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Cycle Separations
Terry Todd
CRESP Short Course -Introduction to Nuclear Fuel C cle
Chemistry
Crystal City, VA
,
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Why separate components of spent fuel?
Recover useful constituents of fuel for reuse
Weapons (Pu)Energy
Recycle
Waste management on on ue or op m ze sposa
Recover long-lived radioactive elements fortransmutation
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Spent (used) Nuclear Fuel
Cs and Sr 0.3%
Plutonium 0.9 %
Long-lived I and Tc 0.1%Other Long-Lived Fission
Products 0.1 %
Uranium 95.6%
Other
Minor Actinides 0.1%
Stable Fission Products 2.9%Without cladding
os ea pro uc on s rom s an r, w c ecay n ~ yr Most radiotoxicity is in long-lived fission products and the minor actinides, which can be transmuted and/or
disposed in much smaller packages
Only about 5% of the energy value of the fuel is used in a once-through fuel cycle!
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Aqueous processing - History
Began during Manhattan Project to recover Pu-239
Seaborg first separated microgram quantities of Pu in 1942 usingbismuth-phosphate precipitation process
Process scaled to kilogram quantity production at Hanford in 1944
A scale-up factor of 109!!!
separation and recovery of both U and Pu and
Reprocessing transitioned from defense to commercial use
Waste managementHanford T-Plant 1944
20 micrograms of plutonium hydroxide
1942
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Bismuth Phos hate Process
Advantages of Bismuth Phosphate Process
Recovery of >95% of Pu
Decontamination factors from fission productsof 107
Disadvantages of Bismuth Phosphate Process
Batch operations
Required numerous cycles and chemicals
Produced large volumes of high-level waste
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REDOX Process First solvent extraction process used
in reprocessing
Continuous process Recovers both U and Pu with high
ield and hi h decontaminationfactors from fission products
Developed at Argonne NationalLaboratory
Tested in pilot plant at Oak Ridge Nat.Lab 1948-49
REDOX plant built in Hanford in 1951
Used at Idaho for highly enricheduranium recovery Hanford REDOX -Plant (1951)
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REDOX Process Hexone (methyl isobutyl ketone) used as the
extractant
Immiscible with water Used to purify uranium ore concentrates
x rac s o urany an p u ony n ra esselectively from fission products
Plutonium oxidized to Pu (VI) for highest recovery
U (VI) and Pu (VI) co-extracted, then Pu is reduced toPu (III) by ferrous sulfamate and scrubbed from thesolvent
Hexone is highly flammable and volatile
Large amounts of nonvolatile salt reagents added to
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BUTEX Process
Developed in late 1940s by British scientists ata ver a ora ory
Utilized dibutyl carbitol as solvent
Nitric acid was used as salting agent
REDOX process
Lower waste volumes
Industrial operation at Windscale plant in UK until1976
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PUREX Process
Tributyl phosphate used as the extractant in a hydrocarbon diluent(dodecane or kerosene)
Suggested by Warf in 1949 for the recovery of Ce (IV) from rare earthn ra es
Developed by Knolls Atomic Power Lab. and tested at Oak Ridge in 1950-1952
-(H-canyon facility begin operation in 1955 and is still operational in 2008)
Replaced REDOX process at Hanford in 1956
Modified PUREX used in Idaho be innin in 1953 first c cle
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PUREX process chemistry
Early actinides have multiple oxidation states available in.
separate U and Pu from fission products
nS
tate
va a e ox at on state
Most stable oxidation state
Oxidation state only seen
in solids
Oxidati
Actinide element
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PUREX process chemistry
The multiple oxidation states of plutonium
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PUREX process chemistry The PUREX solvent is
hydrocarbon diluent(dodecane or kerosene)
O
TBP is derived mainly fromits phosphoryl oxygen atomcoordinating to metal ions:
P=OO
O
TBP is classed as a neutralextractant i.e. it will onlyextract electroneutral
complexes into the organicphase e.g.
n 3- n 3 4
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PUREX process chemistry
As a general rule only metal ions in the +4 and +6 oxidation statesare extracted, this means that all other species present are rejected
An4+ + 4 NO3- + 2TBP An(NO3)4(TBP)2
2+ -
This leads to an effective separation of U and Pu away from nearly
n 2 3 2 3 2 2
Strong extraction (D>>0.5) Weak extraction (D
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PUREX process- Basic principles
TBP is added
TBP Complex
Organic Solvent
UO2+2
Pu4+
UO2+2
1) Mix Phases
Aqueous Solution
UO2+2
UO2+2
Pu4+
FP
FP
FP
Cs+ Sr2+
Am3+
FP
FP
FP
Cs+ Sr2+
Am3+
2) Allow to
Settle
UO22+ + 2NO3
- + 2TBP UO2(NO3)22TBP
Pu4+ + 4NO - + 2TBP Pu NO 2TBP
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Representation of the extracted UO2
2+ complex with TBP
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PUREX process Basic Principals
The amount of metal extracted is quantified by theanal tical concentration of M in the or anic hase to its
analytical concentration in the aqueous phase atequilibrium. This is more commonly known as the
nTBPNOM
[ ]n
orgaq
orgx
MTBPNOK
D -31xm == +
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Nitric acid dependency
102U(VI)
10 Np(VI)
Pu(IV)
UO22+ + 2NO3- + 2TBP UO2(NO3)22TBP
1D
Th(IV)Pu4+ + 4NO3- + 2TBP Pu(NO3)42TBP
10-1
0 1 2 3 4 5 6 710-2
3
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PUREX Process Basic Principals
U, Np and Pu TBP extraction data plotted against eachother, from this it can be seen that extractability (D) of the
UO22+>NpO22+>PuO22+. Conversely, the extractability (D) ofthe tetravalent actinides is seen to increase across the seriesU4+
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PUREX Process Basic Principals
To remove Pu from the organic phase Pu4+ is reduced toinextractable Pu3+ using a reducing agent, usually Fe2+ or U4+
This preferred process leaves UO22+ unaffected and because
U
4+
+ 2Pu
4+
+ 2H2OUO
2
++ 2Pu
++ 4H
+
U4+ is also extractable, then Pu can be selectively backextracted.
owever, u can e uns a e n 3 so u ons ecause othe presence of nitrous acid, and thus hydrazine is added asa nitrous acid scavenger
2HNO2 + N2H4 N2 + N2O + 3H2O
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Example of Process Flowsheet Design
100
10
M(2)
0.1
1D= [M]o/[M]a
0.01
Dilute Concentrated
Acid Concentration
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PUREX Process unit operations
Se arations
Countercurrent PUREX flowsheet (1st cycle or HA cycle)
Solvent Solvent
Loaded
solvent
Coextraction
U and Pu
FP
Scrubbing
U
Scrubbing
Pu
Stripping
U
Stripping
Raffinates Feed Pu Reducin U Diluted
(FP) (U, Pu, FP....) Solution Solution solutionNitric
Acid
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THORP Reprocessing Flowsheet
Primary Separation
Conversion
to UO3
UraniumAcidScrub
UIV +
Hydrazine
Dilute
Acid Strip
Dissolved feed
from Head End
Valency
HA Cycle
Uranium
HA/HS 1BX/1BS 1C
TBP/OK solvent
TBP/OK solvent
for recycle
HAN
ScrubDilute Acid
Strip
HAN
Scrub
Condition
Powder
Accountancy
UP Cycle
r ox ePlutonium
Fission
Products &Transuranics
UP1 UP2 UP3
TBP/OK
Pu,Tc, Ru,
Cs, Ce
U, Np,
Ru
Valency
ConditionSolution
Accountancy
Conversion
to PuO2
Valency
Condition
PP1 PP2
Acid Scrub HAN Strip
TBP/OK solvent
for recycle
Solvent
PP Cycle
Np, Pu,
Ru
Tc, Ru,
Cs,Ce
Plutonium
dioxide
TBP/OK solventfor recycle
Powder
AccountancyTBP/OK
solvent
queous
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Commercial spent fuel reprocessing in the US
West Valley, NY
First lant in US to re rocess commercial SNF
Operated from 1966 until 1972
Capacity of 250-300 MTHM/yr
Shutdown due to high retrofit costs associated with changing safety and
Morris, IL
Construction halted in 1972, never operated
Close-coupled unit operations with fluoride volatility polishing step (dry U feed) Barnwell SC
1500 MTHM capacity
Construction nearly completed- startup testing was in progress
1977 change in US policy on reprocessing stopped construction
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Commercial-scale application of the PUREX
France ~
Magnox plant in La Hague began operation in 1967 (~400 MT/yr)
LWR oxide plant (UP2) began in La Hague in 1976 (800 MT/yr)
LWR oxide plant (UP3) began in La Hague in 1990 (800 MT/yr)
Windscale plant for Magnox fuel began in 1964 (1200-1500 MT/yr)
THORP LWR oxide plant began in 1994 (1000-1200 MT/yr)
Japan Tokai-Mura plant began in 1975 (~200 MT/yr)
Rokkasho plant currently undergoing hot commissioning (800 MT/yr)
Russia Plant RT-1
Began operation in 1976, 400 MT capacity Variety of headend processes for LWR, naval fuel, fast reactor fuel
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PUREX Process Current CommercialOperating Facilities
THORP, UK
La Hague, France
Rokkasho, Japan
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PUREX Process advantages and disadvantages
Advantages
Continuous operation/ High throughput
g pur ty an se ect v ty poss e can e tune y
flowsheet Recycle solvent, minimizing waste
Disadvantages
Solvent degradation due to hydrolysis and radiolysis
,
Historical handling of high-level waste
Stockpiles of plutonium oxide
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Electrochemical Processing Background
Present generation of technology forrecycling or treating spent fuel startedin the 1980s
Electrochemical processes weredeveloped for the fast reactor fuelcycle
The fast reactor fuel does not require a
high degree of decontamination Potential compactness (co-location
with reactor)
Resistance to radiation effects (short-
Criticality control benefits
Compatibility with advanced (metal)
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Example of an electrochemical flowsheet (LWR or FR
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Electrochemical Treatment of Spent Fuel
Electrochemical processing has
engineering-scale withirradiated fuel since 1996
A roximatel 3.5 MTHM of fuel, including highly enricheduranium fuels, have been
treated Installed process equipment
could support throughputsbetween 3 and 5 MTHM per year
technology are a major focus ofGNEP
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Spent Fuel Constituents are Partitioned According to FreeEnergy of Formation of Chlorides at 500C
-10
-20
HgMoW
-30
H
Zn, Cr
FeCd
lorine
) Retainedwithcladdingmaterials
in
the
anode
basketmetalwaste
-40
-
V
Mn, BeZr
o
leofch
Recoveredbyasolidcathode
as
uranium
-60Np
Cm, Am
Y
U
Pukcalper
product fabricationofrecycledfuel
Recoveredbyliquidcadmiumcathodetogetherwithuranium
External
electrical
potentialis
needed
-70 CePr, LaG
Accumulatedinthemoltensalt ceramic
fabricationofrecycledfuel
-
-90
,Sr
K
Ba, Cs, Rb
Li
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Schematic of Electrorefining Process to Treat Spent Fuel
DirectTransport(DT) Vcell=Va+Vc
Ref.Elec.
Va Vc
AnodicDissolution(AD)UU3+ U3+U
LiClKClUCl3PuPu3+
Cs Cs+
CeCe3+
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What about proliferation?
Proliferation of fissile material (i.e. Pu) has been raised as a concern forseveral decades
UREX and pyrochemical technologies were proposed as proliferationresistant technolo ies because Pu could be ke t with other TRU orradioactive fuel components
Critics do not accept this argument
Pyroprocessing now called reprocessing rather than conditioningy -
This has export control ramifications
NA-24 is now basing proliferation resistance on Attractiveness Level s opens e oor o eave w u o u e o a ower
attractiveness level
This is a change from previous policy, that isotopic dilution wasnecessary (i.e. U-233 or 235)
No technology by itself is intrinsically proliferation proof Technology is one aspect of a multifaceted approach that is necessary to
protect fissile material (with safeguards, security, transparency, etc)
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Where are we toda ? Solvent extraction is a mature technology used at
commercial scale to reprocess spent nuclear fuel
Many new extractant molecules have beendeveloped, but not demonstrated at large scale
,achievable
Electrochemical methods have been demonstratedor recovery a eng neer ng-sca e
TRU recovery and salt recycle have not beendemonstrated at en ineerin -scale
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Where are oin ? Research into advanced separation methods as part
of the Advanced Fuel Cycle program in progress
New Aqueous methods
Electrochemical methods
Transformational methods
Integration of separation R&D efforts with waste
No more throw it over the fence approach
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