GB2556944A - Use of decay heat from spent nuclear fuel processed by electro-reduction - Google Patents
Use of decay heat from spent nuclear fuel processed by electro-reduction Download PDFInfo
- Publication number
- GB2556944A GB2556944A GB1620110.5A GB201620110A GB2556944A GB 2556944 A GB2556944 A GB 2556944A GB 201620110 A GB201620110 A GB 201620110A GB 2556944 A GB2556944 A GB 2556944A
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- United Kingdom
- Prior art keywords
- molten salt
- molten
- heat
- electrolyte
- fuel
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- 239000002915 spent fuel radioactive waste Substances 0.000 title claims abstract description 21
- 150000003839 salts Chemical class 0.000 claims abstract description 70
- 238000000034 method Methods 0.000 claims abstract description 40
- 230000008569 process Effects 0.000 claims abstract description 33
- 239000003792 electrolyte Substances 0.000 claims abstract description 31
- 239000000446 fuel Substances 0.000 claims abstract description 21
- 229910052768 actinide Inorganic materials 0.000 claims abstract description 19
- 229910052751 metal Inorganic materials 0.000 claims abstract description 19
- 150000001255 actinides Chemical class 0.000 claims abstract description 18
- 239000002184 metal Substances 0.000 claims abstract description 16
- 229910045601 alloy Inorganic materials 0.000 claims abstract description 15
- 239000000956 alloy Substances 0.000 claims abstract description 15
- 238000005868 electrolysis reaction Methods 0.000 claims abstract description 14
- 229910052770 Uranium Inorganic materials 0.000 claims abstract description 13
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 claims abstract description 13
- 229910052778 Plutonium Inorganic materials 0.000 claims abstract description 12
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 claims abstract description 12
- 230000005611 electricity Effects 0.000 claims abstract description 8
- 239000012530 fluid Substances 0.000 claims abstract description 7
- 239000003758 nuclear fuel Substances 0.000 claims abstract description 6
- 230000002285 radioactive effect Effects 0.000 claims abstract description 6
- SAWLVFKYPSYVBL-UHFFFAOYSA-K uranium(iii) chloride Chemical compound Cl[U](Cl)Cl SAWLVFKYPSYVBL-UHFFFAOYSA-K 0.000 claims abstract description 6
- 229910000711 U alloy Inorganic materials 0.000 claims abstract description 5
- 238000002844 melting Methods 0.000 claims abstract description 5
- 230000008018 melting Effects 0.000 claims abstract description 5
- 238000005253 cladding Methods 0.000 claims abstract description 3
- 230000004992 fission Effects 0.000 claims description 9
- 229910052747 lanthanoid Inorganic materials 0.000 claims description 9
- 150000002602 lanthanoids Chemical class 0.000 claims description 9
- 239000002826 coolant Substances 0.000 claims description 6
- 230000005855 radiation Effects 0.000 claims description 6
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 claims description 5
- 239000000463 material Substances 0.000 claims description 5
- 239000000203 mixture Substances 0.000 claims description 5
- 239000002699 waste material Substances 0.000 claims description 5
- WCUXLLCKKVVCTQ-UHFFFAOYSA-M Potassium chloride Chemical compound [Cl-].[K+] WCUXLLCKKVVCTQ-UHFFFAOYSA-M 0.000 claims description 4
- AIYUHDOJVYHVIT-UHFFFAOYSA-M caesium chloride Chemical compound [Cl-].[Cs+] AIYUHDOJVYHVIT-UHFFFAOYSA-M 0.000 claims description 4
- FGDZQCVHDSGLHJ-UHFFFAOYSA-M rubidium chloride Chemical compound [Cl-].[Rb+] FGDZQCVHDSGLHJ-UHFFFAOYSA-M 0.000 claims description 4
- 239000010935 stainless steel Substances 0.000 claims description 4
- 229910001220 stainless steel Inorganic materials 0.000 claims description 4
- 238000003860 storage Methods 0.000 claims description 4
- 238000009825 accumulation Methods 0.000 claims description 3
- 229910001092 metal group alloy Inorganic materials 0.000 claims description 3
- UXVMQQNJUSDDNG-UHFFFAOYSA-L Calcium chloride Chemical group [Cl-].[Cl-].[Ca+2] UXVMQQNJUSDDNG-UHFFFAOYSA-L 0.000 claims description 2
- OKTJSMMVPCPJKN-UHFFFAOYSA-N Carbon Chemical group [C] OKTJSMMVPCPJKN-UHFFFAOYSA-N 0.000 claims description 2
- UGFAIRIUMAVXCW-UHFFFAOYSA-N Carbon monoxide Chemical class [O+]#[C-] UGFAIRIUMAVXCW-UHFFFAOYSA-N 0.000 claims description 2
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 claims description 2
- WDIHJSXYQDMJHN-UHFFFAOYSA-L barium chloride Chemical compound [Cl-].[Cl-].[Ba+2] WDIHJSXYQDMJHN-UHFFFAOYSA-L 0.000 claims description 2
- 229910001626 barium chloride Inorganic materials 0.000 claims description 2
- 239000001110 calcium chloride Substances 0.000 claims description 2
- 229910001628 calcium chloride Inorganic materials 0.000 claims description 2
- 229910002090 carbon oxide Inorganic materials 0.000 claims description 2
- 229910002804 graphite Inorganic materials 0.000 claims description 2
- 239000010439 graphite Substances 0.000 claims description 2
- 229910052760 oxygen Inorganic materials 0.000 claims description 2
- 239000001301 oxygen Substances 0.000 claims description 2
- 239000001103 potassium chloride Substances 0.000 claims description 2
- 235000011164 potassium chloride Nutrition 0.000 claims description 2
- 229940102127 rubidium chloride Drugs 0.000 claims description 2
- 239000011780 sodium chloride Substances 0.000 claims description 2
- 229910001631 strontium chloride Inorganic materials 0.000 claims description 2
- AHBGXTDRMVNFER-UHFFFAOYSA-L strontium dichloride Chemical compound [Cl-].[Cl-].[Sr+2] AHBGXTDRMVNFER-UHFFFAOYSA-L 0.000 claims description 2
- -1 actinide metals Chemical class 0.000 claims 1
- 229910052741 iridium Inorganic materials 0.000 claims 1
- GKOZUEZYRPOHIO-UHFFFAOYSA-N iridium atom Chemical group [Ir] GKOZUEZYRPOHIO-UHFFFAOYSA-N 0.000 claims 1
- 150000002823 nitrates Chemical class 0.000 claims 1
- 150000002826 nitrites Chemical class 0.000 claims 1
- SIWVEOZUMHYXCS-UHFFFAOYSA-N oxo(oxoyttriooxy)yttrium Chemical compound O=[Y]O[Y]=O SIWVEOZUMHYXCS-UHFFFAOYSA-N 0.000 claims 1
- 229910052715 tantalum Inorganic materials 0.000 claims 1
- GUVRBAGPIYLISA-UHFFFAOYSA-N tantalum atom Chemical compound [Ta] GUVRBAGPIYLISA-UHFFFAOYSA-N 0.000 claims 1
- WFKWXMTUELFFGS-UHFFFAOYSA-N tungsten Chemical compound [W] WFKWXMTUELFFGS-UHFFFAOYSA-N 0.000 claims 1
- 229910052721 tungsten Inorganic materials 0.000 claims 1
- 239000010937 tungsten Substances 0.000 claims 1
- 238000012958 reprocessing Methods 0.000 abstract description 9
- 239000008188 pellet Substances 0.000 description 7
- 238000004519 manufacturing process Methods 0.000 description 5
- 239000007789 gas Substances 0.000 description 3
- 229910000510 noble metal Inorganic materials 0.000 description 3
- 229910000831 Steel Inorganic materials 0.000 description 2
- GKLVYJBZJHMRIY-OUBTZVSYSA-N Technetium-99 Chemical compound [99Tc] GKLVYJBZJHMRIY-OUBTZVSYSA-N 0.000 description 2
- MCMNRKCIXSYSNV-UHFFFAOYSA-N Zirconium dioxide Chemical compound O=[Zr]=O MCMNRKCIXSYSNV-UHFFFAOYSA-N 0.000 description 2
- 238000009434 installation Methods 0.000 description 2
- 230000009467 reduction Effects 0.000 description 2
- 102200052313 rs9282831 Human genes 0.000 description 2
- 239000010959 steel Substances 0.000 description 2
- 229910052712 strontium Inorganic materials 0.000 description 2
- CIOAGBVUUVVLOB-UHFFFAOYSA-N strontium atom Chemical compound [Sr] CIOAGBVUUVVLOB-UHFFFAOYSA-N 0.000 description 2
- 101001023271 Homo sapiens Laminin subunit gamma-2 Proteins 0.000 description 1
- DGAQECJNVWCQMB-PUAWFVPOSA-M Ilexoside XXIX Chemical compound C[C@@H]1CC[C@@]2(CC[C@@]3(C(=CC[C@H]4[C@]3(CC[C@@H]5[C@@]4(CC[C@@H](C5(C)C)OS(=O)(=O)[O-])C)C)[C@@H]2[C@]1(C)O)C)C(=O)O[C@H]6[C@@H]([C@H]([C@@H]([C@H](O6)CO)O)O)O.[Na+] DGAQECJNVWCQMB-PUAWFVPOSA-M 0.000 description 1
- 102100035159 Laminin subunit gamma-2 Human genes 0.000 description 1
- 229910002651 NO3 Inorganic materials 0.000 description 1
- 229910052781 Neptunium Inorganic materials 0.000 description 1
- NHNBFGGVMKEFGY-UHFFFAOYSA-N Nitrate Chemical compound [O-][N+]([O-])=O NHNBFGGVMKEFGY-UHFFFAOYSA-N 0.000 description 1
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 description 1
- 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 description 1
- 230000033228 biological regulation Effects 0.000 description 1
- 230000015572 biosynthetic process Effects 0.000 description 1
- 229910052792 caesium Inorganic materials 0.000 description 1
- TVFDJXOCXUVLDH-UHFFFAOYSA-N caesium atom Chemical compound [Cs] TVFDJXOCXUVLDH-UHFFFAOYSA-N 0.000 description 1
- BRPQOXSCLDDYGP-UHFFFAOYSA-N calcium oxide Chemical compound [O-2].[Ca+2] BRPQOXSCLDDYGP-UHFFFAOYSA-N 0.000 description 1
- 239000000292 calcium oxide Substances 0.000 description 1
- ODINCKMPIJJUCX-UHFFFAOYSA-N calcium oxide Inorganic materials [Ca]=O ODINCKMPIJJUCX-UHFFFAOYSA-N 0.000 description 1
- 159000000006 cesium salts Chemical class 0.000 description 1
- 230000008859 change Effects 0.000 description 1
- 238000012993 chemical processing Methods 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 239000003795 chemical substances by application Substances 0.000 description 1
- 150000003841 chloride salts Chemical class 0.000 description 1
- 238000009826 distribution Methods 0.000 description 1
- 238000005516 engineering process Methods 0.000 description 1
- 239000000284 extract Substances 0.000 description 1
- 150000004673 fluoride salts Chemical class 0.000 description 1
- 230000020169 heat generation Effects 0.000 description 1
- 238000005338 heat storage Methods 0.000 description 1
- 238000010438 heat treatment Methods 0.000 description 1
- 238000009413 insulation Methods 0.000 description 1
- 230000007774 longterm Effects 0.000 description 1
- 229910044991 metal oxide Inorganic materials 0.000 description 1
- 150000004706 metal oxides Chemical class 0.000 description 1
- LFNLGNPSGWYGGD-UHFFFAOYSA-N neptunium atom Chemical compound [Np] LFNLGNPSGWYGGD-UHFFFAOYSA-N 0.000 description 1
- 229910052756 noble gas Inorganic materials 0.000 description 1
- 230000009972 noncorrosive effect Effects 0.000 description 1
- 239000003921 oil Substances 0.000 description 1
- 230000003647 oxidation Effects 0.000 description 1
- 238000007254 oxidation reaction Methods 0.000 description 1
- FGIUAXJPYTZDNR-UHFFFAOYSA-N potassium nitrate Chemical class [K+].[O-][N+]([O-])=O FGIUAXJPYTZDNR-UHFFFAOYSA-N 0.000 description 1
- 235000010333 potassium nitrate Nutrition 0.000 description 1
- PDEDQSAFHNADLV-UHFFFAOYSA-M potassium;disodium;dinitrate;nitrite Chemical class [Na+].[Na+].[K+].[O-]N=O.[O-][N+]([O-])=O.[O-][N+]([O-])=O PDEDQSAFHNADLV-UHFFFAOYSA-M 0.000 description 1
- 238000012545 processing Methods 0.000 description 1
- 230000035755 proliferation Effects 0.000 description 1
- 238000003608 radiolysis reaction Methods 0.000 description 1
- 238000011084 recovery Methods 0.000 description 1
- 238000007670 refining Methods 0.000 description 1
- 239000011435 rock Substances 0.000 description 1
- 229910052708 sodium Inorganic materials 0.000 description 1
- 239000011734 sodium Substances 0.000 description 1
- 235000010344 sodium nitrate Nutrition 0.000 description 1
- 239000007787 solid Substances 0.000 description 1
- 239000011343 solid material Substances 0.000 description 1
- 239000000126 substance Substances 0.000 description 1
- 229910000439 uranium oxide Inorganic materials 0.000 description 1
- 239000002918 waste heat Substances 0.000 description 1
- 229910052726 zirconium Inorganic materials 0.000 description 1
Classifications
-
- 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
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21H—OBTAINING ENERGY FROM RADIOACTIVE SOURCES; APPLICATIONS OF RADIATION FROM RADIOACTIVE SOURCES, NOT OTHERWISE PROVIDED FOR; UTILISING COSMIC RADIATION
- G21H3/00—Arrangements for direct conversion of radiation energy from radioactive sources into forms of energy other than electric energy, e.g. into light or mechanic energy
- G21H3/02—Arrangements for direct conversion of radiation energy from radioactive sources into forms of energy other than electric energy, e.g. into light or mechanic energy in which material is excited to luminesce by the radiation
-
- 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
- Physics & Mathematics (AREA)
- Engineering & Computer Science (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Plasma & Fusion (AREA)
- Electrolytic Production Of Metals (AREA)
Abstract
A method of reprocessing spent oxide nuclear fuel is described comprising: removing the spent oxide fuel from its cladding; electrolytically reducing the spent fuel to an alloy of uranium and other metallic elements in a molten salt at a temperature above the melting point of uranium; continuing the electrolysis without adding further oxide fuel until the majority of plutonium and other actinides have been reduced to metal and have mixed into the molten alloy; and then transferring the molten alloy to a molten metal/molten salt contacting apparatus where the plutonium and higher actinides are extracted into a further molten salt containing uranium trichloride. The process may be repeated multiple times without changing the molten salt electrolyte, the resulting highly radioactive heat generating salt being used to heat a working fluid which is then used to generate electricity. The further molten salt may also be used as fuel for a molten salt nuclear reactor.
Description
(71) Applicant(s):
Ian Richard Scott
Lambourne House, Lower Binton,
STRATFORD UPON AVON, Warwickshire, CV37 9TQ, United Kingdom (72) Inventor(s):
Ian Richard Scott (74) Agent and/or Address for Service:
Ian Richard Scott
Lambourne House, Lower Binton,
STRATFORD UPON AVON, Warwickshire, CV37 9TQ, United Kingdom (51) INT CL:
G21C 19/44 (2006.01) (56) Documents Cited:
US 3298935 A US 3052611 A
Progress in Natural Science: Material International Vol. 25, December 2015, E-Y Choi et al, Electrochemical processing of spent nuclear fuels: An overview of oxide reduction in pyroprocessing technology, pages 572-582 (58) Field of Search:
INT CL G21C, G21F
Other: WPI; EPODOC; Patent fulltext (54) Title of the Invention: Use of decay heat from spent nuclear fuel processed by electro-reduction Abstract Title: Reprocessing spent oxide type nuclear fuel by electro-reduction (57) A method of reprocessing spent oxide nuclear fuel is described comprising: removing the spent oxide fuel from its cladding; electrolytically reducing the spent fuel to an alloy of uranium and other metallic elements in a molten salt at a temperature above the melting point of uranium; continuing the electrolysis without adding further oxide fuel until the majority of plutonium and other actinides have been reduced to metal and have mixed into the molten alloy; and then transferring the molten alloy to a molten metal/molten salt contacting apparatus where the plutonium and higher actinides are extracted into a further molten salt containing uranium trichloride. The process may be repeated multiple times without changing the molten salt electrolyte, the resulting highly radioactive heat generating salt being used to heat a working fluid which is then used to generate electricity. The further molten salt may also be used as fuel for a molten salt nuclear reactor.
Waste uranium alloy out
1/2
Off gas to Oxide pellet feed condenser I
Zirconia insulation
2/2
Heat producing salt in Nitrate (solar) salt heat
in
Π3
Φ t/S
W>
• w
Use of decay heat from spent nuclear fuel processed by electro-reduction
Background
Spent fuel from nuclear reactors continues to produce heat for many years after the nuclear fission process has ended. There have been many suggestions as to how to use this decay heat for productive purposes but all have failed to have any commercial relevance due to a combination of factors.
• Spent fuel in the form it emerges from the reactor rapidly decays to a point after a few years where the heat output is of the order of lkW per tonne of fuel making any meaningful power generator impractically large.
• Spent fuel typically is in a form where it cannot be exposed to high temperatures without risk of leakage, limiting the temperature of whatever medium is used to transfer to heat.
• Spent fuel emits high levels of radiation so any installation must be extensively screened to protect people and the environment • Spent fuel contains plutonium so any installation using it becomes subject to IAEA nuclear safeguards standards which impose an impractically high cost on the relatively small amounts of electricity generated.
• Chemical processing of spent fuel to separate the small quantity of heat generating fission products from the large volume of non heat producing uranium is expensive and typically involves using a process that can produce weapons usable plutonium.
The combination of these factors have prevented any attempt to use the decay heat from spent nuclear fuel on a commercial scale.
Description of the invention
A novel method of reprocessing spent uranium oxide fuel has been devised where the oxide pellets are electrochemically reduced in a suitable electrolyte operating above the melting point of uranium. The high temperature is the key innovation as it prevents formation of layers of metallic uranium in the outer regions of the oxide pellet which otherwise prevent the oxide at the centre of the pellet from being reduced to the metal. Previous attempts to use electrolytic reduction on spent nuclear fuel have been broadly unsuccessful due to this problem of incomplete oxidation of solid oxides at lower temperatures as reviewed in Progress in Natural Science. Materials International,vol25, 572-582 (2015) see end of section
2.1
The electrolyte can be any molten salt that allows significant levels of metal oxides to be dissolved in the salt. Calcium chloride, containing small amounts of calcium oxide is suitable, but salts or mixtures of salts such as barium chloride, strontium chloride, sodium chloride, potassium chloride, caesium chloride, rubidium chloride are also suitable and may be preferred in certain circumstance. Strontium and cesium salts for example are already present in spent nuclear fuel as fission products and therefore do not introduce new elements to the system. Fluoride salts may be used instead of chloride salts.
This reprocessing method is radically simpler and therefore cheaper than aqueous based reprocessing such as the PUREX process or other molten salt based processes which require electrochemical transfer of uranium from one electrode to another as a key part of the process (eg Argonne National Lab reprocessing of EBR2 metallic reactor fuel).
An exemplary apparatus is shown in figure 1. Spent nuclear fuel pellets are fed into the apparatus continually and reduced to uranium metal by passage of electric current from anode to cathode. Initially the molten metal alloy produced contains uranium, neptunium and the noble metals but when sufficient molten alloy has accumulated, addition of fuel pellets is stopped and electrolysis continued. A large portion of the plutonium and other actinides and a fraction of the lanthanides are reduced to metal at this time and mix into the molten alloy.
This alloy is drained from the electro-reduction cell through a counter flow column where it is contacted with a molten salt containing uranium trichloride which extracts the plutonium and higher actinides, and a fraction of the lanthanides from the molten alloy which is then stored for later use. The molten salt from this process is suitable for use in a molten salt fuelled nuclear reactor such as that described in GB2508537.
The process is then repeated many times with fresh batches of spent oxide fuel (though metallic fuel may be added or used instead of oxide fuel) until the heat production or chemical change from the accumulation of fission products in the electrolyte reaches a level where replacement of the electrolyte is required. At this point addition of fuel pellets is stopped and a more exhaustive electrolysis carried out so that essentially all the remaining actinides and the majority of the lanthanides are reduced to metal. This alloy is passed down a counter flow apparatus contacting a molten salt which may be uranium trichloride or another suitable salt so as to extract much of the lanthanides but substantially none of the actinides into the salt and to transfer substantially all the uranium from the salt into the alloy. The salt is either stored for recovery of the lanthanides or returned to the electrolysis apparatus to mix with the spent electrolyte. The alloy containing the actinides, including the uranium transferred from the uranium trichloride is stored to be added back to the apparatus with a fresh batch of electrolyte. The spent electrolyte is drained from the electrolysis cell for disposal, or as set out in an embodiment of the current invention to be used as a useful heat source.
This process produces three outputs • a gas stream of the noble gas fission products plus oxygen or carbon oxides (depending on whether an inert of graphite electrode is used), • a metal alloy comprising uranium, other actinides, some lanthanides and all the noble and semi-noble metal fission products • a molten salt stream containing the majority of the fission products but substantially no actinides
The process has a number of advantages over other spent fuel reprocessing methods.
• It is very simple and capable of high throughput and hence low cost. This is driven by a number of factors but of especial importance is the tolerance of the electro-reduction cell to high current densities which are an order of magnitude higher than are possible in alternative electro-refining cells where uranium is transferred between electrodes.
• It is relatively poor at separating plutonium from other species, especially lanthanides. As such it cannot be reconfigured to produce plutonium of a purity that can be used in a nuclear weapon. It is therefore substantially more proliferation resistant than alternative reprocessing methods.
• It produces an actinide free waste stream containing the majority of the fission products except for the noble metals including technetium-99. This is important for two reasons. The absence of actinides means it is not subject to IAEA nuclear safeguard regulations but only to the safety and security requirements for storage of nuclear waste. The absence of Tc-99 and actinides means that its radioactivity falls rapidly when the heat producing caesium and strontium have decayed making its long term disposal much simpler.
The spent electrolyte has a high heat production capacity. Depending on the amount of spent nuclear fuel processed through it, this may be as high as 100-500 times the heat production per tonne of the input spent fuel. Depending on the time since the spent fuel was removed from the reactor this could be of the order of lOOkW per tonne.
The spent electrolyte can optionally be rendered substantially non corrosive to standard steels by addition of small amounts of metallic zirconium or other reactive metal and packaged in steel (preferably stainless steel) containers which can be welded closed with a gas space inside to allow for expansion of the salt.
This salt has a heat generation capacity high enough to be of practical value but has a high emission of radiation. For optimum power production, its temperature must be around 600C or more so that it can power a standard high efficiency superheated steam cycle, though other options for conversion of the heat to electric power such as Stirling engines, thermoelectric apparatus, Brayton cycle engines etc may be preferred for certain purposes. It must also be effectively screened against its emissions of gamma radiation.
A suitable apparatus to utilise this heat source is shown in figure 2. It is a tank containing a heat storage medium in which the heat generating spent electrolyte salt containers are immersed. Particularly useful media are “solar salt” which is a mixture of sodium and potassium nitrates or alternative salts such as Hitec salt. The molten solar salt can be bulked out with cheap solid materials such as crushed rock. Other media such as high temperature oils are also suitable though radiolysis by the gamma radiation from the heat source would limit their practical life.
Heat exchangers around the periphery of the tank allow the heat in the solar salt to be transferred to a working fluid for the turbo-generator set. Flow of solar salt in the tank is ideally via natural convection resulting in no moving parts internal to the tank though pumped flow is an option. The large volume of the solar salt provides an effective radiation screen but also gives the apparatus a high thermal inertia so that if the heat source emits heat at, say, lOOkW continually, the turbogenerator can withdraw heat at double that rate for half of the day. This power output variation is especially important when an apparatus such as this is used in remote communities where other methods of production/demand management are difficult to achieve.
Waste heat from the turbo-generator can be discarded or used for heating if a distribution system exists.
The coolant in the tank can also be used directly as a source of process heat.
Claims (13)
1) A process for separating spent nuclear oxide type fuel into a new nuclear fuel and waste products comprising • removing spent oxide fuel from its metal cladding, • reducing the spent fuel to an alloy of uranium with other metallic elements by electrolysis in a molten salt in the presence of the oxide fuel at a temperature above the melting point of uranium, • continuing the electrolysis process without adding further oxide fuel until the majority of the plutonium has been reduced to metal, • transferring the molten uranium alloy to a molten metal/molten salt contacting apparatus where the plutonium is transferred into the molten salt
2) The process of claim 1 where the molten salt in the contacting apparatus includes uranium trichloride
3) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and a final electrolysis in the absence of added oxide fuel is used to reduce essentially all the actinides in the electrolyte to a molten alloy after which the molten salt electrolyte is removed for storage or further use
4) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and a final electrolysis in the absence of added oxide fuel is used to reduce essentially all the actinides in the electrolyte to a molten alloy after which the molten salt electrolyte is removed for storage or further use and fresh molten salt electrolyte added together with more oxide spent fuel, this process resulting in accumulation of the lanthanide fission products in the electrolyte
5) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and a further electrolysis in the absence of added oxide fuel is used to reduce essentially all the actinides in the electrolyte to a molten alloy after which the molten salt electrolyte is removed for storage or further use and the molten alloy passed through a molten metal/molten salt contactor to transfer some non actinide metals from the molten metal, the remaining metal alloy being returned to the electrolysis cell together with fresh molten salt electrolyte and spent oxide fuel for continued use thereby preventing accumulation of lanthanides in the electrolysis cell while avoiding actinides entering the waste streams from the process.
6) The process of claim 1 where the molten salt is calcium chloride, strontium chloride, barium chloride, sodium chloride, potassium chloride, caesium chloride, rubidium chloride or a mixture thereof
7) The process of claim 1 where the electrolysis apparatus is lined with or constructed from tantalum or tungsten metal optionally coated with yttrium oxide
8) The process of claim 1 where the anode is graphite and carbon oxides are generated consuming the anode
9) The process of claim 1 where the anode is iridium and oxygen is generated without consuming the anode
10) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and the resulting highly radioactive heat generating salt is used in an apparatus transferring the heat to a working fluid which is then used to generate electricity or provide process heat
11) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and the resulting highly radioactive heat generating salt is used in an apparatus transferring the heat to a working fluid which is then used to generate electricity or provide process heat where the molten salt electrolyte is enclosed in leak proof thermally conductive containers of stainless steel or other material resistant to molten salt attack and immersed in a tank of coolant which is capable of storing the heat output of the heat generating salt for a period of hours allowing temporally flexible use of the heat for electricity generation or process heat.
12) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and the resulting highly radioactive heat generating salt is used in an apparatus transferring the heat to a working fluid which is then used to generate electricity or provide process heat where the molten salt electrolyte is enclosed in leak proof thermally conductive containers of stainless steel or other material resistant to molten salt attack and immersed in a tank of coolant where the coolant is a molten salt of low melting point such as a mixture of nitrates and optionally nitrites
13) The process of claim 1 where the steps are repeated multiple times without changing the molten salt electrolyte and the resulting highly radioactive heat generating salt is used in an apparatus transferring the heat to a working fluid which is then used to generate electricity or provide process heat where the molten salt electrolyte is enclosed in leak proof thermally conductive containers of stainless steel or other material resistant to molten salt attack and immersed in a tank of coolant where the thickness of the layer of coolant surrounding the heat generating salt containers is sufficient to effectively screen the radiation from said containers.
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Application No: Claims searched:
GB1620110.5
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Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| GB1620110.5A GB2556944A (en) | 2016-11-28 | 2016-11-28 | Use of decay heat from spent nuclear fuel processed by electro-reduction |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| GB1620110.5A GB2556944A (en) | 2016-11-28 | 2016-11-28 | Use of decay heat from spent nuclear fuel processed by electro-reduction |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| GB201620110D0 GB201620110D0 (en) | 2017-01-11 |
| GB2556944A true GB2556944A (en) | 2018-06-13 |
Family
ID=58073394
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| GB1620110.5A Withdrawn GB2556944A (en) | 2016-11-28 | 2016-11-28 | Use of decay heat from spent nuclear fuel processed by electro-reduction |
Country Status (1)
| Country | Link |
|---|---|
| GB (1) | GB2556944A (en) |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| GB2563792B (en) * | 2016-03-16 | 2022-03-09 | Richard Scott Ian | Conversion of spent uranium oxide fuel into molten salt reactor fuel |
Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3052611A (en) * | 1961-04-25 | 1962-09-04 | Roger D Piper | Method of producing uranium metal by electrolysis |
| US3298935A (en) * | 1965-04-13 | 1967-01-17 | Thomas A Henrie | Preparation of reactive metal solutions by electrodeposition methods |
-
2016
- 2016-11-28 GB GB1620110.5A patent/GB2556944A/en not_active Withdrawn
Patent Citations (2)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3052611A (en) * | 1961-04-25 | 1962-09-04 | Roger D Piper | Method of producing uranium metal by electrolysis |
| US3298935A (en) * | 1965-04-13 | 1967-01-17 | Thomas A Henrie | Preparation of reactive metal solutions by electrodeposition methods |
Non-Patent Citations (1)
| Title |
|---|
| Progress in Natural Science: Material International Vol. 25, December 2015, E-Y Choi et al, "Electrochemical processing of spent nuclear fuels: An overview of oxide reduction in pyroprocessing technology", pages 572-582 * |
Cited By (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| GB2563792B (en) * | 2016-03-16 | 2022-03-09 | Richard Scott Ian | Conversion of spent uranium oxide fuel into molten salt reactor fuel |
Also Published As
| Publication number | Publication date |
|---|---|
| GB201620110D0 (en) | 2017-01-11 |
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| WAP | Application withdrawn, taken to be withdrawn or refused ** after publication under section 16(1) |