04 todd cresp 2009 presentation separations todd

7/31/2019 04 Todd CRESP 2009 Presentation Separations Todd

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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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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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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

04 todd cresp 2009 presentation separations todd

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