GB1593323A - Method of releasing fission gases from irradiated nuclear fuel - Google Patents
Method of releasing fission gases from irradiated nuclear fuel Download PDFInfo
- Publication number
- GB1593323A GB1593323A GB2020/78A GB202078A GB1593323A GB 1593323 A GB1593323 A GB 1593323A GB 2020/78 A GB2020/78 A GB 2020/78A GB 202078 A GB202078 A GB 202078A GB 1593323 A GB1593323 A GB 1593323A
- Authority
- GB
- United Kingdom
- Prior art keywords
- process according
- oxidant
- nitrogen
- fuel
- irradiated
- 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.)
- Expired
Links
- 238000000034 method Methods 0.000 title claims description 39
- 230000004992 fission Effects 0.000 title claims description 22
- 239000003758 nuclear fuel Substances 0.000 title claims description 16
- 239000007789 gas Substances 0.000 title claims description 14
- JCXJVPUVTGWSNB-UHFFFAOYSA-N Nitrogen dioxide Chemical compound O=[N]=O JCXJVPUVTGWSNB-UHFFFAOYSA-N 0.000 claims description 52
- MWUXSHHQAYIFBG-UHFFFAOYSA-N Nitric oxide Chemical compound O=[N] MWUXSHHQAYIFBG-UHFFFAOYSA-N 0.000 claims description 35
- 239000000446 fuel Substances 0.000 claims description 35
- 239000000047 product Substances 0.000 claims description 29
- YZCKVEUIGOORGS-NJFSPNSNSA-N Tritium Chemical compound [3H] YZCKVEUIGOORGS-NJFSPNSNSA-N 0.000 claims description 23
- 229910052722 tritium Inorganic materials 0.000 claims description 23
- 230000001590 oxidative effect Effects 0.000 claims description 21
- 239000007800 oxidant agent Substances 0.000 claims description 20
- 238000006243 chemical reaction Methods 0.000 claims description 17
- 229910000439 uranium oxide Inorganic materials 0.000 claims description 16
- 239000000203 mixture Substances 0.000 claims description 15
- MGWGWNFMUOTEHG-UHFFFAOYSA-N 4-(3,5-dimethylphenyl)-1,3-thiazol-2-amine Chemical group CC1=CC(C)=CC(C=2N=C(N)SC=2)=C1 MGWGWNFMUOTEHG-UHFFFAOYSA-N 0.000 claims description 14
- 238000007254 oxidation reaction Methods 0.000 claims description 14
- 230000003647 oxidation Effects 0.000 claims description 11
- 239000008188 pellet Substances 0.000 claims description 11
- 239000000843 powder Substances 0.000 claims description 11
- IJGRMHOSHXDMSA-UHFFFAOYSA-N Atomic nitrogen Chemical compound N#N IJGRMHOSHXDMSA-UHFFFAOYSA-N 0.000 claims description 10
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 claims description 9
- 238000010494 dissociation reaction Methods 0.000 claims description 9
- 230000005593 dissociations Effects 0.000 claims description 9
- 229910052760 oxygen Inorganic materials 0.000 claims description 9
- 239000001301 oxygen Substances 0.000 claims description 9
- 238000010438 heat treatment Methods 0.000 claims description 5
- 238000010790 dilution Methods 0.000 claims description 4
- 239000012895 dilution Substances 0.000 claims description 4
- 238000002474 experimental method Methods 0.000 claims description 4
- 229910052757 nitrogen Inorganic materials 0.000 claims description 4
- SHZGCJCMOBCMKK-KGJVWPDLSA-N beta-L-fucose Chemical compound C[C@@H]1O[C@H](O)[C@@H](O)[C@H](O)[C@@H]1O SHZGCJCMOBCMKK-KGJVWPDLSA-N 0.000 claims description 3
- PNDPGZBMCMUPRI-UHFFFAOYSA-N iodine Chemical compound II PNDPGZBMCMUPRI-UHFFFAOYSA-N 0.000 claims description 3
- 229910052743 krypton Inorganic materials 0.000 claims description 3
- DNNSSWSSYDEUBZ-UHFFFAOYSA-N krypton atom Chemical compound [Kr] DNNSSWSSYDEUBZ-UHFFFAOYSA-N 0.000 claims description 3
- 238000012360 testing method Methods 0.000 claims description 3
- 229910052724 xenon Inorganic materials 0.000 claims description 3
- FHNFHKCVQCLJFQ-UHFFFAOYSA-N xenon atom Chemical compound [Xe] FHNFHKCVQCLJFQ-UHFFFAOYSA-N 0.000 claims description 3
- WZECUPJJEIXUKY-UHFFFAOYSA-N [O-2].[O-2].[O-2].[U+6] Chemical compound [O-2].[O-2].[O-2].[U+6] WZECUPJJEIXUKY-UHFFFAOYSA-N 0.000 claims description 2
- 238000007865 diluting Methods 0.000 claims description 2
- 230000000694 effects Effects 0.000 claims description 2
- 239000012467 final product Substances 0.000 claims description 2
- 230000004584 weight gain Effects 0.000 claims description 2
- 235000019786 weight gain Nutrition 0.000 claims description 2
- 229960003753 nitric oxide Drugs 0.000 description 11
- 238000012958 reprocessing Methods 0.000 description 8
- FLDALJIYKQCYHH-UHFFFAOYSA-N plutonium(IV) oxide Inorganic materials [O-2].[O-2].[Pu+4] FLDALJIYKQCYHH-UHFFFAOYSA-N 0.000 description 5
- 239000000243 solution Substances 0.000 description 5
- 239000002245 particle Substances 0.000 description 4
- 239000007787 solid Substances 0.000 description 4
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 4
- 238000002485 combustion reaction Methods 0.000 description 3
- FCTBKIHDJGHPPO-UHFFFAOYSA-N uranium dioxide Inorganic materials O=[U]=O FCTBKIHDJGHPPO-UHFFFAOYSA-N 0.000 description 3
- KRHYYFGTRYWZRS-UHFFFAOYSA-N Fluorane Chemical compound F KRHYYFGTRYWZRS-UHFFFAOYSA-N 0.000 description 2
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 description 2
- 238000005372 isotope separation Methods 0.000 description 2
- 239000007788 liquid Substances 0.000 description 2
- 229910017604 nitric acid Inorganic materials 0.000 description 2
- OOAWCECZEHPMBX-UHFFFAOYSA-N oxygen(2-);uranium(4+) Chemical compound [O-2].[O-2].[U+4] OOAWCECZEHPMBX-UHFFFAOYSA-N 0.000 description 2
- QPLDLSVMHZLSFG-UHFFFAOYSA-N Copper oxide Chemical compound [Cu]=O QPLDLSVMHZLSFG-UHFFFAOYSA-N 0.000 description 1
- 239000005751 Copper oxide Substances 0.000 description 1
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 1
- XLYOFNOQVPJJNP-PWCQTSIFSA-N Tritiated water Chemical compound [3H]O[3H] XLYOFNOQVPJJNP-PWCQTSIFSA-N 0.000 description 1
- 238000005054 agglomeration Methods 0.000 description 1
- 230000002776 aggregation Effects 0.000 description 1
- 230000003466 anti-cipated effect Effects 0.000 description 1
- 239000007864 aqueous solution Substances 0.000 description 1
- 238000000429 assembly Methods 0.000 description 1
- 230000000712 assembly Effects 0.000 description 1
- 229910000431 copper oxide Inorganic materials 0.000 description 1
- 239000013078 crystal Substances 0.000 description 1
- 230000002939 deleterious effect Effects 0.000 description 1
- 238000004090 dissolution Methods 0.000 description 1
- 238000000227 grinding Methods 0.000 description 1
- 229910052739 hydrogen Inorganic materials 0.000 description 1
- 239000001257 hydrogen Substances 0.000 description 1
- 239000000463 material Substances 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 238000010298 pulverizing process Methods 0.000 description 1
- 238000011084 recovery Methods 0.000 description 1
Classifications
-
- E—FIXED CONSTRUCTIONS
- E21—EARTH OR ROCK DRILLING; MINING
- E21B—EARTH OR ROCK DRILLING; OBTAINING OIL, GAS, WATER, SOLUBLE OR MELTABLE MATERIALS OR A SLURRY OF MINERALS FROM WELLS
- E21B7/00—Special methods or apparatus for drilling
- E21B7/26—Drilling without earth removal, e.g. with self-propelled burrowing devices
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
- G21C19/42—Reprocessing of irradiated fuel
- G21C19/44—Reprocessing of irradiated fuel of irradiated solid fuel
- G21C19/48—Non-aqueous processes
-
- B—PERFORMING OPERATIONS; TRANSPORTING
- B25—HAND TOOLS; PORTABLE POWER-DRIVEN TOOLS; MANIPULATORS
- B25D—PERCUSSIVE TOOLS
- B25D9/00—Portable percussive tools with fluid-pressure drive, i.e. driven directly by fluids, e.g. having several percussive tool bits operated simultaneously
- B25D9/14—Control devices for the reciprocating piston
- B25D9/26—Control devices for adjusting the stroke of the piston or the force or frequency of impact thereof
-
- E—FIXED CONSTRUCTIONS
- E21—EARTH OR ROCK DRILLING; MINING
- E21B—EARTH OR ROCK DRILLING; OBTAINING OIL, GAS, WATER, SOLUBLE OR MELTABLE MATERIALS OR A SLURRY OF MINERALS FROM WELLS
- E21B4/00—Drives for drilling, used in the borehole
- E21B4/06—Down-hole impacting means, e.g. hammers
- E21B4/14—Fluid operated hammers
- E21B4/145—Fluid operated hammers of the self propelled-type, e.g. with a reverse mode to retract the device from the hole
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/28—Treating solids
- G21F9/30—Processing
-
- 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
-
- 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
- Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
- Y02W30/00—Technologies for solid waste management
- Y02W30/50—Reuse, recycling or recovery technologies
Landscapes
- Engineering & Computer Science (AREA)
- Physics & Mathematics (AREA)
- Mining & Mineral Resources (AREA)
- Geology (AREA)
- Life Sciences & Earth Sciences (AREA)
- Fluid Mechanics (AREA)
- Mechanical Engineering (AREA)
- Environmental & Geological Engineering (AREA)
- High Energy & Nuclear Physics (AREA)
- General Life Sciences & Earth Sciences (AREA)
- Geochemistry & Mineralogy (AREA)
- General Engineering & Computer Science (AREA)
- Plasma & Fusion (AREA)
- Automation & Control Theory (AREA)
- Inorganic Compounds Of Heavy Metals (AREA)
- Monitoring And Testing Of Nuclear Reactors (AREA)
- Physical Or Chemical Processes And Apparatus (AREA)
Description
(54) METHOD OF RELEASING FISSION GASES FROM IRRADIATED
NUCLEAR FUEL
(71) We, EXXON NUCLEAR COMPANY INC., a Corporation duly organised and existing under the laws of the State of Delaware, United States of America, of Bellevue,
Washington, United States of America, do hereby declare the invention for which we pray that a patent may be granted to us, and the method by which it is to be performed, to be particularly described in and by the following statement:
This invention relates to a process for releasing volatile fission products from irradiated fuel pellets with a nitrogen oxide oxidant, for example either NO2 or a mixture of NO2 with its dissociation products, 2 and NO.
One of the problems encountered in the reprocessing of irradiated nuclear fuel is the extreme difficulty in separating the volatile fission products from the reaction solutions of the reprocessing process. This is especially important with regard to removing tritium in that the tritium isotope is extremely difficult to separate from aqueous solutions. The hydrogen isotope tritium, has a half life of about 12.26 years and is a ternary fission product produced in small mass quantities, approximately 1/1000 of an oz. in each ton of fuel.
However, because of the volume of irradiated fuel anticipated to be produced in order to satisfy the world's electrical needs for the year 2000, there would be significant volumes of tritium produced. Recovery of tritium in nuclear fuel reprocessing plants of conventional designs is not practical because of the very small amounts of tritium in each ton of fuel.
Tritium, in the form of tritiated water, becomes intimately mixed with the thousands of gallons of process water in the plant. Isotopic separation of the tritium from this large volume of water would be required before the water could be released as liquid or vapor.
Since the isotopic separation of tritium in such volumes of water is not practical, it is necessary to develop a technique for the removal of volatile fission products from irradiated fuel prior to aqueous reprocessing.
The problem of removing volatile fission products from irradiated fuel is encountered when reprocessing irradiated uranium dioxide fuel and/or irradiated mixed oxides, i.e.,
UO2 and PuO2. The problem of removing the volatile fission products is increased by the fact that the volatile fission products are trapped within the crystal lattice of the irradiated fuel and thus cannot be removed by grinding or pulverizing the fuel.
A process for the removal of volatile fisshion products from irradiated fuel has been developed by the Oak Ridge National Laboratories, U.S.A., and is referred to as voloxidation and is described in the Oak Ridge National Laboratories Report ORNL-TM3723. Voloxidation is a process for oxidizing irradiated fuel in the presence of oxygen at a controlled temperature in order to form a very fine powder of U308. The voloxidation process is sensitive to temperature and it has been estimated that in a large commercial plant the reaction temperature would have to be maintained within a narrow range probably at about 480"C. plus or minus 10 C. Because of the critically narrow temperature range, the difficulty in operating large scale reprocessing plants would be increased. The present invention relates to a method for releasing the volatile fission products from irradiated fuel in a commercially reasonable time and under commercially reasonable processing conditions thereby resulting in significant reduction of costs in the reprocessing of irradiated nuclear fuels.
Accordingly this invention concerns a process for releasing from irradiated nuclear fuels volatile fission products, i.e., iodine, xenon, krypton and tritium. Pellets of irradiated nuclear fuel e.g. uranium dioxide, plutonium oxide or mixed oxide fuels e.g. UO2 and PuO2 are oxidized with a nitrogen oxide, e.g. nitrogen dioxide. Nitrogen dioxide, at the reaction temperature of this process, is in equilibrium with its dissociation products, nitrogen monoxide and oxygen. Thus the oxidant may be added as nitrogen dioxide, the dissociation products of nitrogen dioxide (nitrogen monoxide and oxygen) or as a mixture of nitrogen dioxide and its dissociation products. The oxidant may be diluted with a gas, such as nitrogen or nitrogen monoxide without any substantial deleterious effect. The oxidation reaction temperature is from 325"C. to 8000C., preferably between 350"C and 780"C. More preferably, the reaction is maintained at a temperature in the range of 350" to 6500C.
The oxidation reaction is maintained for a time sufficient to convert the irradiated fuel pellets to a fine powder (U3O8 and UO3) preferably having a mean particle size of less than 0.105 millimeters and more preferably less than 0.045 millimeters. Heating to reaction temperatures above 800"C. will result in an agglomeration of the fine powder and thus is to be avoided.
The irradiated fuel assemblies after removal from the reactor are generally permitted to cool and then the fuel rods are cut open or chopped-up in order to prepare the irradiated fuel for reprocessing. The irradiated nuclear fuel is then placed in a reactor vessel and nitrogen dioxide alone or in admixture with its dissociation products is passed through the reactor. The system is heated to a reaction temperature between 325" and 800"C. During the oxidation reaction the fission gases, particularly tritium, in nearly quantitative amounts, are released to the off-gas system. During the reaction, UO2 is oxidized to form UO3 and/or U3Os.
The time of the reaction will vary depending upon the volume of material in the reactor, fuel chop size, UO2 particle size, temperature and gas composition. The fine powder will usually have a mean particle diameter of less than 0.105 millimeters and preferably a mean particle diameter of less than 0.045 millimeters.
The released tritium may be collected as tritium gas or it may be oxidized to form THO, cooled and collected as a liquid (TNO3, HTO etc.).
In order to demonstrate the effectiveness of this invention, fuel pellets were oxidized in ordinary laboratory equipment. Fuel pellets were placed in a sample boat and inserted into a combustion tube. A thermocouple was used for temperature control. A tube furnace was placed around the combustion tube to provide heat to the system. The oxidant was passed through the combustion tube as a gas. Tritium liberated during the oxidation was converted to THO in a copper oxide furnace and condensed in a cold finger.
Example I
One gram samples of an irradiated uranium oxide (UO2) fuel and a mixed oxide (UO2/PuO2) fuel were each oxidized for four hours with NO2 at 4000C. At the end of the four hour period, the irradiated fuels were in the form of very fine powders. The resulting powders were dissolved in 15 milliliters of 8 molar HNO3 for between three and four hours at 100"C., filtered and sampled for tritium remaining in solution. An additional one gram sample of each of the irradiated fuels, without NO2 oxidation, was dissolved directly in 8 molar HNO3 for about four hours at 1000C., filtered and sampled for tritium to provide a standard sample. The residual filtered solids for both the standard and test samples were further dissolved with 8 molar HNOX plus 0.005 molar hydrofluoric acid, and the resulting solution sampled to determine the tritium content of the solids. The results of the oxidation of the two fuel samples are set forth in Table I below:
TABLE 1
Tritium Found, Fci/g UO2 H Percent Tritium as
Fuel Dissolver Solution Solids compared to Standard
Standard UO2 159.0 0.05
NO2 Oxidized UO2 0.17 0.12 0.2% Standard UO,/Pu02 20.4 O.11 11 NO2 oxidized UO2/PuO2 0.64 0.4 5%
It is to be secn from the above results that the NO2 oxidation of the fuel prior to dissolution effectively removed the tritium from the fuel. The mixed oxide dissolver solution and solids contained 5% of the initial tritium content.
Example 2
A series of runs were made under isothermal conditions at temperatures of 300"C to 800"C. using samples of UO2 fuel pellets to determine the rate of the reaction. When the oxidation was complete the fuel pellet was converted to a finely divided powder and the fuel sample showed a weight gain of 4 weight percent. The following results obtained are shown in Figure 1 of the drawings. For runs at temperatures of 325"C. to 600"C. the final product was a powder. Runs made at 350 , 500 and 600" appear to be approximately equivalent.
Example 3
A series of experiments at constant rates of heating of 300"C per hour were made to test the effect of diluting the oxidant, NO2, with nitrogen with results as shown in Figure 2 of the drawings in which the mixtures of NO2 and N2 are by volume. The fuel samples were UO2 pellets. It is to be noted that the rate of reaction for UO2 oxidation in NO2 is relatively insensitive to N2 dilution.
Example 4
To determine the influence of NO dilution on UO2 oxidation, an experiment at a constant rate of heating of 300"C. per hour was made with NO2 and NO added as the oxidant mixture with results as shown in Figure 3 of the drawings.
The oxidant used was a 1:1 volume mixture of NO2 and NO. It is observed that increased amounts of NO tend to reduce the rate of reaction for UO2 oxidation in NO2.
WHAT WE CLAIM IS:
1. A process for releasting volatile fission products from irradiated nuclear fuel which comprises reacting irradiated nuclear fuel pellets with a nitrogen oxide oxidant at a reaction temperature between 325" and 800"C for a time sufficient to form a fine powder, thereby releasing the volatile fission products.
2. A process according to claim 1 wherein the irradiated fuel is uranium oxide, plutonium oxide or a mixture thereof.
3. A process according to either of claims 1 and 2 wherein the reaction temperature is maintained between 350" and 650"C.
4. A process according to any one of the preceding claims wherein the volatile fission products are xenon, iodine, krypton and tritium.
5. A process according to any one of the preceding claims wherein the oxidant is nitrogen dioxide.
6. A process according to claim 5 wherein the nitrogen dioxide is in equilibrium with its dissociation products, oxygen and nitrogen monoxide.
7. A process according to any one of claims 1 to 4 wherein the oxidant consists of a mixture of nitrogen dioxide, oxygen and nitrogen monoxide.
8. A process according to any one of claims 1 to 4 wherein the oxidant is a mixture of nitrogen dioxide dissociation products, or nitrogen monoxide and oxygen.
9. A process according to any one of the preceding claims wherein the oxidant is diluted with a gas.
10. A process according to claim 9 wherein the gas is nitrogen.
11. A process according to claim 9 wherein the gas is nitrogen monoxide.
12. A process for releasing volatile fission products from irradiated nuclear fuel according to claim 1 substantially as hereinbefore described with reference to the
Examples.
13. Volatile fission products whenever released from irradiated nuclear fuel by the process according to any one of the preceding claims.
**WARNING** end of DESC field may overlap start of CLMS **.
Claims (13)
1. A process for releasting volatile fission products from irradiated nuclear fuel which comprises reacting irradiated nuclear fuel pellets with a nitrogen oxide oxidant at a reaction temperature between 325" and 800"C for a time sufficient to form a fine powder, thereby releasing the volatile fission products.
2. A process according to claim 1 wherein the irradiated fuel is uranium oxide, plutonium oxide or a mixture thereof.
3. A process according to either of claims 1 and 2 wherein the reaction temperature is maintained between 350" and 650"C.
4. A process according to any one of the preceding claims wherein the volatile fission products are xenon, iodine, krypton and tritium.
5. A process according to any one of the preceding claims wherein the oxidant is nitrogen dioxide.
6. A process according to claim 5 wherein the nitrogen dioxide is in equilibrium with its dissociation products, oxygen and nitrogen monoxide.
7. A process according to any one of claims 1 to 4 wherein the oxidant consists of a mixture of nitrogen dioxide, oxygen and nitrogen monoxide.
8. A process according to any one of claims 1 to 4 wherein the oxidant is a mixture of nitrogen dioxide dissociation products, or nitrogen monoxide and oxygen.
9. A process according to any one of the preceding claims wherein the oxidant is diluted with a gas.
10. A process according to claim 9 wherein the gas is nitrogen.
11. A process according to claim 9 wherein the gas is nitrogen monoxide.
12. A process for releasing volatile fission products from irradiated nuclear fuel according to claim 1 substantially as hereinbefore described with reference to the
Examples.
13. Volatile fission products whenever released from irradiated nuclear fuel by the process according to any one of the preceding claims.
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| US76614277A | 1977-02-07 | 1977-02-07 |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| GB1593323A true GB1593323A (en) | 1981-07-15 |
Family
ID=25075540
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| GB2020/78A Expired GB1593323A (en) | 1977-02-07 | 1978-01-18 | Method of releasing fission gases from irradiated nuclear fuel |
Country Status (8)
| Country | Link |
|---|---|
| JP (1) | JPS6034718B2 (en) |
| BE (1) | BE863675A (en) |
| DE (1) | DE2801744C2 (en) |
| ES (1) | ES466675A1 (en) |
| FR (1) | FR2379884A1 (en) |
| GB (1) | GB1593323A (en) |
| IT (1) | IT1092476B (en) |
| SE (1) | SE429587B (en) |
Families Citing this family (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPH0439019U (en) * | 1990-08-01 | 1992-04-02 |
Family Cites Families (3)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3140151A (en) * | 1959-11-12 | 1964-07-07 | James R Foltz | Method of reprocessing uo2 reactor fuel |
| BE611199A (en) * | 1961-12-06 | 1962-06-06 | Ct D Etude De L En Nucleaire | Process for the reprocessing of nuclear fuels, which include carbon in their structure and / or in their cladding. |
| IT1034322B (en) * | 1975-03-17 | 1979-09-10 | Agip Nucleare Spa | PYROCHEMICAL SEPARATION OF PLUTUS NIUM FROM IRRAYED NUCLEAR FUELS BY THERMODECOMPOSITION IN MELTED NITRATES |
-
1978
- 1978-01-16 DE DE2801744A patent/DE2801744C2/en not_active Expired
- 1978-01-18 GB GB2020/78A patent/GB1593323A/en not_active Expired
- 1978-01-30 IT IT1980278A patent/IT1092476B/en active
- 1978-01-31 JP JP53008985A patent/JPS6034718B2/en not_active Expired
- 1978-02-06 BE BE184922A patent/BE863675A/en not_active IP Right Cessation
- 1978-02-06 ES ES466675A patent/ES466675A1/en not_active Expired
- 1978-02-06 SE SE7801361A patent/SE429587B/en not_active IP Right Cessation
- 1978-02-06 FR FR7803277A patent/FR2379884A1/en active Granted
Also Published As
| Publication number | Publication date |
|---|---|
| JPS5399196A (en) | 1978-08-30 |
| BE863675A (en) | 1978-08-07 |
| IT1092476B (en) | 1985-07-12 |
| IT7819802A0 (en) | 1978-01-30 |
| SE429587B (en) | 1983-09-12 |
| FR2379884B1 (en) | 1984-10-19 |
| FR2379884A1 (en) | 1978-09-01 |
| SE7801361L (en) | 1978-08-08 |
| DE2801744A1 (en) | 1978-08-10 |
| ES466675A1 (en) | 1979-11-16 |
| DE2801744C2 (en) | 1986-12-11 |
| JPS6034718B2 (en) | 1985-08-10 |
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