USRE28894E - Method for the preparation of concentrated anion-deficient salt solutions - Google Patents
Method for the preparation of concentrated anion-deficient salt solutions Download PDFInfo
- Publication number
- USRE28894E USRE28894E US05/587,209 US58720975A USRE28894E US RE28894 E USRE28894 E US RE28894E US 58720975 A US58720975 A US 58720975A US RE28894 E USRE28894 E US RE28894E
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- United States
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- sub
- solution
- anion
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- Expired - Lifetime
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- 150000001450 anions Chemical class 0.000 title claims abstract description 16
- 230000002950 deficient Effects 0.000 title claims abstract description 16
- 239000012266 salt solution Substances 0.000 title claims abstract description 9
- 238000000034 method Methods 0.000 title claims description 14
- 238000002360 preparation method Methods 0.000 title claims description 12
- 229910052768 actinide Inorganic materials 0.000 claims abstract description 7
- VGBPIHVLVSGJGR-UHFFFAOYSA-N thorium(4+);tetranitrate Chemical compound [Th+4].[O-][N+]([O-])=O.[O-][N+]([O-])=O.[O-][N+]([O-])=O.[O-][N+]([O-])=O VGBPIHVLVSGJGR-UHFFFAOYSA-N 0.000 claims abstract description 6
- FLDALJIYKQCYHH-UHFFFAOYSA-N plutonium(IV) oxide Inorganic materials [O-2].[O-2].[Pu+4] FLDALJIYKQCYHH-UHFFFAOYSA-N 0.000 claims abstract description 4
- 229910000310 actinide oxide Inorganic materials 0.000 claims abstract description 3
- 239000007864 aqueous solution Substances 0.000 claims abstract description 3
- JCMLRUNDSXARRW-UHFFFAOYSA-N uranium trioxide Inorganic materials O=[U](=O)=O JCMLRUNDSXARRW-UHFFFAOYSA-N 0.000 claims abstract 5
- FCTBKIHDJGHPPO-UHFFFAOYSA-N uranium dioxide Inorganic materials O=[U]=O FCTBKIHDJGHPPO-UHFFFAOYSA-N 0.000 claims abstract 3
- 239000000243 solution Substances 0.000 claims description 39
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 claims description 6
- 150000001255 actinides Chemical class 0.000 claims description 5
- 229910002005 actinide nitrate Inorganic materials 0.000 claims 1
- OOAWCECZEHPMBX-UHFFFAOYSA-N oxygen(2-);uranium(4+) Chemical compound [O-2].[O-2].[U+4] OOAWCECZEHPMBX-UHFFFAOYSA-N 0.000 claims 1
- 150000003839 salts Chemical class 0.000 claims 1
- 238000003756 stirring Methods 0.000 claims 1
- 239000000919 ceramic Substances 0.000 abstract description 4
- 239000003758 nuclear fuel Substances 0.000 abstract description 3
- 239000002245 particle Substances 0.000 abstract description 3
- -1 actinide metals Chemical class 0.000 abstract description 2
- 229910052751 metal Inorganic materials 0.000 abstract description 2
- 239000002184 metal Substances 0.000 abstract description 2
- 239000007858 starting material Substances 0.000 abstract description 2
- 238000004519 manufacturing process Methods 0.000 abstract 1
- 238000003980 solgel method Methods 0.000 abstract 1
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 18
- 229910017604 nitric acid Inorganic materials 0.000 description 18
- 229910002007 uranyl nitrate Inorganic materials 0.000 description 15
- 125000005289 uranyl group Chemical group 0.000 description 14
- 229910052770 Uranium Inorganic materials 0.000 description 7
- 239000002253 acid Substances 0.000 description 7
- 239000000463 material Substances 0.000 description 6
- 239000012798 spherical particle Substances 0.000 description 6
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 description 6
- 229910002651 NO3 Inorganic materials 0.000 description 5
- NHNBFGGVMKEFGY-UHFFFAOYSA-N Nitrate Chemical compound [O-][N+]([O-])=O NHNBFGGVMKEFGY-UHFFFAOYSA-N 0.000 description 5
- WZECUPJJEIXUKY-UHFFFAOYSA-N [O-2].[O-2].[O-2].[U+6] Chemical class [O-2].[O-2].[O-2].[U+6] WZECUPJJEIXUKY-UHFFFAOYSA-N 0.000 description 4
- 239000007788 liquid Substances 0.000 description 4
- 229940021384 salt irrigating solution Drugs 0.000 description 4
- 239000002699 waste material Substances 0.000 description 4
- 150000001875 compounds Chemical class 0.000 description 3
- 238000000605 extraction Methods 0.000 description 2
- 239000000203 mixture Substances 0.000 description 2
- 239000000843 powder Substances 0.000 description 2
- 238000010992 reflux Methods 0.000 description 2
- 239000000126 substance Substances 0.000 description 2
- ZOXJGFHDIHLPTG-UHFFFAOYSA-N Boron Chemical compound [B] ZOXJGFHDIHLPTG-UHFFFAOYSA-N 0.000 description 1
- 229910003556 H2 SO4 Inorganic materials 0.000 description 1
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 1
- 150000001224 Uranium Chemical class 0.000 description 1
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 description 1
- 229910052796 boron Inorganic materials 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 239000003795 chemical substances by application Substances 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 238000001704 evaporation Methods 0.000 description 1
- 230000008020 evaporation Effects 0.000 description 1
- 239000012467 final product Substances 0.000 description 1
- 238000010438 heat treatment Methods 0.000 description 1
- 239000001257 hydrogen Substances 0.000 description 1
- 229910052739 hydrogen Inorganic materials 0.000 description 1
- 229910000041 hydrogen chloride Inorganic materials 0.000 description 1
- 239000012535 impurity Substances 0.000 description 1
- 238000009434 installation Methods 0.000 description 1
- GQPLMRYTRLFLPF-UHFFFAOYSA-N nitrous oxide Inorganic materials [O-][N+]#N GQPLMRYTRLFLPF-UHFFFAOYSA-N 0.000 description 1
- 230000001590 oxidative effect Effects 0.000 description 1
- 239000001301 oxygen Substances 0.000 description 1
- 229910052760 oxygen Inorganic materials 0.000 description 1
- 229910052761 rare earth metal Inorganic materials 0.000 description 1
- 150000002910 rare earth metals Chemical class 0.000 description 1
- 229920006395 saturated elastomer Polymers 0.000 description 1
- 239000007787 solid Substances 0.000 description 1
- 229910000439 uranium oxide Inorganic materials 0.000 description 1
- 229910052727 yttrium Inorganic materials 0.000 description 1
- VWQVUPCCIRVNHF-UHFFFAOYSA-N yttrium atom Chemical compound [Y] VWQVUPCCIRVNHF-UHFFFAOYSA-N 0.000 description 1
- 150000003755 zirconium compounds Chemical class 0.000 description 1
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
- G21C3/42—Selection of substances for use as reactor fuel
- G21C3/58—Solid reactor fuel Pellets made of fissile material
- G21C3/62—Ceramic fuel
- G21C3/623—Oxide fuels
-
- C—CHEMISTRY; METALLURGY
- C01—INORGANIC CHEMISTRY
- C01G—COMPOUNDS CONTAINING METALS NOT COVERED BY SUBCLASSES C01D OR C01F
- C01G43/00—Compounds of uranium
-
- C—CHEMISTRY; METALLURGY
- C01—INORGANIC CHEMISTRY
- C01G—COMPOUNDS CONTAINING METALS NOT COVERED BY SUBCLASSES C01D OR C01F
- C01G43/00—Compounds of uranium
- C01G43/006—Compounds containing uranium, with or without oxygen or hydrogen, and containing two or more other elements
-
- C—CHEMISTRY; METALLURGY
- C01—INORGANIC CHEMISTRY
- C01G—COMPOUNDS CONTAINING METALS NOT COVERED BY SUBCLASSES C01D OR C01F
- C01G56/00—Compounds of transuranic elements
- C01G56/003—Compounds containing transuranic elements, with or without oxygen or hydrogen, and containing two or more other elements
-
- 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
Definitions
- the invention relates to the preparation of concentrated anion-deficient salt solutions.
- Anion-deficient salt solutions are for instance suitable for the preparation of solid oxide and carbide particles.
- the invention aims at giving improved methods for the preparation of an anion-deficient uranyl nitrate solution. Besides it appeared that anion-deficient actinide salt-solutions could be prepared according to several more methods than was formerly possible.
- one or more actinide oxides as PuO 2 , UO 3 or lower uranium oxides than UO 3 are dissolved in a small volume of an acid reacting liquid.
- the acid reacting liquid consists of a small amount of a strong acid such as a small amount of concentrated HNO 3 , HCl or H 2 SO 4 or an aqueous solution of an actinide salt of a strong acid as for instance UO 2 (NO 3 ) 2 or Th(NO 3 ) 4 .
- uranium concentration is at least 2 molar.
- water soluble boron yttrium, rare earth metals and zirconium compounds.
- anion-deficient solutions of the required nitrate/actinide metal ratio can be obtained by causing lower oxides than UO 3 to react with strong nitric acid, uranyl nitrate solution, thorium nitrate solution or mixtures of these substances in the quantities calculated on the basis of the requirements.
- Lower uranium oxides than UO 3 are the compounds U 3 O 8 and UO 2 . These oxides, along with uranyl nitrate, are the forms in which uranium is obtainable as a basic material. They are also the forms in which uranium is preferably conveyed.
- the required anion-deficient uranyl nitrate solution may be characterized as follows: ##EQU1##
- Difficultly soluble UO 2 is likewise converted by this thermal processing into easily soluble U 3 O 3 .
- UO 2 Very difficultly soluble UO 2 is converted into U 3 O 8 by being sintered in air at 700° C.
- the cubic lattice of UO 2 is thereby changed into he othorhombic lattice of U 3 O 8 .
- the molecular volume of U 3 O 8 is greater than that of UO 2 , since UO 2 is of higher density than U 3 O 8 , the particles are completely crumbled.
- the high specific surface areas of the U 3 O 8 obtained in this way has the effect that it can now be readily dissolved in HNO 3 .
- the preparation of U 3 O 8 as described above is the ideal method of utilizing waste obtained in the preparation of the ceramic fissile material.
- the waste may consist either of unsintered waste material, possibly containing organic filter material, or of sintered final product composed of UO 2 .
- the quantities of nitric acid used can be determined by calculation.
- Example II deals with the processing of spherical particles of unsintered UO 3 .
- Example III deals with the conversion of waste material from spherical particles of UO 2 sintered at high temperatures.
- Examples IV relates to the dissolving of U 3 O 8 in uranyl nitrate solution.
- the solution obtained was found to have an NO 3 '/U ratio of 1.6.
- the HNO 3 /U 3 O 8 ratio used 2300/415 ⁇ 5.5. According to the gross equation (2) an NO 3 '/U ratio ⁇ 1.5 may be reckoned with.
- the clear solution obtained had a 2.49 molar content of uranium and an NO 3 '/U ratio of 1.62.
Landscapes
- Chemical & Material Sciences (AREA)
- Organic Chemistry (AREA)
- Engineering & Computer Science (AREA)
- Inorganic Chemistry (AREA)
- Life Sciences & Earth Sciences (AREA)
- Geology (AREA)
- General Life Sciences & Earth Sciences (AREA)
- Physics & Mathematics (AREA)
- Ceramic Engineering (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Inorganic Compounds Of Heavy Metals (AREA)
Abstract
Concentrated anion-deficient salt solutions are prepared of the actinide oxides, PuO2, UO2, UO3 and U3 O8 by dissolving one or more oxides in an aqueous solution of thorium nitrate at a concentration of 4 molar or greater and at a temperature of 60° C. or more. Anion-deficient salt solutions of actinide metals so produced are useful as starting materials for the manufacture of ceramic nuclear fuel particles by the sol-gel process.
Description
This is a division of application Ser. No. 106,922, filed Jan. 15, 1971, now abandoned.
The invention relates to the preparation of concentrated anion-deficient salt solutions.
Anion-deficient salt solutions are for instance suitable for the preparation of solid oxide and carbide particles.
For the preparation of spherical particles of ceramic nuclear fuel an anion-deficient solution of uranylnitrate can successfully be used as a starting material.
In the prior art these solutions have been prepared according to the following methods:
(1) By dissolving UO3 in concentrated uranyl nitrate solutions,
(2) By the extraction of nitric acid from stoichiometric, possibly slightly acid uranyl nitrate solutions.
These methods show, however, the following drawbacks.
For the purpose of the first method it is necessary to have at one's disposal a UO3 of such a texture that this substance easily dissolves in the uranyl nitrate solution.
As to the second method it is observed that extraction, whereby nitric acid is withdrawn from a stoichiometric or weakly acid uranyl nitrate solution, can only be applied to dilute uranyl nitrate solutions. Moreover, a special installation is needed for this. After removal of the nitric acid the solution obtained has to be brought to the required degree of concentration, e.g. by evaporation.
The invention aims at giving improved methods for the preparation of an anion-deficient uranyl nitrate solution. Besides it appeared that anion-deficient actinide salt-solutions could be prepared according to several more methods than was formerly possible.
According to the invention one or more actinide oxides as PuO2, UO3 or lower uranium oxides than UO3 are dissolved in a small volume of an acid reacting liquid. The acid reacting liquid consists of a small amount of a strong acid such as a small amount of concentrated HNO3, HCl or H2 SO4 or an aqueous solution of an actinide salt of a strong acid as for instance UO2 (NO3)2 or Th(NO3)4.
Mixtures of the above-mentioned liquids can be used too.
With a small amount of liquid is meant that in case of an anion-deficient uranyl nitrate solution the uranium concentration is at least 2 molar.
It is possible to incorporate during the preparation or thereafter small amounts of compounds of other elements in the anion-deficient actinide salt solution in order to improve the properties of nuclear fuel material prepared from this solution.
By other compounds are meant water soluble boron, yttrium, rare earth metals and zirconium compounds.
Examples of the preparation of mixed anion-deficient actinide salt solutions are the dissolving of PuO2 in uranyl nitrate solution and of UO3 in thorium nitrate solution.
It has surprisingly been found that anion-deficient solutions of the required nitrate/actinide metal ratio can be obtained by causing lower oxides than UO3 to react with strong nitric acid, uranyl nitrate solution, thorium nitrate solution or mixtures of these substances in the quantities calculated on the basis of the requirements.
The use of lower uranium oxides than UO3 has the advantage of better solubility in acid solutions than UO3. The difficulty of preparing a UO3 with a suitable texture namely can be avoided.
Lower uranium oxides than UO3 are the compounds U3 O8 and UO2. These oxides, along with uranyl nitrate, are the forms in which uranium is obtainable as a basic material. They are also the forms in which uranium is preferably conveyed.
It is therefore of importance to convert these oxides in the eaiest possible manner into the solution required for the process to be employed.
The required anion-deficient uranyl nitrate solution may be characterized as follows: ##EQU1##
It is observed that this uranium concentration is higher than that of the saturated stoichiometric uranyl nitrate solution.
For the preparation of ceramic fissile material a solution of this kind is first mixed with ammonia-liberating agent and then solidified by being dispersed in a phase of sufficiently high temperature, non-miscible with water. With this method it is of great importance to start with highly concentrated uranium solutions.
In order to make the rate of solution of the uranium oxide in nitric acid as high as possible, it is important to prepare the U3 O8 by heating in an oxidizing atmosphere, such as air or oxygen, at temperatures between 600° and 900° C. At these temperatures the most volatile and/or combustible impurities are removed and the texture of the material is still conducive to solution.
Difficultly soluble UO2 is likewise converted by this thermal processing into easily soluble U3 O3.
Very difficultly soluble UO2 is converted into U3 O8 by being sintered in air at 700° C. The cubic lattice of UO2 is thereby changed into he othorhombic lattice of U3 O8. As the molecular volume of U3 O8 is greater than that of UO2, since UO2 is of higher density than U3 O8, the particles are completely crumbled. The high specific surface areas of the U3 O8 obtained in this way has the effect that it can now be readily dissolved in HNO3.
The preparation of U3 O8 as described above is the ideal method of utilizing waste obtained in the preparation of the ceramic fissile material. For this purpose the waste may consist either of unsintered waste material, possibly containing organic filter material, or of sintered final product composed of UO2.
In accordance with the undermentioned gross equations (1) and (2), the quantities of nitric acid used can be determined by calculation.
2UO.sub.2 + 5HNO.sub.3 →
2{UO.sub. 2 (NO.sub.3).sub.1.5 (OH).sub. 0.5 } + 2H.sub.2 O + NO+NO.sub.
2 (1)
2U.sub.3 O.sub.8 + 11HNO.sub.3 →
6{UO.sub.2 (NO.sub. 3).sub.1.5 (OH).sub.0.5 } + 4H.sub.2 O
(2)O+NO.sub.2
The invention is further elucidated below by reference to a number of examples.
Example I deals with the preparation of an anion-deficient uranyl nitrate solution by dissolving UO2 powder in nitric acid.
Example II deals with the processing of spherical particles of unsintered UO3.
Example III deals with the conversion of waste material from spherical particles of UO2 sintered at high temperatures.
Examples IV relates to the dissolving of U3 O8 in uranyl nitrate solution.
A solution test was carried out with natural UO2 powder in nitric acid with the undermentioned quantities of UO2 and HNO3.
weighed-out
UO.sub.2 :
11.4854 g.
= 42.5 mmol of UO.sub.2
HNO.sub.3 :
3 × 42.5
= 127.5 mmol of HNO.sub.3,
diluted with water to 100 ml. In this example UO2 was added in portions to the hot (˜80° C.) HNO3 solution.
On account of the fact that during solution in an open beaker some losses of nitric acid occurred, slightly more nitric acid was used than was equivalent to equation (1).
The solution obtained was found to have an NO3 '/U ratio of 1.6.
A quantity of spherical particles of UO3 was heated slowly in air to 700° C. and then kept at this temperature for another four hours. The following was obtained: ##EQU2##
This quantity was added in portions to a heated HNO3 solution consisting of 160 ml. of concentrated HNO3 (14.4 M) and 258 ml. of water in a beaker. The total volume amounted to 160+258= 418 ml., so that after solution the uranium concentration is about 3 M.
The HNO3 /U3 O8 ratio used= 2300/415≈5.5. According to the gross equation (2) an NO3 '/U ratio ≦1.5 may be reckoned with.
Analysis of the solution obtained gave the following results: ##EQU3## density 1.965 g./cm.3 (20.6° C.).
The solution tests were repeated with two quantities of spherical particles of UO3 with a 20% and 40% enrichment respectively, after they had first been converted into U3 O8.
The results obtained in this way are set forth below in Table A.
It was observed that by operating in a three-necked flask with a reflux cooler the nitrous vapors had reformed a quantity of HNO3.
TABLE A
__________________________________________________________________________
[U]
Degree of Density,
enrich-
U.sub.3 O.sub.8,
G. mol
Ml. HNO.sub.3,
Mol HNO.sub.3 /
H.sub.2 O,
g./cm..sub.3,
Meas-
Calcu-
[NO.sub.3 ']/
ment grams
U.sub.3 O.sub. 8
14.4M mol U.sub.3 O.sub.8
ml. 21°C.
ured
lated
[U]
__________________________________________________________________________
20% 770.1
0.917
350 5.50 500 1.866
2.82
2.82
1.76
40% 622.3
0.743
270 5.25 300 1.904
2.95
2.97
1.58
__________________________________________________________________________
644.1 grams of spherical particles of UO2 (sintered at 1400° C. in an atmosphere containing hydrogen), were slowly heated to 750° C. and then kept for four hours at this temperature. In this way 662.5 grams of U3 O8 were obtained, which could readily be passed into solution according to the method indicated in Example II.
In this example a quantity of 116 g. of
UO.sub.2 (NO.sub.3).sub.2.sup.. 6H.sub.2 O
was dissolved in 72 ml. of water and then boiled under reflux with 13.7 g. of U3 O8 for 21/2 hours.
The clear solution obtained had a 2.49 molar content of uranium and an NO3 '/U ratio of 1.62.
Claims (2)
1. A method for preparing a concentrated anion-deficient actinide salt solution containing at least one actinide oxide selected from the group consisting of PuO2, UO2, UO3 and U3 O8, said method including dissolving at a temperature of at least 60° C., said .[.salt.]. .Iadd.oxide .Iaddend.in an aqueous solution of thorium nitrate having a concentration of at least 4 molar.
2. A method for the preparation of a concentrated anion-deficient actinide nitrate solution wherein at least one member selected from the group consisting of uranium dioxide, uranium trioxide and U3 O8, is dissolved by stirring in a heated thorium nitrate solution of a temperature of at least 60° C., and of a concentration of at least 4 molar, and the solution thus obtained is thereafter diluted with water.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| US05/587,209 USRE28894E (en) | 1970-01-16 | 1975-06-16 | Method for the preparation of concentrated anion-deficient salt solutions |
Applications Claiming Priority (6)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| NL7000674 | 1970-01-16 | ||
| NL7000674A NL7000674A (en) | 1970-01-16 | 1970-01-16 | Method for preparing an anion-deficient uranyl nitrate solution |
| AU24159/71A AU428956B2 (en) | 1971-01-08 | A method for the preparation of concentrated anion-deficient salt solutions | |
| AU24159/71 | 1971-01-08 | ||
| US00283291A US3838062A (en) | 1971-01-15 | 1972-08-24 | Method for the preparation of concentrated anion-deficient salt solutions |
| US05/587,209 USRE28894E (en) | 1970-01-16 | 1975-06-16 | Method for the preparation of concentrated anion-deficient salt solutions |
Related Parent Applications (2)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| US10692271A Division | 1971-01-15 | 1971-01-15 | |
| US00283291A Reissue US3838062A (en) | 1970-01-16 | 1972-08-24 | Method for the preparation of concentrated anion-deficient salt solutions |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| USRE28894E true USRE28894E (en) | 1976-07-06 |
Family
ID=27422771
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| US05/587,209 Expired - Lifetime USRE28894E (en) | 1970-01-16 | 1975-06-16 | Method for the preparation of concentrated anion-deficient salt solutions |
Country Status (1)
| Country | Link |
|---|---|
| US (1) | USRE28894E (en) |
Citations (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3171715A (en) * | 1963-05-22 | 1965-03-02 | Alfred T Kleinsteuber | Method for preparation of spherical thorium dicarbide and thorium-uranium dicarbide particles |
| US3307772A (en) * | 1964-07-29 | 1967-03-07 | Garrod Norman John | Gramophone record sleeves |
| US3361676A (en) * | 1967-02-10 | 1968-01-02 | Atomic Energy Commission Usa | Urania sol forming method in the presence of formic acid and a palladiumon-thoria catalyst |
| US3401122A (en) * | 1964-06-24 | 1968-09-10 | Cnen | Process for producing dense particles of oxides of actinides usable as fuels for nuclear reactors |
-
1975
- 1975-06-16 US US05/587,209 patent/USRE28894E/en not_active Expired - Lifetime
Patent Citations (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3171715A (en) * | 1963-05-22 | 1965-03-02 | Alfred T Kleinsteuber | Method for preparation of spherical thorium dicarbide and thorium-uranium dicarbide particles |
| US3401122A (en) * | 1964-06-24 | 1968-09-10 | Cnen | Process for producing dense particles of oxides of actinides usable as fuels for nuclear reactors |
| US3307772A (en) * | 1964-07-29 | 1967-03-07 | Garrod Norman John | Gramophone record sleeves |
| US3361676A (en) * | 1967-02-10 | 1968-01-02 | Atomic Energy Commission Usa | Urania sol forming method in the presence of formic acid and a palladiumon-thoria catalyst |
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