DEVELOPMENT OF FRONTEND PROCESSING TO ALLOW USE OF HIGH-DENSITY LEU - PowerPoint PPT Presentation

DEVELOPMENT OF FRONTEND PROCESSING TO ALLOW USE OF HIGH-DENSITY LEU FOIL TARGETS IN CURRENT Mo-99 PRODUCTION FACILITIES M. Alex Brown, James L. Jerden Jr., Artem V. Gelis, Dominque C. Stepinski, Stan Wiedmeyer, Amanda Youker, Andrew Hebden, George F. Vandegrift Chemical Sciences & Engineering Division Argonne National Laboratory June 29, 2014 Mo-99 Topical, Washington, D.C

URANIUM ENRICHMENT > 90% U-235 U-Foil Nuclear Chemical < 20% U-235 UAl x - Target

ARGONNE HD-TARGET FRONTEND PROCESSES LEU IRRADIATE ACID BASE PROCESS PROCESS Mo-99 Mo-99 PURIFICATION 3

ARGONNE HD-TARGET FRONTEND PROCESSES ACID ELECTROCHEMICAL PROCESS PROCESS 4

ARGONNE HD-TARGET FRONTEND PROCESSES  Full-scale  Prototype that design can be scaled  250-g U/batch up  Resistant to  20-g U/batch radiation,  Resistant to corrosion, and radiation, hot-cell corrosion, and compatible hot-cell  Cold test (Ni) compatible  Warm test (DU)  Warm test (DU)  Hot test (irradiated  Hot test (irradiated LEU) LEU) 5

LEU IRRADIATIONS AT ARGONNE  LEU foils: 6 – 15 grams  Mimic fission recoil barriers: Al (electrochemical) / Ni (acid)  Thermal neutron flux: ~10 11 n × cm -2 × s -1  10 minute irradiation  Over-night cooling  Calculations: 50-100 µCi 99 Mo 6

THE ACID PROCESS  Uranium foil dissolved in nitric acid U + 4HNO 3  UO 2 (NO 3 ) 2 + 2H 2 O + 2NO  Nickel fission-recoil barrier and all other components dissolve also  Product fed to titania column for Mo recovery/separation and conversion to alkaline solution  Alkaline Mo-product solution to current purification process 7

THE ACID PROCESS - DISSOLUTION  Tested with Ni alone, DU, and finally with 242 g DU + 6 g irradiated LEU.  All components dissolve in 500 mL of nitric acid  100% of Ni and U foil were dissolved in 2 hours 8

THE ACID PROCESS – Mo RECOVERY 140 Ba, 141 Ce, 133 I, 99 Mo HNO 3 + Uranium 147 Nd, 151 Pm, 105 Rh, + Mo + Fission 127 Sb 103 Ru, 153 Sm, 91 Sr, Products titania 132 Te, 237 U, 93 Y, 95 Zr 99 Mo + 127 Sb + NaOH  Mo recovered on a titania column  Fission products  Acid wash followed by hydroxide strip  ~85% of fission products passed through; >90% removed after first wash  Column step completed in < 1 hour  99.3% Mo loaded; 98.4% Mo stripped 9

THE ELECTROCHEMICAL PROCESS Irradiated LEU target  Dissolve Al in NaOH Decladding  Dissolve U-foil in NaHCO 3 CCD - CCD - CCD - PEG PEG PEG Al(OH) 4 NaOH Aluminum  Precipitate U + FP with CaO CCD - CCD - CCD - PEG PEG PEG NaHCO 3 Uranium  Alkaline Mo-product solution to current U, Np, TRUEX TRUEX TRUEX Precipitation purification process CaO Pu, FP’s TALSPEAK TALSPEAK TALSPEAK 99 Mo sorption

THE ELECTROCHEMICAL DISSOLVER  Anode / Cathode connections to a Magna-Power supply. BEFORE  SS basket with external heating  ~2L of solution AFTER

THE ELECTROCHEMICAL PROCESS  Al dissolved in ~30 minutes  Operated at 9 V and N 2 Stir Feed Motors 40 Amps  Gases swept with N 2 Thermo- couple Dissolver  15 grams of LEU dissolved in 3.5 hours Mixing Collection Vessel Tanks (98%)  600 mL of carbonate Anode In-line solution after Basket Filter dissolution 12

PRECIPITATION & PRODUCT  Clear color  Uranium precipitated  pH 13.0 with ~100 grams CaO  Tc-99m, Mo-99, I-131  Water rinse  Trace amounts of 237 U  10 µm in-line filter  Fission Products  Strong signals from uranium and Fission 133 I Products 99 m Tc 99 Mo 131 I 135 Xe 135 Xe 99 Mo 99 Mo 132 I

Mo-99 RECOVERY IODINE RECOVERY CCD CCD CCD - PEG - PEG - PEG Al Digest Trace I-133 No Mo-99 ~ 32 µCi I-133 ~ 28 µCi Mo-99 CCD CCD CCD - PEG - PEG - PEG U Digest ~ 1.6 µCi I-131 2 µCi Mo-99 TRUEX TRUEX TRUEX Precipitation 11 µCi I-133 26 µCi Mo-99 TALSPEAK TALSPEAK TALSPEAK Product 0.9 µCi I-131 92% 30-60% Mo-99 Recovered Iodine Recovered

CONCLUSIONS  Two frontend processes were developed and tested at Argonne to treat irradiated LEU foil for Mo-99 production.  An acid process used nitric acid to dissolve LEU followed by Mo-99 recovery/separation on a titania column.  An electrochemical process utilized anodic dissolution of LEU in carbonate followed by calcium precipitation.  Both processes demonstrated > 90% Mo-99 recovery.  Both processes can be fed into known Mo-purification procedures. 15

ACKNOWLEDGMENT  Vakhtang Makarashvili (Argonne)  Bill Brown (Argonne)  Roman Gromov (Argonne)  ANL Central Shops  Sergei Chemerisov (Argonne)  James Grudzinksi (Argonne)  Lohman Hafenrichter (Argonne)  Steve Sherman (ORNL) Jim Byrnes (Argonne)  CSE Division (ANL)   Dave Rotsch (Argonne)  NNSA / GTRI Work supported by the U.S. Department of Energy, National Nuclear Security Administration's (NNSA's) Office of Defense Nuclear Nonproliferation, under Contract DE-AC02-06CH11357. Argonne National Laboratory is operated for the U.S. Department of Energy by UChicago Argonne, LLC. Thank you. Questions? 16

The submitted manuscript has been created by UChicago Argonne, LLC, Operator of Argonne National Laboratory (“Argonne”). Argonne, a U.S. Department of Energy Office of Science laboratory, is operated under Contract No. DE-AC02-06CH11357. The U.S. Government retains for itself, and others acting on its behalf, a paid-up nonexclusive, irrevocable worldwide license in said article to reproduce, prepare derivative works, distribute copies to the public, and perform publicly and display publicly, by or on behalf of the Government. 17

EXTRA SLIDES 18

ARGONNE HD-TARGET PROCESSES BASE ACID  Dissolution  Dissolution  Iodine  Iodine  NO x gas  NO x gas  UREX  UREX  Purification  Purification 19

URANIUM TARGETS U Metal UO 2 orthorhombic fluorite ρ = 19.1 g/cm 3 ρ = 10.9 g/cm 3 U-U = 2.8Å U-O = 2.3Å

URANIUM DISSOLUTION  Al dissolved in ~30 minutes  Operated at 9 V and 40 m Amps  Gases swept with N 2  15 grams of LEU dissolved in 3.5 hours (98%)  600 mL of carbonate solution after dissolution

DISSOLVED URANIUM SOLUTION 1  Light-green color U(VI) No U(IV) Absorbance  pH 10.0 0.5 0 99 m Tc 400 500 600 700 800 900 nm 97 Nb 97 Zr 133 I 237 U

PRECIPITATION AND FILTRATION  Uranium precipitated with ~100 grams CaO  Mixing vessel rinsed with water  Slurry fed through 10 µm in-line filter  ~1.2 L product solution

Mo PURIFICATION  Product solution contacted with AG-MP-1 anion exchange resin  Iodine and Molybdenum retained  K d (Mo) = ~150 mL/g  α -Benzoin oxime precipitated Mo-carrier after acidification http://www.sigmaaldrich.com/catalog/product/aldrich/b8 908?lang=en&region=US

FUTURE  More low-burnup and DU tests at ANL  Improve hot-cell compatibility  High-burnup tests  More XRD studies on Na-Ca-UO 2 -CO 3 precipitate

waste

Low Temperature Low Pressure Alkaline Dissolution Process Scheme Irradiated LEU foil target Mechanical Decladding Dissolution of NaOH/ CCD CCD CCD - - - PEG PEG PEG NaOH NaAl(OH) 4 Al barrier 1 M U electrolysis CCD CCD CCD - - - PEG PEG PEG NaHCO 3 CaCO 3 , Ca(OH) 2 , CaO TRUEX TRUEX TRUEX U precipitation An, FP’s solid filtrate TALSPEAK TALSPEAK TALSPEAK 99 Mo sorption 27

Target Dissolution Nickel Uranium Foil Foil _ Two-step process + Stirrer 1. Dissolution of Al fission recoil barrier using NaOH 2. Anodic dissolution (1 M NaHCO 3 in a beaker with Ni/SS clips intense stirring)  8.8g DU foil dissolved in 45 minutes (0.0042 g/min·cm 2 )  22g foil dissolved in 90 minutes Ni Mesh Basket 20 cathodic 15 H 2 1 M NaHCO 3 10 Current, A 5 Ni basket 0 U foil -5 U 0 → UO 2 → UO 2+x → UO 2 (CO 3 ) n 2-2n -10 anodic -15 O 2 U(VI) fast surface rate-limiting step -20 reactions 2.0 1.5 1.0 0.5 0.0 -0.5 -1.0 E vs Hg/HgO/0.1 M NaOH, V 28

Uranium Precipitation XRD of Precipitate • Addition of CaO excess is  CaCO 3 followed by a filtration  2+ )-(CO 3 ) phase A mixed Na-Ca-(UO 2 step  Would also contain insoluble FPs, Pu, • The precipitate is very Np easy to filter using a paper  SEM and TEM analysis will follow filter under gravity Precipitate Filtrate Solution  2- <1 mM CO 3  Trace U  Saturated Ca(OH) 2  Would also contain soluble FPs  pH 12.7  2- is co-precipitated ! No MoO 4  K d ( 99 Mo) ~ 340 mL/g on AG-MP1 29

New Dissolver Design 30

Tc-99 m  The most important medical isotope in the world http://backreaction.blogspot.com/2 008/11/technetium-99.html

Mo-99/Tc-99 m PRODUCTION • Fission of highly-enriched U-235 can make profitable amounts of Mo-99 • Canada produces half • 2016 deadline • A domestic supply is needed http://www.cins.ca/scat.html Aecl.ca

Mo-99 PRODUCTION STEPS DIGEST CHEMICAL PURIFY URANIUM SEPARATIONS Mo-99 Well known for acid, Works with what about base? acid and base.

LEU IRRADIATE ACID BASE PROCESS PROCESS Mo-99 Mo-99 PURIFICATION

SELECTED FISSION PRODUCT H 2 O CHEMISTRY Zr La Mo I 100 130 MASS ACID BASE Zirconium Zr 4+ , ZrO 2 Zr(OH) x 4 -x Polymer Molybdenum MoO 2 2+ MoO 4 2- Lanthanum La 3+ La(OH) x 3 -x Iodine I 2 I - , I 3 - , IO 3 -