily made to an explosive.
Claims (11)
1. A process for molten-salt nuclear chemical reaction characterized by using a single-fluid type accelerator molten-salt breeder and effecting the reaction while containing a fissile material in the molten salt.
2. The process as set forth in claim 1 wherein said fissile material is 233UF4.
3. The process as set forth in claim 1 wherein said fissile material is 235UF4.
4. The process as set forth in claim 1 wherein said fissile material is 239put3.
5. An apparatus for molten-salt nuclear chemical reaction which comprises a singlefluid type accelerator molten-salt breeder comprising a molten-salt nuclear reactor vessel, a heat exchanger and a pump for circulating a molten salt between said nuclear reactor vessel and said heat exchanger, said nuclear reactor vessel being a cylindrical vessel having closed top and bottom made of Hastelloy N provided with a graphite shield alone the inner wall and having an opening on the top, and on which a linear accelerator for generating a fast charge particle such as proton is set up so that the particle is injected directly onto the liquid surface of molten salt through the opening, and which is characterized by provided with a target molten-salt bypass system having such as a liquid metal contact reaction means and a nickel filter.
6. In a system comprising a combination of high gain type accelerator molten-salt breeder and molten-salt converter reactor, a process for circulating a fissile material produced by said AMSB which comprises elevating the concentration of 233UF4 in a target salt containing 0.2 to 1 mol% of 233UF4 from said
AMSB by adding 233UF4 obtained by reprocessing a fuel salt withdrawn from said converter reactor to said target salt when adding said target to the fuel salt of said converter reactor to supply 233UF4 thereto.
7. The process as set forth in claim 6 wherein the composition of molten salt is 7LiF
BeF2-ThF4-233UF4.
8. The process as set forth in claim 6 wherein the target salt from said AMSB is subjected to an operation of removing impurities.
9. The process as set forth in claim 6 wherein the ThF4 containing salt cleaned in reprocessing facilities is adjusted to almost the same composition as that of target salt by adding ThF4 and a small amount of BeF2 thereto and then is added to said AMSB as a salt for dilution.
10. A process substantially as hereinbefore described with reference to, and as illustrated in, the accompanying drawings.
11. An apparatus substantially as hereinbefore described with reference to, and as illustrated in, the accompanying drawings.
1 2. A process substantially as herein before described with reference to any one or more of Examples 1 to 5.
Applications Claiming Priority (2)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP56074897A JPS57190293A (en) | 1981-05-20 | 1981-05-20 | Melt salt nuclear reaction method and device |
| JP56194443A JPS5896277A (en) | 1981-12-04 | 1981-12-04 | Method of circulating fuel of accelerator melted salt breeder reactor |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| GB2098788A true GB2098788A (en) | 1982-11-24 |
Family
ID=26416065
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| GB8214599A Withdrawn GB2098788A (en) | 1981-05-20 | 1982-05-19 | Process and apparatus for molten-salt nuclear chemical reaction process for circulating a fuel salt in a combination system of the apparatus with a molten- salt converter reactor |
Country Status (3)
| Country | Link |
|---|---|
| DE (1) | DE3218852A1 (en) |
| FR (1) | FR2519792A1 (en) |
| GB (1) | GB2098788A (en) |
Cited By (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| EP0617430A1 (en) * | 1993-03-24 | 1994-09-28 | Kazuo Furukawa | Plutonium annihilating nuclear reactor with use of liquid nuclear fuel |
| GB2468892A (en) * | 2009-03-25 | 2010-09-29 | Tate & Lyle Technology Ltd | A molten salt treatment system and process |
| CN112349436A (en) * | 2020-11-06 | 2021-02-09 | 西安交通大学 | Liquid metal cooling wire winding positioning molten salt reactor core |
Family Cites Families (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3385759A (en) * | 1967-05-08 | 1968-05-28 | Atomic Energy Commission Usa | Fast burst neutronic reactor |
| DE2922608C2 (en) * | 1979-06-02 | 1982-02-25 | Kernforschungsanlage Jülich GmbH, 5170 Jülich | Spallation source targets, methods for their cooling and use |
| CA1183287A (en) * | 1980-04-15 | 1985-02-26 | Kazuo Furukawa | Single fluid type accelerator molten-salt breeder |
-
1982
- 1982-05-19 DE DE3218852A patent/DE3218852A1/en not_active Withdrawn
- 1982-05-19 FR FR8208785A patent/FR2519792A1/en not_active Withdrawn
- 1982-05-19 GB GB8214599A patent/GB2098788A/en not_active Withdrawn
Cited By (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| EP0617430A1 (en) * | 1993-03-24 | 1994-09-28 | Kazuo Furukawa | Plutonium annihilating nuclear reactor with use of liquid nuclear fuel |
| GB2468892A (en) * | 2009-03-25 | 2010-09-29 | Tate & Lyle Technology Ltd | A molten salt treatment system and process |
| CN112349436A (en) * | 2020-11-06 | 2021-02-09 | 西安交通大学 | Liquid metal cooling wire winding positioning molten salt reactor core |
| CN112349436B (en) * | 2020-11-06 | 2021-10-19 | 西安交通大学 | Liquid metal cooling wire winding positioning molten salt reactor core |
Also Published As
| Publication number | Publication date |
|---|---|
| FR2519792A1 (en) | 1983-07-18 |
| DE3218852A1 (en) | 1982-12-09 |
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Legal Events
| Date | Code | Title | Description |
|---|---|---|---|
| WAP | Application withdrawn, taken to be withdrawn or refused ** after publication under section 16(1) |