EP1091908A1 - Nuclear fuel reprocessing - Google Patents
Nuclear fuel reprocessingInfo
- Publication number
- EP1091908A1 EP1091908A1 EP19990925171 EP99925171A EP1091908A1 EP 1091908 A1 EP1091908 A1 EP 1091908A1 EP 19990925171 EP19990925171 EP 19990925171 EP 99925171 A EP99925171 A EP 99925171A EP 1091908 A1 EP1091908 A1 EP 1091908A1
- Authority
- EP
- European Patent Office
- Prior art keywords
- phase
- aqueous
- organic phase
- technetium
- zirconium
- 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.)
- Withdrawn
Links
- 238000012958 reprocessing Methods 0.000 title claims abstract description 23
- 239000003758 nuclear fuel Substances 0.000 title claims description 5
- 238000000034 method Methods 0.000 claims abstract description 58
- 239000012074 organic phase Substances 0.000 claims abstract description 55
- 239000008346 aqueous phase Substances 0.000 claims abstract description 47
- 229910052713 technetium Inorganic materials 0.000 claims abstract description 40
- 229910052726 zirconium Inorganic materials 0.000 claims abstract description 38
- GKLVYJBZJHMRIY-UHFFFAOYSA-N technetium atom Chemical compound [Tc] GKLVYJBZJHMRIY-UHFFFAOYSA-N 0.000 claims abstract description 37
- QCWXUUIWCKQGHC-UHFFFAOYSA-N Zirconium Chemical compound [Zr] QCWXUUIWCKQGHC-UHFFFAOYSA-N 0.000 claims abstract description 35
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 claims abstract description 26
- 229910017604 nitric acid Inorganic materials 0.000 claims abstract description 26
- 230000004992 fission Effects 0.000 claims abstract description 25
- 239000012071 phase Substances 0.000 claims abstract description 25
- 229910052770 Uranium Inorganic materials 0.000 claims abstract description 24
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 claims abstract description 23
- 229910052778 Plutonium Inorganic materials 0.000 claims abstract description 22
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 claims abstract description 22
- 239000002915 spent fuel radioactive waste Substances 0.000 claims abstract description 18
- 229910052751 metal Inorganic materials 0.000 claims abstract description 13
- 239000002184 metal Substances 0.000 claims abstract description 13
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 claims abstract description 12
- 229910052768 actinide Inorganic materials 0.000 claims abstract description 6
- 150000001255 actinides Chemical class 0.000 claims abstract description 5
- 229910052781 Neptunium Inorganic materials 0.000 claims description 15
- LFNLGNPSGWYGGD-UHFFFAOYSA-N neptunium atom Chemical compound [Np] LFNLGNPSGWYGGD-UHFFFAOYSA-N 0.000 claims description 15
- 239000011260 aqueous acid Substances 0.000 claims description 12
- 239000002904 solvent Substances 0.000 claims description 10
- 239000002253 acid Substances 0.000 claims description 9
- 238000000605 extraction Methods 0.000 claims description 6
- 239000000446 fuel Substances 0.000 claims description 6
- 239000000463 material Substances 0.000 claims description 4
- 239000004594 Masterbatch (MB) Substances 0.000 claims description 2
- 238000011001 backwashing Methods 0.000 claims description 2
- 238000005192 partition Methods 0.000 claims description 2
- 239000008188 pellet Substances 0.000 claims description 2
- 239000000843 powder Substances 0.000 claims description 2
- -1 actinide metals Chemical class 0.000 claims 1
- WJWSFWHDKPKKES-UHFFFAOYSA-N plutonium uranium Chemical group [U].[Pu] WJWSFWHDKPKKES-UHFFFAOYSA-N 0.000 claims 1
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 description 6
- 238000000926 separation method Methods 0.000 description 5
- 230000015572 biosynthetic process Effects 0.000 description 3
- 230000000694 effects Effects 0.000 description 2
- 230000007613 environmental effect Effects 0.000 description 2
- 238000000638 solvent extraction Methods 0.000 description 2
- 239000004215 Carbon black (E152) Substances 0.000 description 1
- 239000000284 extract Substances 0.000 description 1
- 229930195733 hydrocarbon Natural products 0.000 description 1
- 150000002430 hydrocarbons Chemical class 0.000 description 1
- 239000003350 kerosene Substances 0.000 description 1
- 229910052747 lanthanoid Inorganic materials 0.000 description 1
- 150000002602 lanthanoids Chemical class 0.000 description 1
- WRIRWRKPLXCTFD-UHFFFAOYSA-N malonamide Chemical compound NC(=O)CC(N)=O WRIRWRKPLXCTFD-UHFFFAOYSA-N 0.000 description 1
- 239000000203 mixture Substances 0.000 description 1
- 238000001556 precipitation Methods 0.000 description 1
- 238000011084 recovery Methods 0.000 description 1
- AAORDHMTTHGXCV-UHFFFAOYSA-N uranium(6+) Chemical compound [U+6] AAORDHMTTHGXCV-UHFFFAOYSA-N 0.000 description 1
Classifications
-
- C—CHEMISTRY; METALLURGY
- C01—INORGANIC CHEMISTRY
- C01G—COMPOUNDS CONTAINING METALS NOT COVERED BY SUBCLASSES C01D OR C01F
- C01G56/00—Compounds of transuranic elements
- C01G56/001—Preparation involving a liquid-liquid extraction, an adsorption or an ion-exchange
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/0204—Obtaining thorium, uranium, or other actinides obtaining uranium
- C22B60/0217—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes
- C22B60/0221—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching
- C22B60/0226—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using acidic solutions or liquors
- C22B60/0239—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes by leaching using acidic solutions or liquors nitric acid containing ion as active agent
-
- C—CHEMISTRY; METALLURGY
- C22—METALLURGY; FERROUS OR NON-FERROUS ALLOYS; TREATMENT OF ALLOYS OR NON-FERROUS METALS
- C22B—PRODUCTION AND REFINING OF METALS; PRETREATMENT OF RAW MATERIALS
- C22B60/00—Obtaining metals of atomic number 87 or higher, i.e. radioactive metals
- C22B60/02—Obtaining thorium, uranium, or other actinides
- C22B60/0204—Obtaining thorium, uranium, or other actinides obtaining uranium
- C22B60/0217—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes
- C22B60/0252—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes treatment or purification of solutions or of liquors or of slurries
- C22B60/026—Obtaining thorium, uranium, or other actinides obtaining uranium by wet processes treatment or purification of solutions or of liquors or of slurries liquid-liquid extraction with or without dissolution in organic solvents
-
- 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
- G21C19/46—Aqueous processes, e.g. by using organic extraction means, including the regeneration of these means
-
- 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
Definitions
- This invention relates to nuclear fuel reprocessing and is particularly concerned with the separation of uranium, plutonium and neptunium from zirconium and technetium.
- the organic phase is subjected to separation of fission products by solvent extraction before the so-called U/Pu split.
- This process is typically completed in two stages: separation of the fission products, except technetium (e.g. in the so-called
- a concentrated acid feed (ca. 4 - 6 M nitric acid) with high flowrate, relative to the dissolved spent fuel flowrate, is required in the Tc rejection contactors to efficiently backwash technetium.
- the essence of the present invention is that a process or plant for reprocessing or treating spent fuel uses a technique in which both zirconium and technetium as well as other fission products are simultaneously separated from uranium and plutonium.
- the present invention provides a method for treating or reprocessing spent nuclear fuel in which an organic phase is contacted with an aqueous nitric acid phase containing zirconium, technetium and at least one extractable metal selected from uranium, plutonium and neptunium, the aqueous phase having a relatively low acidity and a relatively high flowrate such that the at least one extractable metal is extracted into the organic phase while the zirconium and technetium predominantly remain in the aqueous phase.
- the aqueous phase normally contains other fission products in addition to zirconium and technetium, and these also predominantly remain in the aqueous phase.
- the aqueous phase usually contains all of the uranium, plutonium and neptunium.
- the method normally further comprises contacting the organic phase into which at least one extractable metal has been extracted with a second aqueous nitric acid phase to strip into the second aqueous phase zirconium, technetium and other fission products which have entered the organic phase.
- the second aqueous phase usually becomes incorporated into the first aqueous phase which is contacted with the organic phase.
- the second aqueous phase feed is preferably of intermediate acidity, e.g. has a nitric acid concentration from 2 to 4 M.
- the invention also provides a Purex reprocessing plant in which there are arranged in series along a solvent stream flowpath (i) a unit for extraction of uranium, plutonium and neptunium from an aqueous acid stream containing dissolved spent fuel into the solvent stream while zirconium, technetium and other fission products predominantly remain in the aqueous phase, (ii) a unit for stripping of zirconium, technetium and other fission products from the solvent stream into an aqueous acid phase which passes to the unit (i) where a combination thereof with a dissolved spent fuel feed forms said aqueous acid stream, and, optionally, (iii) a unit for backstripping of acidity from the solvent stream into a low acid strip stream which passes to the unit (ii) where a combination thereof with an intermediate acidity strip stream forms said aqueous acid phase.
- Figure 1 is a standard single cycle Purex flowsheet
- Figure 2 is a partial flowsheet of a Purex reprocessing process incorporating the method of the invention.
- the flowsheet contains the units shown in Tables 1 & 2.
- Figure 2 illustrates a portion of a reprocessing plant which contains apparatus (unit HA) in which is performed a method for treating or reprocessing spent nuclear fuel in which an organic phase, in this case 30 % TBP/OK, is contacted with an aqueous nitric acid phase containing zirconium, technetium and at least one extractable metal selected from uranium, plutonium and neptunium, the aqueous phase having a relatively low acidity and a relatively high flowrate such that the at least one extractable metal is extracted into the organic phase while the zirconium and technetium predominantly remain in the aqueous phase.
- an organic phase in this case 30 % TBP/OK
- the organic stream SI is contacted with an aqueous nitric acid phase comprising the combined aqueous streams Al, A2 and A3 in the unit HA, which is a multi-stage contactor in the illustrated embodiment.
- the organic stream extracts uranium, plutonium and neptunium into the organic phase.
- the fission products including zirconium and technetium remain in the aqueous phase and exit the contactor in PI, the highly active raffmate.
- the nitric acid concentration of the aqueous phase product is preferably within the range of 2 to 4M, preferably 2 to 3M and preferably best at 2.2 M.
- the ratio of the organic phase flowrate to aqueous phase (Al + A2 + A3) flowrate to is preferably 0.7 and 1.5, preferably between 0.8, eg 8.8 to 1.3 and preferably set at 0.91; the preferred numerical parameters are generally applicable to all embodiments of the invention.
- the organic phase loaded with uranium, plutonium and neptunium goes from unit HA to unit HS, in this case a multi-stage contactor unit, where fission products including zirconium and technetium are stripped from the organic phase into a second aqueous nitric acid phase (Al plus A2).
- the organic phase loaded with uranium, plutonium and neptunium goes from unit HS to unit HSS, in this case a multi-stage contactor unit, where acidity is backwashed from the organic phase.
- This is not an essential requirement of the invention because it is performed to reduce the organic phase acidity prior to the U/Pu split. It has little effect on zirconium and technetium stripping from the organic phase.
- the aqueous stream Al is contacted with the organic stream in the unit HSS.
- the aqueous stream is a low acidity (e.g. 0.05 to 0.2 M nitric acid), low flowrate stream so that acid can be stripped from the organic phase.
- the solvent : aqueous flowrate ratio may be from 0.05 to 0.20, depending on the contactor equipment performance.
- the aqueous product of the unit HSS is combined with the aqueous stream A2 and is contacted with the organic stream in the unit HS.
- the aqueous stream A2 is of intermediate acidity (ca. 4 to 2 M nitric acid) although this is not an essential requirement of the invention. It is also of reasonably high flowrate to ensure efficient stripping of fission products including zirconium and technetium.
- the second aqueous phase (combined Al and A2) therefore usually has intermediate acidity, usually from 2 to 4 M, and the ratio of the second organic phase flowrate to aqueous phase flowrate to is usually between 1.0 to 2.0, preferably 1.1 to 2.0 but desirably about 1.3.
- the aqueous product of the unit HS is combined with the aqueous stream A3 and is contacted with the organic stream in the unit HA.
- the aqueous stream A3 is the dissolved spent fuel feed.
- the A3 stream is again of intermediate acidity but normally less than 3 M nitric acid. Optimum performance of this invention is achieved when the acidity of this stream is minimised.
- the invention is not restricted as to the manner in which the organic product II is treated.
- II is sent to a uranium, plutonium split operation, for example a conventional process as used in a commercial plant.
- Table 1 Illustrative flowrates and concentrations for some of the streams shown in Figure 2.
- the method of the invention dispenses with the separation of technetium in technetium rejection contactors. Accordingly, the plant may be smaller, resulting in both environmental and economic benefits. The method also features lower nitric acid inventory so again resulting in both environmental and economic benefits.
- a further benefit of the preferred methods of the invention is that relatively relaxed flowsheet control is possible in the HA/ ⁇ S/HSS contactors.
- the preferred method uses the same solvent inventory but avoids a high uranium loading in the HS contactor cascade. Accordingly, variation in flowrates and feed compositions has significantly less effect on flowsheet performance.
- a yet further benefit enabled by the method of the invention is that excessive zirconium recycle in the HA/ ⁇ S HSS contactors is completely avoided because zirconium is effectively routed to PI, the highly active raffmate. Accordingly, the risk of third phase formation is significantly reduced.
- the invention includes a Purex reprocessing method in which an aqueous acid phase containing dissolved spent fuel is contacted with an organic phase, characterised in that the aqueous phase contacted with the organic phase has a relatively low acidity and a relatively high flowrate such that uranium, plutonium and neptunium are extracted from the aqueous phase into the organic phase while zirconium, technetium and other fission products predominantly remain in the aqueous phase.
- the organic phase and the aqueous phase are contacted in a first contactor unit, and the organic phase is fed from the first contactor unit to a second contactor unit in which it contacts a second aqueous nitric acid phase to strip into the second aqueous phase zirconium, technetium and other fission products which have entered the organic phase, the second aqueous phase being combined with the aqueous acid phase containing the dissolved spent fuel before the latter contacts the organic phase in the first contactor.
- the organic phase is fed from the second contactor unit into a third contactor unit in which the organic phase is contacted with an aqueous low acid strip phase for backwashing of acidity from the organic phase into the low acid strip phase.
- the methods of the invention are typically methods for reprocessing nuclear fuel to form a fissile material optionally in the form of a gel, a powder, a master batch material, a fuel pellet, a fuel pin or a fuel assembly.
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- Chemical & Material Sciences (AREA)
- Engineering & Computer Science (AREA)
- General Life Sciences & Earth Sciences (AREA)
- Organic Chemistry (AREA)
- Life Sciences & Earth Sciences (AREA)
- Mechanical Engineering (AREA)
- Manufacturing & Machinery (AREA)
- Materials Engineering (AREA)
- Geology (AREA)
- Metallurgy (AREA)
- Environmental & Geological Engineering (AREA)
- Physics & Mathematics (AREA)
- Plasma & Fusion (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Inorganic Chemistry (AREA)
- Inorganic Compounds Of Heavy Metals (AREA)
Abstract
There is described a method for treating or reprocessing spent nuclear fuel in which an organic phase is contacted with an aqueous nitric acid phase containing zirconium, technetium and at least one extractable actinide metal such that the at least one extractable metal is extracted into the organic phase characterised in that the zirconium and technetium predominantly remain in the aqueous phase. There is also described a Purex reprocessing plant, and the use of such a plant in simultaneously separating both zirconium and technetium along with other fission products from uranium and plutonium.
Description
NUCLEAR FUEL REPROCESSING
This invention relates to nuclear fuel reprocessing and is particularly concerned with the separation of uranium, plutonium and neptunium from zirconium and technetium.
Most commercial reprocessing plants use the Purex process, in which the spent fuel is dissolved in nitric acid and the dissolved uranium, plutonium and neptunium are subsequently extracted from the nitric acid solution into an organic phase of tributyl phosphate (TBP) dissolved in an inert hydrocarbon such as odourless kerosene. The organic phase is then subjected to various solvent extraction techniques to remove fission products and also partition the uranium from the plutonium and neptunium.
More particularly, the organic phase is subjected to separation of fission products by solvent extraction before the so-called U/Pu split. This process is typically completed in two stages: separation of the fission products, except technetium (e.g. in the so-called
HA/HS contactors), and subsequent separation of technetium (in Tc rejection contactors).
There are disadvantages with such a process:
• Excellent flowsheet control is required in the HA/HS contactor cascade to maintain high uranium loading in the organic phase. Otherwise loss of uranium, plutonium and neptunium can occur to the highly active raffinate.
• Excessive zirconium recycle can occur in the HA/HS contactors thereby resulting in third phase formation.
• A concentrated acid feed (ca. 4 - 6 M nitric acid) with high flowrate, relative to the dissolved spent fuel flowrate, is required in the Tc rejection contactors to efficiently backwash technetium.
US Patent No. 5,132,092, Musikas et al describes a process for recovery of uranium (VI) and/or plutonium (IV). However, the prior art does not use tri-n-butyl phosphate (TBP), rather it uses a novel range of N,N-alkylamides. In addition, Musikas et al, describes that both Tc and Zr can be effectively backwashed together, however, this is far from certain because, inter alia, Musikas provides extraction values for Zr and Tc as a function of nitric acid only. Musikas draws the conclusion that because extraction values for Zr/Tc
into nitric acid are low, then they could be readily backwashed. However, we have found that it is the formation of complexes between Tc and U(VI) or ZR(VI) which cause Tc/Zr to be extractable. Therefore, Musikas does not solve the problem of the separate extraction requirement for Tc.
Furthermore, US Patent No. 5,223,232, Cuillerdier, et al, describes a method of separating Fe and Zr from the actinides or lanthanides. The process described by Duillerdier uses a propane diamide rather than TBP. Moreover, Cuillerdier again does not address the problem of the separate extraction step for the removal of Tc.
The essence of the present invention is that a process or plant for reprocessing or treating spent fuel uses a technique in which both zirconium and technetium as well as other fission products are simultaneously separated from uranium and plutonium.
Thus, according to the invention, we provide a method for treating or reprocessing spent nuclear fuel in which an organic phase is contacted with an aqueous nitric acid phase containing zirconium, technetium and at least one extractable actinide metal such that the at least one extractable metal is extracted into the organic phase characterised in that the zirconium and technetium predominantly remain in the aqueous phase.
More particularly, the present invention provides a method for treating or reprocessing spent nuclear fuel in which an organic phase is contacted with an aqueous nitric acid phase containing zirconium, technetium and at least one extractable metal selected from uranium, plutonium and neptunium, the aqueous phase having a relatively low acidity and a relatively high flowrate such that the at least one extractable metal is extracted into the organic phase while the zirconium and technetium predominantly remain in the aqueous phase.
The aqueous phase normally contains other fission products in addition to zirconium and technetium, and these also predominantly remain in the aqueous phase. The aqueous phase usually contains all of the uranium, plutonium and neptunium.
The method normally further comprises contacting the organic phase into which at least
one extractable metal has been extracted with a second aqueous nitric acid phase to strip into the second aqueous phase zirconium, technetium and other fission products which have entered the organic phase. The second aqueous phase usually becomes incorporated into the first aqueous phase which is contacted with the organic phase. The second aqueous phase feed is preferably of intermediate acidity, e.g. has a nitric acid concentration from 2 to 4 M.
The invention also provides a Purex reprocessing plant in which there are arranged in series along a solvent stream flowpath (i) a unit for extraction of uranium, plutonium and neptunium from an aqueous acid stream containing dissolved spent fuel into the solvent stream while zirconium, technetium and other fission products predominantly remain in the aqueous phase, (ii) a unit for stripping of zirconium, technetium and other fission products from the solvent stream into an aqueous acid phase which passes to the unit (i) where a combination thereof with a dissolved spent fuel feed forms said aqueous acid stream, and, optionally, (iii) a unit for backstripping of acidity from the solvent stream into a low acid strip stream which passes to the unit (ii) where a combination thereof with an intermediate acidity strip stream forms said aqueous acid phase.
The present invention is further described by way of example only with reference to the accompanying drawings, in which:
Figure 1 is a standard single cycle Purex flowsheet; and
Figure 2 is a partial flowsheet of a Purex reprocessing process incorporating the method of the invention.
The following symbols are used in the Figures:
Ai = Aqueous feeds Si = Organic feeds Ii = Intermediate organic streams Pi = Product streams Double arrows = Organic streams Single arrows = Aqueous streams
The flowsheet contains the units shown in Tables 1 & 2.
Table 1. Units used in the Purex reprocessing plant in Figure 1.
The aqueous feeds, intermediate organic streams and product streams shown in Figure 2 are as follows:
Al Low acid strip. A2 Intermediate acidity strip. A3 Dissolved spent fuel feed. SI : Organic feed. PI : Highly active raffinate. II : Organic product from fission product removal.
Figure 2 illustrates a portion of a reprocessing plant which contains apparatus (unit HA) in which is performed a method for treating or reprocessing spent nuclear fuel in which an organic phase, in this case 30 % TBP/OK, is contacted with an aqueous nitric acid phase containing zirconium, technetium and at least one extractable metal selected from uranium, plutonium and neptunium, the aqueous phase having a relatively low acidity
and a relatively high flowrate such that the at least one extractable metal is extracted into the organic phase while the zirconium and technetium predominantly remain in the aqueous phase.
More particularly, the organic stream SI is contacted with an aqueous nitric acid phase comprising the combined aqueous streams Al, A2 and A3 in the unit HA, which is a multi-stage contactor in the illustrated embodiment. The organic stream extracts uranium, plutonium and neptunium into the organic phase. The fission products including zirconium and technetium remain in the aqueous phase and exit the contactor in PI, the highly active raffmate. The nitric acid concentration of the aqueous phase product is preferably within the range of 2 to 4M, preferably 2 to 3M and preferably best at 2.2 M. The ratio of the organic phase flowrate to aqueous phase (Al + A2 + A3) flowrate to is preferably 0.7 and 1.5, preferably between 0.8, eg 8.8 to 1.3 and preferably set at 0.91; the preferred numerical parameters are generally applicable to all embodiments of the invention.
The organic phase loaded with uranium, plutonium and neptunium goes from unit HA to unit HS, in this case a multi-stage contactor unit, where fission products including zirconium and technetium are stripped from the organic phase into a second aqueous nitric acid phase (Al plus A2).
The organic phase loaded with uranium, plutonium and neptunium goes from unit HS to unit HSS, in this case a multi-stage contactor unit, where acidity is backwashed from the organic phase. This is not an essential requirement of the invention because it is performed to reduce the organic phase acidity prior to the U/Pu split. It has little effect on zirconium and technetium stripping from the organic phase. Thus it is preferred but not essential that the aqueous stream Al is contacted with the organic stream in the unit HSS. The aqueous stream is a low acidity (e.g. 0.05 to 0.2 M nitric acid), low flowrate stream so that acid can be stripped from the organic phase. The solvent : aqueous flowrate ratio may be from 0.05 to 0.20, depending on the contactor equipment performance.
The aqueous product of the unit HSS is combined with the aqueous stream A2 and is
contacted with the organic stream in the unit HS. The aqueous stream A2 is of intermediate acidity (ca. 4 to 2 M nitric acid) although this is not an essential requirement of the invention. It is also of reasonably high flowrate to ensure efficient stripping of fission products including zirconium and technetium. The second aqueous phase (combined Al and A2) therefore usually has intermediate acidity, usually from 2 to 4 M, and the ratio of the second organic phase flowrate to aqueous phase flowrate to is usually between 1.0 to 2.0, preferably 1.1 to 2.0 but desirably about 1.3.
The aqueous product of the unit HS is combined with the aqueous stream A3 and is contacted with the organic stream in the unit HA. The aqueous stream A3 is the dissolved spent fuel feed. The A3 stream is again of intermediate acidity but normally less than 3 M nitric acid. Optimum performance of this invention is achieved when the acidity of this stream is minimised.
The invention is not restricted as to the manner in which the organic product II is treated. In the illustrated embodiment, II is sent to a uranium, plutonium split operation, for example a conventional process as used in a commercial plant. In one alternative, the uranium and plutonium may be co-processed into a MOX [MOX = mixed oxide] product, for example via a gel precipitation route as described in the UK patent application No 9722497.6.
Illustrative flowrates and concentrations for some of the streams shown in Figure 2 appear in table 1 :
Table 1 : Illustrative flowrates and concentrations for some of the streams shown in Figure 2.
The method of the invention dispenses with the separation of technetium in technetium rejection contactors. Accordingly, the plant may be smaller, resulting in both environmental and economic benefits. The method also features lower nitric acid inventory so again resulting in both environmental and economic benefits.
A further benefit of the preferred methods of the invention is that relatively relaxed flowsheet control is possible in the HA/ΗS/HSS contactors. The preferred method uses the same solvent inventory but avoids a high uranium loading in the HS contactor cascade. Accordingly, variation in flowrates and feed compositions has significantly less effect on flowsheet performance.
A yet further benefit enabled by the method of the invention is that excessive zirconium recycle in the HA/ΗS HSS contactors is completely avoided because zirconium is effectively routed to PI, the highly active raffmate. Accordingly, the risk of third phase formation is significantly reduced.
It will be appreciated that the invention includes a Purex reprocessing method in which an aqueous acid phase containing dissolved spent fuel is contacted with an organic phase, characterised in that the aqueous phase contacted with the organic phase has a relatively low acidity and a relatively high flowrate such that uranium, plutonium and neptunium are extracted from the aqueous phase into the organic phase while zirconium, technetium and other fission products predominantly remain in the aqueous phase. Normally the organic phase and the aqueous phase are contacted in a first contactor unit, and the organic phase is fed from the first contactor unit to a second contactor unit in which it contacts a second aqueous nitric acid phase to strip into the second aqueous phase zirconium, technetium and other fission products which have entered the organic phase, the second aqueous phase being combined with the aqueous acid phase containing the dissolved spent fuel before the latter contacts the organic phase in the first contactor. Optionally, the organic phase is fed from the second contactor unit into a third contactor unit in which the organic phase is contacted with an aqueous low acid strip phase for backwashing of acidity from the organic phase into the low acid strip phase.
The methods of the invention are typically methods for reprocessing nuclear fuel to form a fissile material optionally in the form of a gel, a powder, a master batch material, a fuel pellet, a fuel pin or a fuel assembly.
Claims
1. A method for treating or reprocessing spent nuclear fuel in which an organic phase is contacted with an aqueous nitric acid phase containing zirconium, technetium and at least one extractable actinide metal such that the at least one extractable metal is extracted into the organic phase characterised in that the zirconium and technetium predominantly remain in the aqueous phase.
2. A method according to claim 1 characterised in that the actinide is selected from uranium plutonium and n and neptunium.
3. A method according to claim 1 characterised in that the aqueous phase has a relatively low acidity.
4. A method according to claim 1 characterised in that the aqueous phase has a relatively high flow rate.
5. A method of claim 1 in which the aqueous phase further contains other fission products in addition to zirconium and technetium, which fission products predominantly remain in the aqueous phase.
6. A method of claim 2 characterised in that the aqueous phase contains all of the uranium, plutonium and neptunium.
7. A method of any of claims 1 to 3 characterised in that the concentration of nitric acid in the aqueous phase product is from 1.0 to 4.0M..
8. A method of claim 7 characterised in that the nitric acid concentration is 2.2 M.
9. A method of any of claim 1 characterised in that the ratio of the organic phase flowrate : aqueous phase flowrate is from 0.9 to 1.3.
10. A method of claim 9 characterised in that the ratio is 0.91.
11. A method of claim 9 characterised in that the ratio is no more than 1.3.
12. A method of claim 1 characterised in that it further comprises contacting the organic phase, into which the at least one extractable metal has been extracted, with a second aqueous nitric acid phase to strip any zirconium, technetium and other fission products which have entered the organic phase into the second aqueous phase.
13. A method of claim 12 characterised in that the second aqueous phase becomes incorporated into the first aqueous phase which is contacted with organic phase.
14. A method of claim 12 characterised in that the second aqueous phase is of intermediate acidity.
15. A method of claim 14 characterised in that the second aqueous phase has a nitric acid concentration of from 2 to 4 M.
16. A method of claim 12 characterised in that the ratio of the second organic phase flowrate : aqueous phase flowrate is from 1.1 to 2.0.
17. A method of claim 16 characterised in that the ratio is 1.3.
18. A method of claim 16 characterised in that the ratio is 1.3.
19. A Purex reprocessing method in which an aqueous acid phase containing dissolved spent fuel is contacted with an organic phase, characterised in that the actinide metals are extracted from the aqueous phase into the organic phase while zirconium, technetium and other fission products predominantly remain in the aqueous phase.
20. A Purex reprocessing method of claim 19 characterised in that the aqueous phase contacted with the organic phase has a relatively low acidity.
21. A Purex reprocessing method of claim 19 characterised in that the aqueous phase contacts with the organic phase has a relatively high flowrate.
22. A method of claim 9 which further includes the feature(s) recited in one or more of claims 7 to 11.
23. A method of claim 19 characterised in that the organic phase and the aqueous phase are contacted in a first contactor unit, and the organic phase is fed from the first contactor unit to a second contactor unit in which it contacts a second aqueous nitric acid phase to strip into the second aqueous phase zirconium, technetium and other fission products which have entered the organic phase, the second aqueous phase being combined with the aqueous acid phase containing the dissolved spent fuel before the latter contacts the organic phase in the first contactor.
24. A method of claim 23 which further includes the feature(s) recited in one or more of claims 15 to 18.
25. A method of claim 23 characterised in that the organic phase is fed from the second contactor unit into a third contactor unit in which the organic phase is contacted with an aqueous low acid strip phase for backwashing of acidity from the organic phase into the low acid strip phase.
26. A method of any one of the preceding claims characterised in that the organic phase after stripping of the zirconium, technetium and other fission products is fed to a uranium, plutonium split operation.
27. A method of any of claims 1 to 25 characterised in that the organic phase after stripping of the zirconium, technetium and other fission products is not treated to partition uranium and plutonium.
28. A method of any one of the preceding claims which is a method for reprocessing nuclear fuel to form a fissile material optionally in the form of a gel, a powder, a master batch material, a fuel pellet, a fuel pin or a fuel assembly.
29. A Purex reprocessing plant in which there are arranged in series along a solvent stream flowpath (i) a unit for extraction of uranium, plutonium and neptunium from an aqueous acid stream containing dissolved spent fuel into the solvent stream while zirconium, technetium and other fission products predominantly remain in the aqueous phase, (ii) a unit for stripping of zirconium, technetium and other fission products from the solvent stream into an aqueous acid phase which passes to the unit (i) where a combination thereof with a dissolved spent fuel feed forms said aqueous acid stream, and, optionally, (iii) a unit for backstripping of acidity from the solvent stream into a low acid strip stream which passes to the unit (ii) where a combination thereof with an intermediate acidity strip stream forms said aqueous acid phase.
30. The use of a plant of claim 29 to perform a method of any of claims 1 to 28.
31. The use in a Purex reprocessing method or plant of a technique for simultaneously separating both zirconium and technetium along with other fission products from uranium and plutonium, in which technique an aqueous nitric acid stream of relatively low acidity and high flowrate is contacted with an organic stream.
Applications Claiming Priority (5)
Application Number | Priority Date | Filing Date | Title |
---|---|---|---|
GB9811727 | 1998-06-02 | ||
GBGB9811727.8A GB9811727D0 (en) | 1998-06-02 | 1998-06-02 | Nuclear fuel reprocessing |
GB9818380 | 1998-08-24 | ||
GBGB9818380.9A GB9818380D0 (en) | 1998-08-24 | 1998-08-24 | Nuclear fuel reprocessing |
PCT/GB1999/001711 WO1999062824A1 (en) | 1998-06-02 | 1999-05-28 | Nuclear fuel reprocessing |
Publications (1)
Publication Number | Publication Date |
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EP1091908A1 true EP1091908A1 (en) | 2001-04-18 |
Family
ID=26313776
Family Applications (1)
Application Number | Title | Priority Date | Filing Date |
---|---|---|---|
EP19990925171 Withdrawn EP1091908A1 (en) | 1998-06-02 | 1999-05-28 | Nuclear fuel reprocessing |
Country Status (3)
Country | Link |
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EP (1) | EP1091908A1 (en) |
JP (1) | JP2002516810A (en) |
WO (1) | WO1999062824A1 (en) |
Families Citing this family (4)
Publication number | Priority date | Publication date | Assignee | Title |
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FR2880180B1 (en) * | 2004-12-29 | 2007-03-02 | Cogema | IMPROVEMENT OF THE PUREX PROCESS AND ITS USES |
FR2903025B1 (en) * | 2006-07-03 | 2008-10-10 | Cogema | PROCESS FOR SEPARATING A CHEMICAL ELEMENT FROM URANIUM FROM AN ACOUSTIC AQUEOUS PHASE IN A URANIUM EXTRACTION CYCLE |
CN103426489B (en) * | 2012-05-17 | 2016-01-27 | 中国原子能科学研究院 | Method for improving technetium washing effect in post-treatment extraction separation process |
CN109735859B (en) * | 2019-03-18 | 2020-10-09 | 中国原子能科学研究院 | Application of 3-pentylhydrazine and salt thereof |
Family Cites Families (3)
Publication number | Priority date | Publication date | Assignee | Title |
---|---|---|---|---|
FR2583989B1 (en) * | 1985-06-26 | 1987-07-31 | Commissariat Energie Atomique | PROCESS FOR AVOIDING THE EXTRACTION OF TECHNETIUM AND / OR RHENIUM IN PARTICULAR DURING THE EXTRACTION OF URANIUM AND / OR PLUTONIUM, BY AN ORGANIC SOLVENT |
FR2607823B1 (en) * | 1986-12-03 | 1989-02-17 | Commissariat Energie Atomique | PROCESS FOR SEPARATING TECHNETIUM FROM AN ORGANIC SOLVENT WITH ZIRCONIUM AND AT LEAST ANOTHER METAL SUCH AS URANIUM OR PLUTONIUM, ESPECIALLY USEFUL FOR THE TREATMENT OF IRRADIATED NUCLEAR FUELS |
JPH0712987A (en) * | 1993-06-21 | 1995-01-17 | Japan Atom Energy Res Inst | Concentration, separation and recovering process for technetium in co-decontamination step in reprocessing of spent nuclear fuel |
-
1999
- 1999-05-28 EP EP19990925171 patent/EP1091908A1/en not_active Withdrawn
- 1999-05-28 WO PCT/GB1999/001711 patent/WO1999062824A1/en not_active Application Discontinuation
- 1999-05-28 JP JP2000552044A patent/JP2002516810A/en active Pending
Non-Patent Citations (1)
Title |
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See references of WO9962824A1 * |
Also Published As
Publication number | Publication date |
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JP2002516810A (en) | 2002-06-11 |
WO1999062824A1 (en) | 1999-12-09 |
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