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Contract No. W-7405-eng-26

Chemical Technology Division

MOLTEN-SALT FLUORIDE VOLATILITY PILOT PLANT: RECOVERY OF ENRICHED URANIUM FROM ALUMINUM-CLAD FUEL ELEMENTS

W. H. Carr* L. J. King F. G. Kitts W. T. McDuffee F. W. Miles**

"Present address: Allied Chemical Nuclear Products, Inc., Florham Park, N. J. 07932 *Present address: General Electric Company, Midwest Fuel Recovery Plant, Morris, Ill. 60450

APRIL 1971

This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Atomic Energy Commission, nor any of $heir employees, nor any of their contractors, rrubcontraetors, or their employees, makes any warranty, express or Implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disclosed, or represents that its use would not infringe privately owned rights.

OAK RIDGE NATIONAL LABORATORY O a k Ridge, Tennessee operated by UNION CARBIDE CORPORATION for the U.S. ATOMIC ENERGY COMMISSION

iii

CONTENTS

Page

............................ Introduction . . . . . . . . . . . . . . . . . . . . . . . . Description of the Molten-Salt Volatility Flowsheet . . . .

Abstract

3. Preparation of the K-Zr-Al Fluoride Dissolvent Salts

....

4. Descriptive Data for the Irradiated U-A1 Fuel Elements Processed in Runs RA-1, -2, -3, and -4 . . 4.1 4.2 4.3

.... .... Configuration, Composition, and Method of Handling . . Irradiation History . . . . . . . . . . . . . . . . . . Significant Fission Products . . . . . . . . . . . . .

3 8 11 11 12

5. Dissolution of Fuel Elements, and Volatilization of UF6 fromthe Melt by Fluorination

..............

5.1 Dissolution of Spent Assemblies in Molten K-Zr-A1

................... ....... ......

Fluoride Salts Conversion of the Dissolved UF4 to uF6, and Transfer of the Volatilized UF6 from the Melt to NaF Absorber Beds Using Elemental Fluorine 5.3 Distribution of Significant Fission Products During Dissolution, Fluorination, and Desorption

5.2

6. Quality of UF6 Product, and Uranium Inventory

.......

6.1 Fission Product Decontamination Factors for the

6.2 6.3

8. 9.

..................... ......................

17 22

26 34

uF6

Product.. Primary Radioactive Contaminants in the Product from RunRA-4 Removal of Technetium, Neptunium, and Plutonium from the uF6 Product Nonradioactive Cationic Contaminants Cumulative Material Balances for Salt and Uranium

................... 6.4 ......... 6.5 ... Release of Fission Products to the Environment . . . . . . . Acknowledgments . . . . . . . . . . . . . . . . . . . . . . References . . . . . . . . . . . . . . . . . . . . . . . . .

41 43 43

47 50 52

Page 10

Appendixes 10.1

10.2 10.3

10.4

10.5 10.6

........................

Appendix A: Fission Product Content of I r r a d i a t e d Fuel Elements Appendix B: Dissolution of Fuel Elements Appendix C: Decontamination of P i l o t Plant Equipment Appendix D: Corrosion of Vessels Appendix E: Radiation Safety Appendix F: Index of V o l a t i l i t y P i l o t Plant Log Books..

................... ...... ....................... .......... ............ .....................

54 56 62

63 68

MOLTEN-SALT FLUORIDE VOLATILITY PILOT PLANT:

RECOVERY OF

ENRICHED URANIUM FROM ALUMINUM-CLAD FUEL ELEMENTS W. H. Carr, L. J. King, F. G. K i t t s , W. T. McDuffee, and F. W. Miles

ABSTRACT W e have developed and successfully demonstrated a moltensalt fluoride-volatility process f o r recovering decontaminated uranium from spent uranium-aluminum alloy ORR and LITR f h e l elements clad i n aluminum. The f a c i l i t i e s and t h e process were e s s e n t i a l l y t h e same as those used f o r zirconium- and Zircaloy-clad f u e l s except that an aluminum-potassium-zirconium fluoride mixture w a s used as the carrier salt. The development program included the processing of both unirradiated and irradi a t e d f u e l elements. Fission product decontamination factors ( f u e l t o UF6 product) f o r t h e uF6 products i n t h e four hot runs were generally lo6 t o l o l o . The uranium concentration i n t h e s a l t a f t e r fluorination ranged from less than 0.1 t o 9 ppm; t o t a l nonrecoverable losses i n scrubbers and waste salt averaged less than 0.9%. Dissolution of t h e f u e l elements required 8 t o 17 h r f o r 90% completion, and 12 t o 25 h r f o r 100% complet i o n ; average dissolution rates were 0.6 and 0.4 kg of aluminum/ h r , respectively. The release of f i s s i o n products t o t h e atmosphere during the first three hot runs w a s confined t o 120 m C i of technetium, 5 m C i of ruthenium (which occurred i n one run), undetectable amounts of 1311, and 47 t o 60 C i ( c a l c u l a t e d ) of 85Kr.

In t h e fourth run, an ORR element t h a t had been cooled less than four weeks w a s processed. Radiation exposures t o personnel w e r e controlled within tolerable l i m i t s . The decontamination factors (DF's) i n t h i s run ranged from 2 x lo5 t o l o 8 . Two major exceptions were the DF's f o r 99M0 and 125Sb, which were 36 and about 500 respectively. The roduct had a high radioactivity l e v e l d u e t o t h e presence of 8 3 7 u , The uranium concentration i n t h e salt after fluorination i n t h i s run w a s approximately 0.1 ppm, and t h e t o t a l nonrecoverable loss w a s 0.4%. In t h e short-cooled run (RA-4), 24 C i of 85Kr and 2260 C i of 133Xe (calculated) were released t o t h e atmosphere during hydrofluorination; 20 C i of technetium, along with barely detectable amounts of 1311 Ci) and ruthenium (low3C i ) , were released during fluorination. The "Kr and 133Xe were released over an extended period so t h a t a c t u a l ground-level concentrations did not exceed a s m a l l fraction of t h e maximum permissible concentrations (MPC'S) a t any t i m e .

1. INTRODUCTION

A t ORNL, uranium-zirconium alloy fuels containing highly enriched

uranium and i r r a d i a t e d t o burnups of 32% have been successfully processed,using the molten-salt fluoride-volatility process, a f t e r cooling periods as short as s i x months.'

However, the anticipated quantity of

spent uranium-zirconium alloy f u e l i s i n s u f f i c i e n t t o justia a molten salt-fluoride v o l a t i l i t y processing plant.

To be economically practi-

c a l , such a plant would have t o be capable of processing other types of nuclear reactor f u e l as well. In order t o provide a l a r g e r processing load (and thereby improve plant economics) f o r a molten-salt

fluoride-volatility plant, a devel-

opment program was undertaken t o extend t h e use of t h e v o l a t i l i t y process mentioned above t o the decontamination and recovery of unburned

uranium i n uranium-aluminum a ~ o y( u - ~ 1 ) f u e l elements.

This program

culminated i n t h e processing of highly enriched f u e l s , w i t h approximately 30% of the i n i t i a l 235U fissioned, within 25 days of discharge from a reactor operating at a flux of greater than 2 x 1014 neutrons cm-2 sec-1. Five cold runs and four hot runs were made i n t h e V o l a t i l i t y P i l o t Plant* (VPP) at ORNL.

I n the first two cold runs, multiplate aluminum

dummy f u e l elements were dissolved; dummy f u e l and unirradiated UF4 were processed i n the remaining three cold runs.

In t h e hot runs, f u e l

from t h e Low Intensity Test Reactor (LITR) and t h e Oak Ridge Research Reactor ( O M ) w a s processed a f t e r cooling periods ranging from l e s s than 30 days t o 18 months.

The hot runs were followed by an additional

dummy dissolution, four barren salt flushes, aqueous decontamination, corrosion measurements, and, f i n a l l y , equipment sectioning. The purpose of t h i s report i s t o present t h e information obtained i n the VPP runs; primary emphasis i s on the four hot runs (LITR f u e l 1

cooled 18 months and ORR f u e l cooled 6 months, 3 months, and 25 days respectively).

The molten-salt fluoride-volatility flowsheet and a

summary of the operational experience and r e s u l t s f o r t h e processing

*Located i n c e l l s 1 and 2 and adjacent areas, Bldg. 3019.

' W i

of aluminum-base fuels i n t h e VPP are included.

The i r r a d i a t i o n his-

t o r i e s of t h e f u e l elements t h a t were dissolved, t h e compositions of t h e molten salts employed, and the two p r i n c i p a l chemical reaction steps

of hydrofluorination and fluorination are discussed i n d e t a i l .

Special

a t t e n t i o n i s given t o the distribution and release of f i s s i o n products. Data regarding the purity of the m6 products, t h e nonrecoverable urani u m losses, t h e uranium and s a l t balances, and t h e radiation experience

(radiation i n t e n s i t y measurements as well as personnel exposures) are a l s o presented.

Equipment design and performance, and operating proce-

dures that d i f f e r from those used i n t h e e a r l i e r zirconium program, are described elsewhere.2

Complete discussions of equipment and procedures i n t h e zirconium program may a l s o be found elsewhere. 1,334 I n t h i s report, t h e molten-salt fluoride-volatility process as applied t o U-A1 f'uels i s referred t o as "the v o l a t i l i t y process," even though it i s only one of many v o l a t i l i t y processes.

This p a r t i c u l a r

v o l a t i l i t y process e s s e n t i a l l y consists of four steps:

(1) dissolution

of f u e l elements i n a molten fluoride s a l t , by reaction with anhydrous I

HF, t o produce UF4 and AlF3; (2) removal and p a r t i a l decontamination of

t h e uranium by fluorination with fluorine, which converts t h e UF4 t o uF6; ( 3 ) f u r t h e r purification of t h e uF6 by passage through beds of NaF

and MgF2 p e l l e t s , u t i l i z i n g sorption and desorption techniques; and

( 4 ) recovery of t h e

uF6 product by solid-phase condensation.

DESCRIPTION OF THE MOLTEN-SALT VOLATILITY FLOWSmET

A simplified schematic diagram, o r flowsheet, of t h e equipment

used i n the VPP i s shown i n Fig. 2.1.

I n accordance with t h i s flowsheet,

each i r r a d i a t e d f i e 1 element w a s brought i n t o t h e p i l o t plant i n t h e shielded carrier-charger (FV-9501).

The carrier-charger w a s centered

over t h e charging chute, and t h e multiplate element was lowered (by a zirconiumwire) d i r e c t l y i n t o t h e 5-in.-diam

A more

d e t a i l e d description of t h e process follows i n t h e next section. 2.

bottom section of the

dissolver (FV-lOOO), where it rested on t h e d i s t r i b u t o r plate. most instances, a second element was placed on top of t h e f i r s t .

In The

ORNL Ow9 64-9234 R-l

caustic Scrubber FV-150 KOH Surge Tank FV- 152

c0 Transfer ;~;5pd' Tank

P I -

NOTES: S h o M vessels ore those which were deconfamihoied NORMAL FLOW LEGâ&#x201A;ŹND U HF -b--B-

Molten SON Service

Fig. 2.1.. Schematic Flowsheet of Equipment in Volatility Pilot Plant.

K-Zr-U

fluoride salt w a s mixed as a powder, added t o t h e barren salt

t r a n s f e r tank (FV-1500), melted (mp = Q J ~ O O O C ) , sampled, and transferred t o the dissolver ( a l s o called t h e hydrofluorinator), which had been preheated t o 600-650'~. hydrous HF w a s d i s t i l l e d batchwise i n t o the system through t h e HF cooler (FV-2004) t o the HF accumulator

(FV-1006).

A stream of li-

quid HF w a s pumped from t h e accumulator t o the HF vapor generator (FV-l2O7), where it w a s vaporized; t h e vapor w a s superheated t o 100Â°C,

metered, and fed t o t h e dissolver beneath t h e d i s t r i b u t o r plate. HF dissolved i n the s a l t and reacted w i t h t h e elements

The

t o produce

A l F 3 and UF4 (which became p a r t of t h e melt), and hydrogen.

The hydro-

gen, unreacted HF vapor, and i n e r t gases from instrument purges l e f t the dissolver and entered t h e f l a s h cooler (FV-1001), where they cbnt a c t e d a second stream of l i q u i d HF pumped from t h e HF accumulator. Solids t h a t had been entrained from t h e dissolver were removed here, i n the condenser (FV-2001), and i n t h e HF catch tank (FV-1003). Solids

I i

t h a t collected i n the catch tank remained there u n t i l the end of the '

dissolution step, when the contents of the tank were transferred t o

the caustic neutralizer (FV-1009). In turn, the contents of t h e neut r a l i z e r were pumped t o the hot chemical waste.

The HF w a s d i s t i l l e d

from the HF' reboiler (FV-1005), and w a s collected i n the HF accumulator a f t e r passing through f i l t e r FV-7001C and the HF cooler.

The hydrogen

and i n e r t gases passed through t h e -5OOC â&#x201A;ŹIF condenser (FV-2005), which removed traces of HF, and were then bubbled through approximately 2 KOH i n the caustic neutralizer.

This off-gas stream then joined the

c e l l off-gas stream, received another caustic scrub, passed through absolute f i l t e r s (AEC type),and was then released t o the atmosphere through the 3020 stack. After dissolution w a s complete (i.e.,

there w a s no further decrease

i n off-gas volume o r HF inventory),the HF recirculation w a s stopped, ~

and the m e l t was sparged with nitrogen t o remove the dissolved HF.

The

molten salt (mp = QJ55OoC) w a s then transferred t o t h e fluorinator

(FV-lOOX which had been preheated t o approximately 600Oc; enough salt was l e f t behind t o f i l l the horizontal section of the connecting l i n e

and thus form a plug,or "freeze valve;' t o separate t h e hydrofluorinat i o n equipment and t h e fluorination equipment. The fluorinator w a s sparged w i t h nitrogen t o mix t h e new charge w i t h any "heel" t h a t remained from t h e previous run, and then t h e salt

w a s sampled by lowering a copper l a d l e (on a chain) d i r e c t l y i n t o t h e molten salt and "dipping" a s m a l l volume from beneath t h e surface of the melt.

From such a "feed salt" sample, the ufanium and f i s s i o n

product concentrations a f t e r hydrofluorination (md before fluorination) could be determined.

After t h e sampling procedure w a s complete, ele-

mental fluorine w a s passed through the melt t o convert t h e UF4 t o uF6 and t o thereby remove it from the melt.

"he only important higher

fluorides of f i s s i o n products t h a t were formed during fluorination and were not retained by NaF were MoF6, TeF6, and TcF6. Fluorine at 1 2 t o 60 p s i g w a s supplied by a tank t r a i l e r parked outside Bldg. 3019; it entered through a NaF t r a p ( i n l e t end heated t o 100째C), which removed HF.

The p u r i f i e d fluorine flowed i n t o t h e fluori-

nator through a draft tube, which induced circulation of t h e melt and improved gas-liquid contacting.

Volatile UF6 a v o l a t i l e f i s s i o n product

fluorides, 'and unreacted fluorine passed out of t h e fluorinator through the movable bed absorber (FV-105); the higher fluorides of most of the f i s s i o n products are nonvolatile,

and they remained i n t h e salt.

This

gas stream passed, f i r s t , through a section of t h e movable bed' absorber containing NaF p e l l e t s a t 400OC.

Here, t h e bulk of t h e f i s s i o n product

fluorides t h a t were v o l a t i l i z e d o r entrained were deposited; t h e F2, e s s e n t i a l l y a l l of t h e U , Mo, Np, and T c , and s i g n i f i c a n t quantities of Zr, Nb, Ru, 12, and Te proceeded t o t h e next section containing NaF

p e l l e t s a t 150째C.

The m6 and most of t h e contaminants were sorbed

under these conditions, while t h e fluorine, MoF6, and some tellurium fluorides passed on t o the chemical t r a p (FV-121), which contained NaF

at ambient temperature. by t h i s trap.

The MoF6 and any t r a c e s of uF6 were removed

Fluorine w a s removed i n a caustic scrubber (FV-150).

The off-gas w a s then vented t o the c e l l off-gas system (which included another caustic scrubber) and was f i l t e r e d before being exhausted. Generally, a s m a l l amount of tellurium w a s released i n t h e off-gas.

7 Desorption of UF6 (but not f i s s i o n products) from the 150째C NaF p e l l e t s i n t h e movable bed absorber w a s achieved by heating t o 40OoC i n a fluorine sweep. impurity t r a p (FV-UO),

This gas stream passed, f i r s t , through the containing MgF2 at 100째C, f o r the removal of

any technetium and neptunium present and, then, through t h e product f i l t e r (FV-723) i n t o the s m a l l product cylinder (FV-223) maintained a t -7OOC by d r y ice--trichloroethylene

slush.

About 70 t o 100% of t h e

uF6 was removed; the remainder was deposited i n the uF6 cold t r a p s

(FV-220 and FV-222) held a t -50 t o -6oOc.

The off-gas exited through

the chemical t r a p (FV-121) t o remove any traces of uranium and then passed t o t h e caustic scrubber as previously.

After HF had been

flashed from the UF6 product under vacuum a t O째C, the small product cylinder w a s removed from the system, weighed, sampled, and assayed t o confirm weight, composition, and enrichment of t h e product. After fluorination, t h e melt i n t h e fluorinator w a s sampled t o determine t h e degree of removal of UF4 f r o m t h e salt,

Analytical re-

sults were received (usually <3 pg of uranium per gram of s a l t ) before a portion of the NaF p e l l e t s from the 40OoC section of t h e movable bed absorber w a s dumped i n t o the fluorinator.

In t h e event t h a t t h e uran-

i u m concentration w a s higher than desired, t h e fluorination could be continued u n t i l an acceptable value of residual uranium w a s obtained. The NaF p e l l e t s t h a t were transferred t o the fluorinator were from t h e l o w e r section ( 4 O O O C ) of t h e absorber; since these p e l l e t s

were the f i r s t t o be contacted by t h e fluorination off-gas, they had the highest concentration of sorbed f i s s i o n products.

The salt w a s

sparged with nitrogen t o a i d i n the p e l l e t dissolution; then another waste s a l t sample w a s taken t o determine t h e amount of uranium held by the p e l l e t s .

After t h i s uranium analysis (usually <8 ppm) w a s

received, t h e waste salt was transferred t o a waste salt can (FV-112) located inside a shielded c a r r i e r .

Enough salt w a s l e f t i n t t h e trans-

f e r l i n e t o form a freeze valve, as was done f o r t h e molten s a l t l i n e between the dissolver and fluorinator.

After cooling, t h e waste salt

c a r r i e r was transported t o t h e burial ground, where t h e waste salt can w a s dropped i n t o an underground vault f o r long-term storage.

PREPARATION OF THE K-Zr-A1 FLUORIDE DISSOLVBNT SALTS

The ternary salt KF-ZrFb-AlF3 w a s considered f o r the dissolvent i n the aluminum campaign i n the VPP since Thoma, Sturm, and Guinn5 had shown it t o be t h e most s u i t a b l e solvent system f o r t h e processing of aluminum-uranium fuels. The use of t h i s salt would permit us t o operate a t r e l a t i v e l y low temperatures, thus minimizing corrosion and avoiding the d i f f i c u l t i e s t h a t would otherwise r e s u l t from t h e low melting point (660Oc)

of aluminum.

A portion of t h e revised triangular plot of liquidus temperature

as a function of composition f o r the system KF-ZrF4-AlF3 i s shown i n Fig. 3.1.

This portion includes t h e only region w i t h melting points

less than 600Oc.

It i s easily seen t h a t any dissolution path (e.g.,

heavy, dashed l i n e s ) t h a t i s chosen t o maximize capacity w i l l start a t t h e m a x i m u m allowable melting point, cross a region of lower melting point, and terminate a t t h e maximum melting point t h a t i s allowable during fluorination.

Obviously, t h e higher t h e temperature t h a t can

be t o l e r a t e d , t h e greater t h e capacity of t h e salt f o r aluminum.

maximum melting point of 600Oc w a s chosen f o r the beginning salt.

This

permitted operation a t a temperature s l i g h t l y above t h e melting point, and s t i l l allowed f o r a reasonable temperature r i s e (due t o reaction heat) without a t t a i n i n g the melting point of aluminum.

The melting

point a t the end of dissolution w a s held t o 55OoC t o l i m i t t h e corrosive e f f e c t of elemental fluorine on the nickel fluorinator. For all four hot runs, barren

s a l t containing

64.3 t o 63.0 mole

% KF and 35.7 t o 37.0 mole % ZrF4 (mp, ~ 6 0 0 O C )w a s transferred t o t h e hydrofluorinator.

These salts, when mixed with t h e s m a l l "heels"

c a r r i e d over i n t h e hydrofluorinator, gave t h e desired i n i t i a l compositions.

The binary salts were prepared by dry-mixing commercial grades

of K2ZrF4 (containing 27% potassium and 32.1% zirconium, by weight) and ZrF4 (54 t o 54.5% zirconium).

The granular salts and mixtures were

handled i n a i r ; no special precautions were taken, except t h a t a reasonable e f f o r t was made t o minimize t h e time during which t h e salt was exposed t o moisture.

ORNL Dwp. 64 -7982

-0

POINT A E F G

COMPOSITION, mole 96 COMPOSITION, mole % KF ZrFq AIF3 ZrF4 AIF3 POINT KF 55.5 28.5 16.0 J 64.7 21.1 14.2 64.3 35.7 0.0 K 57.0 10.0 33.0 55.0 30.6 14.4 0.0 M 75.3 24.7 56.9 18.6 24.5 0.0 N 66.0 34.0 3.2 63.9 32.9

Fig. 3.1. Portion of the KF-Z~FI+-AI.F~ System That is Applicable to the Dissolution of U-A1 Fuels.

The mixture w a s melted i n a closed vessel; a nitrogen purge w a s maintained through the vapor space, and a nitrogen sparge w a s used t o promote mixing when melting began.

There w a s no evidence (thermal) of C

H20 evolution at any temperature, although t h e odor of HF w a s quite evident. barren

The melt w a s c l e a r and had a low viscosity; t r a n s f e r of t h e

salt i n t o the system w a s accomplished without d i f f i c u l t y .

Although t h e binary salt j u s t described provided a highly satisfactory s t a r t i n g material f o r processing U-Al f'uels, we wanted t o demonstrate dissolution w i t h a salt i n i t i a l l y containing aluminum.

The

aluminum could e a s i l y be supplied by leaving an aluminum-rich heel i n the hydrofluorinator.

The remainder of the charge would then consist

of K2ZrF6 and KF'; t h e substitution of low-cost KF f o r ZrF4 would greatly reduce the cost of the salt components.

Although t h e addition of s o l i d

Kz&F~-KF, as outlined, could not be done i n VPP equipment because of design limitations , p a r t i a l t r a n s f e r ( i . e . ,

terminating a molten salt

t r a n s f e r a t a specified point) had been demonstrated during t h e z i r conium campaign. We encountered d i f f i c u l t y i n all early attempts t o prepare a t e r I

nary KpZrF4-AlF3 barren

salt.

Although salt materials (i.e.,

commer-

c i a l K2ZrF6 and K F , and specially-dried AlF3) were carefully pre-mixed, fusion was always incomplete.

A sediment, having t h e consistency of

coarse sand, w a s evident i n the bottom of t h e melt vessel, while a layer of undissolved material f l o a t e d on t h e surface of t h e melts. When t h e ternary phase diagram (liquidus temperature vs composition) 1

w a s examined, revised data showed that these melts had liquidus temperat u r e s about 85Oc higher than those indicated by t h e former triangular plot.

However, mixtures whose compositions had been adjusted t o t h e

data of t h e new diagram exhibited t h e same c h a r a c t e r i s t i c s .

Although

the dissolvent salt could not be fused completely, experiments showed

t h a t t h e product salt, a f t e r aluminum dissolution and HF sparging, w a s

a single-phase melt and could be transferred readily i n t h e l i q u i d state.

Hence, f o r i n i t i a l operations i n t h e VPP, t h e ternary salt

w a s prepared by mixing, melting as much as was possible, allowing t h e

salt t o freeze, and breaking t h e frozen salt i n t o chunks of

non-homogeneous solid. ver.

These chunks were then dropped i n t o t h e dissol-

This experimental procedure w a s used only f o r cold runs ( l i q u i d

binary s a l t w a s used i n the hot runs); it w a s reasonably s a t i s f a c t o r y f o r cold runs, but would be hazardous if used t o charge a highly contaminated hydrofluorinator. A f t e r the l a s t run i n which i r r a d i a t e d fuel w a s processed

a cleanout run w a s made using a dummy aluminum element.

(RA-4),

In the latter

run, a molten W-ZrFq-AlF3 ternary salt w a s t r a n s f e r r e d successfully i n t o t h e system. The t a r g e t composition w a s 63-21-16 mole % KF-ZrF4-AlF3, which w a s thought t o be t h e composition of a ternary e u t e c t i c with t h e lowest-melting salt i n t h e immediate area of i n t e r e s t .

Heating t o

650Oc appeared t o c l a r i f y t h e m e l t ; however, a f e w inches of sediment w a s found i n t h e bottom of t h e melt vessel.

When ZrFq w a s added, t o

increase t h e ZrFq content t o 22.33 mole %, t h e sediment disappeared. It did not reappear upon cooling t o 600Oc.

Thus t h e f i n a l composition

of t h e barren salt t h a t w a s t r a n s f e r r e d i n t o t h e system w a s 61.94-22.33-

15.73 mole % KF-ZrFq-AlF3.

DESCRIPTIVE DATA FOR THE IRRADIATED U-A1 FUEL ELEMENTS PROCESSED I N RUNS RA-1, -2, -3, AND -4

4.1

Configuration, Composition, and Method of handling

The i r r a d i a t e d U-Al f u e l elements processed i n t h i s campaign were of t h e multiplate box type, w i t h curved f u e l p l a t e s , t h a t i s used i n t h e Oak Ridge Research Reactor (ORR) and t h e Low I n t e n s i t y T e s t Reactor (LITR).

When received, t h e elements had o v e r a l l lengths of 26 t o 27

i n . ( t h e end boxes had been removed), and t h e o v e r a l l cross section of each w a s e s s e n t i a l l y 3.0 by 3 . 1 i n .

There were nineteen 2.8-in.-wide

fuel-bearing p l a t e s brazed o r swaged between two s l o t t e d ( i n e r t ) s i d e plates. 11 c

The cladding and s t r u c t u r a l material w a s aluminum, and t h e

active" core a l l o y was approximately 18%uranium--82% aluminum.

The

17 "inside" fuel plates were 24.625 i n . long (active length, 23.625 i n . ) and 0.050 in. t h i c k o v e r a l l (core alloy, 0.020 in. ; cladding, 0.015 i n .

each s i d e ) .

The two f u e l p l a t e s comprising two sides of t h e box had

t h e same alloy core as t h e "inside" plates,but were 2.520 in. longer (overall), and

t h e aluminum cladding w a s 1-1/2 times as thick ( f o r an

overall thickness of 0.065 in.

P r i o r t o processing, t h e elements were stored i n a canal under approximately 10-1/2 f t of water.

To t r a n s f e r t h e elements t o t h e pro-

cessing plant, a shielded ( 9 in. of lead) charger-carrier w a s lowered t o the bottom of the canal.

A 0.060-in.-diam

zirconium wire, which

had been threaded between the p l a t e s of the elements and through an

opening i n the closed end of t h e c a r r i e r , w a s used t o s l i d e t h e elements, horizontally, from a loading t a b l e i n t o the cavity of t h e chargercarrier.

The c a r r i e r w a s closed, hoisted from t h e canal, and, a f t e r

appropriate procedures t o prevent the spread of contamination, brought i n t o t h e processing plant.

It w a s placed on t h e enclosure that shielded

t h e top of the charging chute (centered above t h e dissolver), w i t h i t s

axis v e r t i c a l and the closed end up.

The c a r r i e r drawer and the charg-

ing chute valves were opened, and the elements were lowered d i r e c t l y i n t o t h e 5-in.-diam

section of t h e dissolver

by means of t h e zirconium i

wire that w a s used t o load t h e elements i n t o t h e c a r r i e r .

The wire was

cut while s t i l l under load (the elements were allowed t o drop a f e w inches); t h e severed wire withdrew i n t o the charging chute s u f f i c i e n t l y t o c l e a r the valves closing t h e top of the chute. 4.2

I r r a d i a t i o n History

The elements processed i n runs FtA-2,

-3, and -4 had been used i n

ORR cores, while t h e FtA-1 charge w a s p a r t of an LITR fueling. In the O M , elements are usually i r r a d i a t e d f o r periods of 10 t o 1 4 days; then they are removed from t h e reactor and allowed t o cool f o r 12 t o 180 days before being reinserted f o r t h e next cycle.

Burnups of 2 t o

10% per cycle a r e accumulated over four o r f i v e cycles u n t i l a t o t a l burnup of about 28 t o 32% i s achieved.

(Here, burnup i s defined as the

number of atoms of 235U t h a t have fissioned, divided by t h e number of atoms of 235U i n i t i a l l y present; thus burnup does not include the 235U depletion suffered i n t h e formation of 236U by neutron capture.)

The i r r a d i a t i o n cycles f o r the f u e l elements t h a t were processed i n t h e VPP during the aluminum campaign are shown i n Table f u e l elements

74C, 47C,

4.1. For

~ O D ,58D, and 93D, each l i n e represents an

i r r a d i a t i o n cycle f o r a specified time (shown i n hours) i n the reactor a t t h e indicated average neutron f l u x f o r t h e elementâ&#x20AC;&#x2122;s p a r t i c u l a r location.

Both the burnup achieved during each cycle and the cooling

( o r decay) period observed p r i o r t o t h e subsequent cycle are given. The decay time l i s t e d f o r the l a s t cycle f o r each element i s the time from reactor shutdown t o the beginning of the dissolution s t e p i n the This value i s usedthroughout t h i s report t o designate t h e cool-

VPP.

ing time f o r the element(s) i n each run. In the LITR, elements are placed i n the core, where they remain u n t i l the desired burnup i s achieved.

New elements a r e i n i t i a l l y placed

on t h e periphery of the core and l a t e r , during the appropriate shutdown, a r e transferred t o the center section. Table

The i r r a d i a t i o n times l i s t e d i n

4.1 f o r the f u e l elements processed

times the elements were i n t h e reactor.

i n run RA-1 a r e the t o t a l

These times are much longer

than those for ORR fuels because of the much lower f l u x i n the LITR.

4.3 Significant Fission Products The quantities of the principal f i s s i o n products thatwere present i n the i r r a d i a t e d f u e l elements at t h e time of t h e i r dissolution a r e shown i n Table 4.2.

These f i s s i o n products are divided i n t o two

s l i g h t l y overlapping categories:

nuclides contributing a s i g n i f i c a n t

p a r t of t h e t o t a l radioactivity, and nuclides t h a t , due t o v o l a t i l i t y at various times during processing, must receive special consideration regardless of the amount present.

The quantities given i n the t a b l e

were calculated, using t h e CRUNCH code, and take i n t o account t h e a l t e r nate periods of i r r a d i a t i o n and decay described i n Sect. 4.2.6

(The

complete tabulation of the quantities of f i s s i o n products present i s given i n Table A-1 i n Appendix A . ) The most important determinant of the radioactivity contributed by a p a r t i c u l a r nuclide at a specific cooling time i s i t s h a l f - l i f e

Table 4.1.

Irradiation History of Fuel Elements Processed i n the VPP During the Alminum Campaign

No.

Fuel Element No.

RA-1

U03

0.18

9417

21.9

552

3-5-63

9-9-64

RA-1

N349

0.18

ll417

25.3

582

2-5-63

9-9-64

RA-2

74c

1.48 1.08 2.17 2.17 2.31

242 185 261 224

6.0 2.7 8.6 6.7 8.3 32.3

Run

(10'; cm

Flux neutrons sec-1)

Irradiation Time

(hr)

291 1203

47c

RA-2

1.10

1.42 1.74 2.22

2.27

60D

RA-3

1.48 1.48 1.08 1.72 2.09

296 415 276 18 291

1296 80 229 325 292

276 1202

58D

RA-3

1.65 1.30

RA-4

93D

U -'P

59 361

Burnupa

(b)

(days)

5.4

180.2

6.9 0.5 8.4

30.6

13.0

5.5 5.4

32.4 17.3 111.7 77.0

1.6 7.7

8.6 27.9

1.58 1.38 0.88 2.35

290 297 301 240 1128

7.9 7.6 6.2 6.4

10.0

28.1

Date of VPP

Processing

4-23-64 10-14-64

15.9

23.4 118.8 174.0

2.0

Reactor Discharge

131.0

47.0 174.0

9.2

Date of

69.1 16.5

2.16 2.35

Decw Timeb

43.4 14.3 116.9

4-23-64 10-14-64

8-24-64 11-11-64

77.0

8-24-64

26.4 12.0 134.4 24.8

11-15-64

11-11-64

12-9-64

atoms of 23% fissioned = atoms of zjau i n i t i a l l y present'

'Decay time is cooling time between irradiation cycles or between f i n a l discharge of the f u e l element and the start of processing i n the VPP.

Table 4.2.

P r i n c i p a l Fission Products Present i n I r r a d i a t e d Fuel Elements Processed Fission product a c t i v i t i e s a r e corrected t o time of d i s s o l u t i o n ~

Nuclides Contributing S i g n i f i c a n t l y t o t h e T o t a l A c t i v i t y of I r r a d i a t e d U-Al Fuels Calculated A c t i v i t y i n Fuel Charge a t Time of Processing Half -Idf e curies Days RA-1 RA-! d-3 RA-bO Nuclidea Years

10.27

85Kr 9 0 ~ r

377 58 63 35

91Y 95 b e

9 5 ~ ,fe

' 3Ru

41 167

1.0 1311

8.05

33 ~ e 137csC

5-27 12.80

14ke

32 13.7 290 ll. 3

14 3 ~ r l44~e

2.6

6036 40 3 7973 9189 14505 2766

6306

383

7928 8333 7523 39 86

491

2338

7703

342 331 4059 442 8964

1081

1730

1676

761

567

174

47.3

10.27 ~

2.8

41 1.0

125%

l2 7Sb 127Tef

129Te 1311 132Te 13 3 ~ e

2.7 3.9 90 33 8.05 3.2 5.27

59.5

54.1

0.5

748

167 9.3

445 14.7

1.7

51.4 33.8

2766

491 14.1 100

171 8.5

24.2 148 3986 242 6.5 30.9 48.2 341 2293 2 32 2261

8243 8559 8378 4550 3126

734

Cooling Time (&,SI

3Ru

2293 2261 323

140Ba

lb7Nd 147pmf

31 70

450 3131 3874 7342 748 445

V o l a t i l e and/or Otherwise Troublesome Nuclides Calculated A c t i v i t y i n Fuel Charge at Time of Processing Half-Life (curies Days RA-1 RA-2 2-3 RA-4b Nuclidea Years 85Kr

2252

* 9 ~ r

Cooling Time (days )

567

174

%he first nuclide i n each decay chain t h a t contributes a s i g n i f i c a n t f r a c t i o n of t h e t o t a l . a c t i v i t y a t t h e time of processi n g is tabulated. A "true" t o t a l a c t i v i t y would a l s o include t h e a c t i v i t y of t h e short-lived daughters i n s e c u l a r equilibrium with t h e nuclides l i s t e d . bThe fuel charge i n t h i s run was only one element.

Two elements were used i n each of t h e o t h e r t h r e e runs.

'Less than 1/21 of t h e t o t a l of the a c t i v i t i e s tabulated. %ess

than 0.01 curie.

eIn addition t o contributing s i g n i f i c a n t l y t o t h e t o t a l dose, t h e nuclide i s a l s o v o l a t i l e under c e r t a i n conditions. A' second s i g n i f i c a n t nuclide i s shown because t h e parent-daughter r e l a t i o n s h i p s do not f u l f i l l t h e requirements f o r s e c u l a r equilibrium.

The radiation i n t e n s i t y [determined by i t s decay constant

&-sec-l 1 of a long-lived nuclide i s r e l a t i v e l y low.

'I2 0.6 (A) =

On t h e

other h a k , a short-lived nuclide may almost completely disappear (i.e.,

it decreases by a f a c t o r of

%lo3a f t e r

t e n half-lives) i n a r e l a t i v e l y

short t i m e ; however, as long as any s i g n i f i c a n t f r a c t i o n remains, t h e radiation i n t e n s i t y may be quite high due t o t h e higher A.

For these

reasons, the quantities of long-lived nuclides shown i n Table 4.2 decrease only s l i g h t l y as decay times increase.

Examples are 90Sr,

lo6Ru, and 137Cs, which comprise a large fraction of the a c t i v i t y i n t h e long-cooled run, (RA-l), but a r e r e l a t i v e l y insignificant i n t h e short-cooled runs (RA-3 and RA-4).

Nuclides with intermediate half-

= 30 t o 90 d a y s ) , such as 32-day 14'Ce, lives (t 1/2 shorter cooling times. Insignificant i n run RA-1,

do not appear u n t i l 141Ce increases from

minor significance i n RA-2 t o become t h e highest l i s t e d i n t e n s i t y i n Four important s h o r t - l i v e d ( t = <6 days) nuclides 1/2 99M0, 127Sb, 132Te, and 133Xe are not encountered u n t i l run RA-4;

run RA-4.

four others

- 1311,

140Ba, 143Pr, and 147Nd -with

s l i g h t l y longer half-

4 days), a r e hardly notable until run RA-4. Obviously, l i v e s (t 1/2 the decay time i s the variable t h a t determines which nuclides w i l l have decayed t o insignificance and which remain. The cyclic i r r a d i a t i o n has an e f f e c t on t h e f i s s i o n product spec-

trum a t the time the f'uel i s discharged, but it i s not nearly as import a n t as the cooling time p r i o r t o processing. Since t h e cooling time between i r r a d i a t i o n cycles i s generally l e s s than 135 d a y s , t h e concen>290 days) nuclides increase steadily. t r a t i o n s of long-lived (t 1/2 e90 d a y s ) a l t e r The nuclides with intermediate l i v e s (30 days e t 1/2

nately grow i n and decay t o give, when p l o t t e d , a "sawtooth" curve The short-lived ( t e14 days) 1/2 nuclides decrease t o near 50% or below on each cooling cycle.

having gradually increasing maxima.

DISSOLUTION OF FUEL ELEMENTS, AND VOLATILIZATION OF UF6 FROM THEMELT BY FLUORINATION

5.1 Dissolution of Spent Assemblies i n Molten K-Zr-A1 Fluoride S a l t s During t h e dissolution (hydrofluorination) of U-Al f u e l elements, HF i s vaporized i n a steam-jacketed v e s s e l , superheated t o 100째C i n an

e l e c t r i c a l l y heated c o i l , and fed t o the dissolver (hydrofluorinator) through a l i n e maintained at an average temperature of approximately 500째C (by autoresistance heating).

The HF, which enters t h e dissolver

beneath a d i s t r i b u t o r p l a t e , i s assumed t o be a t t h e temperature of t h e salt by t h e time it contacts t h e element.

Although t h e HF feed rates

may appear low (20 t o 130 g/min), they represent f a i r l y high volumetric r a t e s at 400 t o 600Oc.

For example, each gram of HF, f u l l y dissociated,

represents about 0.1 f t 3 a t h O o C and 0.127 f t 3 at 60OoC (average salt temperature). The conditions and r e s u l t s (including data f o r t h e f u e l charge,

salt compositions, dissolver temperatures, dissolution r a t e s and times required t o a t t a i n 90% and 100% completion, and HF feed r a t e s , consumpt i o n , and u t i l i z a t i o n ) f o r t h e t e n aluminum dissolutions i n t h e VPP

are shown i n Table 5.1. shown.

The runs were made i n t h e chronological order

The runs i n the FtA (Radioactive, Aluminum) series a r e l i s t e d i n

order of decreasing cooling time for t h e f u e l elements processed:

RA-1,

approximately 540 days ; RA-2, approximately 180 days ; RA-3 , approximately

80 days; and EA-4, 25 days.

A cleanout run, DA-3, is discussed only be-

cause it involved a dissolution of aluminum comparable t o others i n t h e

series. The elements dissolved i n the DA (Dummy, Aluminum) and UA (Uranium-

Aluminum) runs w e r e l7-plate aluminum dummies cut t o t h e appropriate length t o give t h e desired weight.

I n t h e UA runs, t h e uranium was

added as f i n e l y powdered UF4 packaged i n aluminum f o i l t h a t was less than 1m i l t h i c k .

A double charge of UF4 w a s added i n run UA-1 so t h a t a mea-

surable amount of u F 6 product (>lo0 g) could be withdrawn from t h e system after a uranium inventory near the normal steady-state value was established.

Weight mm No.

h l t e n Wt Ccamositioa (mole $1 Initial0 Find zr Al IC Zr

Ifrdroiluorinator Temeratun("C 1c . salt vapor section 8cction

M-1

7.81

64.0

36.0

53.8

30.2

16.0

610

DA-2

7.9

64.4

22.0

U.6

55.1

18.8

26.1

UA-1

6.V

592

63.9

20.8

15.3

55.1

18.0

UA-2

6.62

304

63.2

33.5

3.3

.55.0

UA-3

4.2

304

64.3

35.7

RA-1

8.80

323

64.3

35.7

RA-2

8.24

306

62.4

e-3

8.17

324

RA-4

4.2

M-3

3.96 %ere,

of Fuel Al U (kd (8)

R a te

(&in)

Mssolution T i w s 905 For 100% Completion Caapletlon . (IW) (hr) For

Al Dissolutian Rates (k/hr) TO go% To 100% Dissolution

Dissolution

auantitr

Utilization - Dissolution - A t 90% Masolution RF

cans\rmed

A t 100%

(he)

of Theoret

19.2

ll0.

9.6

9.3

lll.

29.8

22.8

480

125-80

23.1

28.4

0.304

0.275

616

500

70-120

8.9

12.8

0.799

0.616

26.9

604

460

l00-h

12.3

16.8

0.497

0.403

16.0

106.

21.8

19.2

29-1

15.9

620

505

130-50

0.271

0.255

. 15.1)

104.

8.7

8.1

57.9

32.2

9.9

615

475

125-40

16.8

22.6

0.225

0.186

10.9

115.

8.U

7.84

55.0

30.6

14.4

608

467

125-75

16.6

25.0

0.494

0.364

20.2

100.

18.0

14.3

34.6

3.0

55.4

30.8

13.8

603

475

100

15-5

24.5

0.503.

0.353

19.7

102.

18.6

13.1

62.8

36.8

0.4

55.1

32.3

12.6

575

k70

40, 100

8.0

i5.5

0,969

0.555

20.4

m6.

50.7

24.2

159

62.1

36.4

1.5

55.5

32.6

11.9

575

440

20-100

10.o

12.3

0. 378

0.33

9.8

. 105.

18.7

16.0

61.5

23.0

15.5

56.1

21.0

22.9

585

490

60,

9.7

10.7

0.37

9.9

ll2.

14.9

%ydrOflWriMtiOn" and

bCorrected for heel 'Average.

HF Flow

"dissoluticn* an used interchsngedu.

f r o m pnvlous

run.

Represents c y c l i c behavior

(within

15OC of the stated value).

100

17.6

.: .

. .. , .

.... . ... .,

I n t h e RA runs, t h e charges were LITR (RA-1) and ORR elements (RA-2 through RA-4) t h a t had been i r r a d i a t e d t o burnups of 22 t o 32%. The barren salts f o r all the runs except three (DA-2,

UA-1,

and

DA-3) were binary mixtures consisting of approximately 64 mole % KF and 36 mole % ZrF4. and ZrFq.

They were prepared from commercial-grade KzZrF6

The " i n i t i a l " salts (Table 5.1) i n four of the runs con-

tained 0.4 t o 3 . 3 mole % AlF3 supplied by heels l e f t i n the hydrofluorinator from the previous runs.

In runs DA-2 and UA-1,

the high

i n i t i a l AlF3 w a s obtained by adding a ternary salt t o the hydrofluorinator as a s o l i d ; i n run DA-3, t h e t a r g e t composition w a s a ternary e u t e c t i c with a melting point s u f f i c i e n t l y low t o permit the ternary salt t o be transferred as a liquid. A l l the "final" melts contained 54

t o 58 mole % KF and e i t h e r 29 t o 32.5 mole % ZrF4-10

t o 16 mole %

AlF3 o r 18 t o 21 mole % ZrF4--23 t o 27 mole % AlF3, depending on the i n i t i a l AlF3 content; each w a s readily transferable. --

Temperature w a s not investigated as a process variable; the t e m peratures used w e r e simply those which we believed would approach t h e

minima needed t o give acceptable operational performance. The range of the â&#x201A;ŹIF' dissolvent flow r a t e w a s also chosen from the standpoint of operational experience.

Systematic changes i n t h i s

variable were not made, and no attempt at optimization with respect t o any p a r t i c u l a r parameter w a s made.

It was not necessary t o keep the

HF flow rate low, since t h e HF t h a t w a s not u t i l i z e d i n a p a r t i c u l a r

pass w a s recycled; t h e only additional costs r e l a t e d t o high flow r a t e s were those connected w i t h pumping and heating ( o r cooling).

A t very

high gas flow rates, salt entrainment could be troublesome.

In t h e

equipment a t B l d g . 3019, the p r a c t i c a l upper l i m i t of the HF flow r a t e

(%130 g/min) i s determined by the refrigeration capacity at low HF u t i lization.

A t high utilization,pressurization of t h e off-gas system

can impose an upper l i m i t on HF feed r a t e . 1

Heat evolution during d i s -

solution w a s never a problem during t h e aluminum campaign.

A lower

l i m i t would probably be imposed by t h e control system, but t h i s was

not explored; t h e equipment functioned very s a t i s f a c t o r i l y at t h e lowest

HF r a t e used, 20 g/min.

20 I

The 9- t o 23-hr periods required f o r 90% dissolution, and the 13t o 28-hr periods required f o r 100% dissolution, a r e based on t h e actual volumes of HF consumed, which were recorded as a function of t i m e during each run.

These volumes ranged from 100 t o 116%of t h e t h e o r e t i c a l

HF consumptions. The tabulated HF u t i l i z a t i o n values range from 8 t o 50% and correl a t e r a t h e r well w i t h the HF flow r a t e s ; t h a t i s , t h e higher the flow r a t e , the lower the u t i l i z a t i o n . w a s too rapid t o be effective.

This means t h a t t h e r a t e of HF sparging

However, i f t h i s added throughput in-

creased t h e reaction r a t e even s l i g h t l y , by such means as increased turbulence, b e t t e r film coefficients, gas phase reaction, o r improved temperature distribution, it might be j u s t i f i e d . Correlation of aluminum dissolution rates w i t h run parameters i s d i f f i c u l t since t h r e e aluminum-uranium systems, t w o s a l t compositions

(as w e l l as intermediate ones), and two schemes of feeding HF were used.

The aluminum dissolution r a t e and the HF feed r a t e are p l o t t e d vs run time f o r each of the t e n dissolutions i n Appendix A. A-10

Figures A-1 through

show t h a t the average rates and times f o r dissolution a r e not en-

t i r e l y representative of what i s taking place.

Some runs show high r a t e s

i n i t i a l l y , while others s t a r t low and gradually increase f o r several hours.

Some decrease rapidly, whereas others tend t o become constant

at d i f f e r e n t l e v e l s .

A l l are characterized by a " t a i l i n g out" as t o t a l

dissolution i s approached, although t h e length of t h i s t a i l i n g varies widely. A s mentioned e a r l i e r , alminum dissolution r a t e s and the completion

of dissolution are i n f e r r e d from periodic readings of t h e inventory of l i q u i d HF i n t h e recirculating system.

These readings are subject t o

t h e usual instrument and reading e r r o r s ; t h i s m a y explain some of t h e roughness of the p l o t s , which were smoothed t o some extent by averaging. The i n i t i a l dissolution rate i s probably the l e a s t r e l i a b l e of all since changes i n inventories i n pipes, vapor spaces, t h e salt i t s e l f , i

etc. are a l s o involved. very d i f f i c u l t t o detect.

The point at which dissolution is complete i s The 90% l e v e l of dissolution is considered

t o be t h e point at which 90% of t h e HF required t o achieve 100%

dissolution ( t h e point a f t e r which no f u r t h e r decrease i n HF inventory occurs) has been consumed.

This time i s then used t o calculate t h e

90% dissolution r a t e .

Although the data f o r 100% dissolution a r e admittedly not precise, and t h e dissolution r a t e curves are somewhat e r r a t i c , t h e dissolution times and average r a t e s based on 90% dissolution a r e f a i r l y accurate and meaningful and can be used t o d r a w some v a l i d conclusions regarding t h e p r i n c i p a l variables.

By comparing runs DA-1 and DA-2, which were

similar except f o r the aluminum contents of t h e i n i t i a l s a l t charges, we must conclude t h a t t h e salt with t h e higher aluminum content e x h i b i t s dissolution r a t e s two t o t h r e e times t h a t of s a l t containing no a l u m i num i n i t i a l l y .

The short dissolution time i n run DA-3 ( i . e . ,

9.7 h r t o

90% completion) confirms t h i s . (Note t h a t t h e dissolution r a t e i n DA-3 i s only one-half t h a t of DA-2; however, t h e times a r e consistent since t h e weight of f u e l used i n run DA-2 w a s twice t h a t i n DA-3 and hence twice t h e a r e a w a s available f o r dissolution.)

The salt dissolved i n

run UA-1 w a s similar t o t h a t dissolved i n DA-2, except t h a t it contained

.:.

a r e l a t i v e l y large amount of UF4.

We believe t h a t t h e longer dissolu-

t i o n time t h a t w a s required t o achieve 90% completion i n run UA-1 (see Table 5.1) i s a s t a t i s t i c a l variation since t h e RA- runs give no indication t h a t t h e presence of uranium i n h i b i t s dissolution.

Run UA-2 w a s

similar t o DA-1 except t h a t s m a l l amounts of aluminum and uranium were

included i n t h e i n i t i a l salt; t h e same slow dissolution w a s observed. From a comparison of runs UA-3 and RA-I,, we conclude t h a t i r r a d i a t i o n has no 'discernible e f f e c t on dissolution r a t e ; dissolution times f o r equal-sized salt charges were equal, although t h e weight of metal i n t h e charges d i f f e r e d by a f a d t o r of 2.

Further j u s t i f i c a t i o n f o r

t h i s conclusion w a s provided by the r e s u l t s obtained i n run RA-2, which

w a s a duplicate of run RA-1 except t h a t t h e salt f o r RA-2 i n i t i a l l y contained 3 mole % A P 3 . r

Dissolution r a t e s were considerably higher (from t h e standpoint of t o t a l t i m e required o r i n terms of r a t e per u n i t area of aluminum) i n runs RA-3 and -4 than i n runs RA-1 and -2.

Three variables, which

could cause d i f f e r e n t dissolution rates, were: temperature, and t h e HF flow rate.

the ZrF4 content, the

The e f f e c t s of these parameters

on dissolution r a t e have not been studied.

Based on r e s u l t s i n other

runs and on previous operating experience, we would not expect t h e

first two t o cause significant changes i n dissolution rate.

In a l l the

runs p r i o r t o RA-3, the â&#x201A;ŹIF flow rate w a s increased up t o t h e maximum

(125 t o 130 g/min) as soon as possible.

A s cooling times became shorter,

however, we decided t o react the bulk of t h e elements at lower flow r a t e s and not r i s k the production of "bursts" of off-gas, which, i n turn, would cause pressurization of t h e off-gas system. i

It i s possible

that l o c a l cooling e f f e c t s of t h e excess gas o r hydraulic e f f e c t s

(especially while t h e multichannel configuration i s s t i l l i n t a c t ) could a f f e c t dissolution rate. 5.2

Conversion of the Dissolved UF4 t o UF6, and Transfer of t h e Volatilized uF6 from t h e Melt t o

NaF Absorber Beds Using Elemental Fluorine During the fluorination step, elemental fluorine i s admitted at the bottom of the fluorinator through a draft tube. t h i s tube i s t o increase circulation within t h e melt.

The purpose of In the fluorina-

t o r , the fluorine contacts t h e molten salt and r e a c t s w i t h t h e dissolved UF4.

Both the UFg t h a t i s formed and the excess fluorine pass up through

the vapor space i n t o the movable-bed (NaF p e l l e t s ) absorber.

The lower,

o r nearly horizontal,* section of t h i s absorber, which i s maintained

*The movable-bed absorber was o r i g i n a l l y constructed w i t h a horizontal section and a v e r t i c a l section. During i n i t i a l t e s t i n g , it became apparent t h a t t i l t i n g t h e u n i t down (with respect t o t h e entry point) would result i n a longer e f f e c t i v e length of horizontal section due t o t h e angle of repose of the p e l l e t s ( ~ 4 5 ' ) . Absorption i n t h e horizontal section, rather than i n the v e r t i c a l section, i s desirable since t h e horizontal piston can e a s i l y move any s i n t e r e d p e l l e t s i n t h i s section. Hence, the unit w a s t i l t e d 15O i n t h e VPP, r e s u l t i n g i n t h e displacement of the horizontal and t h e v e r t i c a l sections 15' from t h e horizontal and the v e r t i c a l axes, respectively.

. sj

a t 4OO0C, removes most of t h e f i s s i o n product fluorides t h a t a r e vola-

t i l i z e d from t h e melt; the UF6 passes t o t h e upper, o r v e r t i c a l , section (maintained a t 1 5 O o C ) , where it i s sorbed quantitatively.

A t 150째C t h e

MOF6 passes through t o a 20 t o 3OoC NaF t r a p , which i s used t o remove any UF6 that could have passed through the 1 5 O o C t r a p .

The off-gas i s

scrubbed with an aqueous c a u s t i c solution (thus removing t h e f l u o r i n e ) and then combined with t h e c e l l off-gas; the r e s u l t i n g off-gas stream i s scrubbed with caustic and, f i n a l l y , i s f i l t e r e d and vented t o t h e

stack. The conditions and r e s u l t s f o r seven fluorination-sorption and desorption runs and two simulations are presented i n Table 5.2.

In each

run, the melt-gas i n t e r f a c e w a s within t h e lower 15-1/4-in.-ID

cylin-

d r i c a l section of t h e fluorinator.

The temperature of t h e salt w a s

measured by a thermocouple i n a well s i t u a t e d beneath t h e surface of the m e l t .

This temperature w a s held only as high as w a s considered

necessary t o maintain the salt as a l i q u i d ; the purpose of such temperat u r e control, of course, w a s t o minimize corrosion.

The density of t h e

s a l t w a s measured by a differential-pressure c e l l t h a t w a s placed across

two bubbler probes stationed 5 i n . apart. The uranium concentration i n t h e fluorination s a l t w a s obtained by analyzing a sample of t h e salt i n t h e fluorinator.

S a l t samples

were taken from beneath t h e surface of t h e melt by using a copper l a d l e suspended on a Monel chain.

Each sample t h a t w a s w i t h d r a w n w a s a c t u a l l y

a composite of t h e dissolution product from t h e current run (including t h e h e e l l e f t i n t h e hydrofluorinator from t h e previous run) and t h e heel l e f t i n t h e f l u o r i n a t o r from t h e previous run.

The l a t t e r h e e l

contained very l i t t l e uranium; thus it acted only as a diluent. Fluorination r a t e s f o r t h e hot runs were 6 and 11 s t d liters/min. The lower rate w a s used i n t h e first p a r t s of t h e runs i n an e f f o r t t o produce more uniform loading of t h e NaF p e l l e t bed during t h e period when uF6 evolution w a s most rapid.

The higher r a t e , which served t o

improve the c i r c u l a t i o n of t h e salt through t h e draft tube, was used l a t e r i n t h e runs i n an attempt t o minimize t h e amount of uranium l o s t

Table 5.2.

Conditions and Results for Fluorination-Sorption and Desorption Runs in the VPP

First two runs were practice runs with no uranium; solid UF4 Was added to the salt in the next three runs; in the last four runs, irrsdiated fuel elements were processed.

Rm NO.

DA-1

cycle' Fl-Sorp

Volume (liters)

77.0

Fluorination Salt Level Above TempDraft Tube erature Density (in.) ("C) (g/cc)

10.9

605

2.40

DeSOrp

M-2

F1-Sorp

73.2

9.7

615

2.20

UA-1

F1-Gorp Desorp

63.3

6.4

560

2.24

Fl-Sorp Desorp

74.0

F1-Sorp

63.3

UA-2

UA-3

10.0

2.46

2.314

1.808

RA-2

6.4

600

2.46

0.976

1.048

RA-4

Time (inin)

F2 Utiliration (%)

U in Waste salt (ppd

Absorber Temueraturec ("C) zone 1 Zone 2 Zone 3 Zone 4

6, 13

17s 15 100

395 392

170

300

150 300

290

1, 18

8, 30

180

Amb.'

Amb.d

Amb.'

392 392

125 375

160 380

141 375

398 400

155 375

137 418

143 400

ll0

6. 15. 11

101, 20, 19

230

7, 12

81. 24 218

3.7

76, 24 271

2.0

402 406

130 375

171 375

150 375

1.4 1.6

71, 24 191

2.5

401

400

149 400

146 430

146 380

1.6 1.6

6. 11 1

2 .%

2.6

Fl-Sorp Desorp

70.6

F1-Sorp No. 1

69.7

8.5

550

2.46

0.616

6. 11

73, 22

1.5

396

175

154

138

1.5

Fl-Sorp No. 2

54.3

3.3

550

2.50

1.433

6. 11

68; 22

2.9

393

162

154

136

5.2

255

395

398

428

400

16.7

8.8

610

2.45

DeSOrp

RA-3

Ratg (SW )

DeSOrp

RA-1

F2 Flow

U Concentration (g/kg salt)

6, ll

F1-Sorp No. 1

64.8

6.9

550

2.52

0.994

6, 11

65, 20

2.5

405

146

160

150

8.65

F1-Gorp No. 2 Desorp

40.0

-1.6

550

2.48

0 769

6, 11

60, 18

1.3

406

148

170

149

6.95

226

403

385

433

398

13.80

F1-Sorp

54.3

6, 11 1

55, 20215

398 400

154 395

147 430

153 398

e0.1-

DeSOrp

3.3

550

2.51

1.351

3.2

5.35 ~

~~~

'LF1 = fluorination; ~ o r p= sorption; ~ c s o r p= desorption. b S W = standard liters per minute. Where more than one rate is given, the rates correspond to the times listed in the adjacent column. 'Zone 1: 400째C zone; zone 2: temperature transition zone; zone 3: U F g sorber; zone 4: cleanup section ( i n flow-path only during desorption). 'Amb = ambient.

bsl) 3

i n t h e waste salt.

Fluorine u t i l i z a t i o n w a s calculated by assuming

t h a t one mole of f l u o r i n e reacted per mole of UF4 present i n the salt.

The absorber temperatures shown i n Table 5.2 a r e averages over t h e fluorination period. The most important values i n Table 5.2 a r e t h e uranium concentra-

t i o n s i n t h e waste s a l t (shown i n t h e l a s t column).

The f i r s t value

l i s t e d f o r runs UA-2 through RA-1 and f o r RA-4 i s t h e uranium content of the melt a f t e r fluorination; t h e second value i s t h e concentration a f t e r t h e UFg had been desorbed from the NaF-filled movable-bed absorber and a batch of NaF p e l l e t s had been discharged from the horizontal sec-

t i o n , which i s maintained at 4OO0C, i n t o t h e fluorinator.

I n runs RA-2

and -3, two fluorinations were made without an intervening desorption o r p e l l e t discharge.

All salt samples (except i n run UA-1) were taken

from the f l u o r i n a t o r i n t h e manner described e a r l i e r .

As i n t h e zirconium campaignY1 t h e uranium concentration i n t h e melt after fluorination seemed t o vary i n a random fashion when it w a s only a few parts per million.

Any correlation of t h i s concentration

w i t h fluorination time, temperature, i n i t i a l concentration, irradia-

t i o n , batch s i z e , o r melt depth w a s not obvious i n the range of t h e variables investigated

The amount of uranium returned t o t h e f l u o r i n a t o r i n t h e NaF p e l l e t s

from t h e 40OoC absorber bed should be a function of sorption and desorpt i o n conditions, but not of fluorination conditions.

The higher l o s s e s

i n runs RA-2 and -3 were probably t h e result of loading the bed v i a two fluorination-sorption cycles p r i o r t o desorption.

Possibly, longer

times at t h e maximum desorption temperature (400OC) would have reduced t h e amount of uranium t h a t w a s returned. I n instances where t h e amount

of uranium returned might be g r e a t e r than t h a t which could be economically discharged, t h e melt could be fluorinated repeatedly u n t i l t h e uranium concentration reached t h e desired terminal l e v e l .

5.3

Distribution of Significant Fission Products During Dissolution, Fluorination, and Desorption

During the dissolution step, unreacted HF w a s condensed i n t o a catch tank and then revaporized and returned t o t h e recycle system. Hydrogen ( a

Any p a r t i c u l a t e matter w a s collected i n t h i s catch tank.

reaction product) and noncondensables passed through t o a tank containing caustic ( 2 M KOH), called the caustic neutralizer ( K N ) , where the gaseous mixture was bubbled through approximately 3 f t of liquid.

the end of each run, the HF heel i n t h e catch tank was discharged t o the KN; thus, all of the f i s s i o n product material t h a t w a s collected from t h e hydrofluorination off-gas w a s represented by samples from the KN.

In a somewhat similar manner, the unreacted fluorine from the

fluorinator, a f t e r passing through an NaF absorber and a uranium clean-up t r a p , entered a scrubbing tower where it was reacted w i t h c i r culating 2

KOH.

of the radioactive

A sample of t h e scrub solution w a s representative

material t h a t w a s evolved during fluorination or

desorption and removed i n the scrubber. The amount of radioactive material t h a t was released t o t h e stack w a s calculated from the analyses of a charcoal t r a p through which a samThe radio-

ple of gas from the bottom of the stack had been drawn.

active noble gases were assumed t o be released upon dissolution (measurements of the radioactivity l e v e l of the off-gas being discharged through the stack were made, but t h e r e l a t i v e l y s m a l l amount of radioactive material t h a t w a s released during dissolution precluded quantitative r e s u l t s using t h i s method ).

The molten salt w a s not sampled u n t i l it

had been transferred t o the fluorinator a f t e r hydrofluorination.

Here

it was sampled both before and after fluorination, and after the NaF p e l l e t s had been discharged from t h e absorber bed.

Agreement between

samples taken before and after fluorination w a s not goodc

I n general,

t h e values obtained f o r most f i s s i o n products were higher (as predicted

by laboratory investigation) a f t e r fluorination; t h i s w a s unexpected, based on representative samples of a homogeneous melt.

In one run,

radiochemical analyses were made of samples withdrawn before and a f t e r

t h e N U p e l l e t discharge, but the s e n s i t i v i t y of the r e s u l t s w a s not high enough t o permit an estimation of the a c t i v i t y returned t o t h e

m e l t by the p e l l e t discharge.

In all cases, the "salt" a c t i v i t y

values presented l a t e r w i l l be the higher of the pre- and postfluorinat i o n samples. The amounts of significant f i s s i o n products found i n the salt, HF, fluorine, off-gas scrubber solutions, the UF6 product, and the stack effluent a r e tabulated l a t e r i n t h i s section f o r each hot run.

As mentioned i n Sect. 4 . 3 , a p a r t i c u l a r f i s s i o n product may be import a n t because it i s troublesome ( v o l a t i l e , e t c . ) o r because it contributes s i g n i f i c a n t l y t o t h e general radiation background.

In t h e e a r l i e r

runs, t h e long-lived f i s s i o n products were t h e most important; however, as cooling times decreased, t h e nuclides with intermediate and short

half-lives received more attention.

In run RA-4,

emphasis was on nu-

clides with half-lives on the order of eight days o r less; these nuclides had not been present previously i n appreciable quantities i n molten salt processing.

In comparing quantities of the various nuclides

expected at t h e time of processing (machine calculation by the CRUNCH code) with t h e t o t a l s actually found by radiochemical analyses, we concluded t h a t agreement of these values within a f a c t o r of 2 i s satisfactory and t h a t agreement within an order of magnitude i s not t o be considered grossly i n error. In run RA-1 (two LITR elements cooled 19 months, Table 5.31, t h e long-lived nuclides 90Sr, lo6Ru, 137Cs,

and 144Ce provided the major

percentage of the radioactivity; i n addition, some 95Zr-95Nb

(tl12=

65 days) and 127Te (tl12= 90 days) were s t i l l present a f t e r a cooling period of 570 days.

Only a low y i e l d of 12'Sb (0.023%) w a s obtained,

but t h i s nuclide has a long h a l f - l i f e (2.7 years) and i s expected t o be v o l a t i l e (as i n the zirconium campaign)

during hydrofluorination.

Results confirmed t h a t t h e antimony and t h e tellurium were v o l a t i l e , while the other nuclides remained i n the s a l t . *

In run RA-2 (two ORR elements cooled 175 days, Table 5.4), three additional nuclides

- 89Sr,

lo3Ru, and 129Te

- a r e present;

however,

Table 5.3. Distribution of Significant Fission Products in Run RA-1

Fission Product

salt (curies)

Caustic Used in HF System (curies1

Distribution of Radioactivitya,b in Run Caustic Used in Off-Gas uF6 Fp System Scrubber Product bci1 (mci1 (PCi1

Released to Stack (mci1

Total (curies)

Calculated Radioactivity of Fuel Element (curies)

Anal./Calc.

(%I

377

95 ~ r

<lC

38.3

125a

l(O.1) 1

27Te

0.95

9.3

10.2

0.03

1.7

1.9

237

323

1162

2338

aQuantities were determined by analysis. bNumbers in parentheses are percentages of the total, as determined by analysis. 'Below

analytical limits.

'Quantity 'Quantity

ignored in total. of 1311 a l s o less than 1 mCi.

Table 5.4. Distribution of Significant Fission Products in Run RA-2

Fission Product

Salt (curies)

Caustic Used in HF System (curies)

Distribution of Radioactivityasb in Run Caustic m6 Used in Fp System Scrubber Product fmci) (mci) (PCi)

Released to Stack (mci)

8 9 ~ r 1850

Total (curies)

Calculated Radioactivity of Fuel Element (curies

hd./Calc.

(dl

1850

2252

430

450

9O~r

403

95 ~ r

2716

<lC

2716

3874

95Nb

6036(99.2 1

6087

7342

3Ru Io6Ru %b 127,129~~ 137cs l44ce

51.4(0.84 )

<lC

19.8 31.8( 98.7) 20ge

4.24(93.1) 527(99.9)

0.15 (0.47)

259(0.81)

1.31(99.2)

10( 0.8

0.15 (3.3)

122(2.7)

0.44(0.08)

3626

74 <lC

lO(0.22)

30(0.66 ) <1

<lC

19.

748

2.6

32.2d

445

7.2

1.32

14.7

8.8

4.55

5.4

527

383

138

3626

7703

a Quantities were determined by analysis.

bNumbers in parentheses are percentages of the total, as determined by analysis. C

Below analytical limits.

' I n addition to the quantities listed, the following quantities of fission products were found on the nickel wool trap (FV-154) between the Fg system caustic scrubber and the off-gas scrubber: 44 pCi of lZ9Te, 164 pCi of lo6Ru, and 17 pCi of lo3Ru. eValue was disregarded since it was later found that niobium coextracts with antimony in the analysis.

I i i

i I

i ~

they do not represent any new chemical species.

In instances where we

are interested i n two isotopes of t h e same element (e.g.,

strontium

and ruthenium), we d i d not f e e l t h a t it was necessary t o obtain a comp l e t e s e t of analyses f o r both nuclides since it was assumed that t h e i r chemical behavior i s similar.

The material balances f o r t h e nonvola-

t i l e elements i n t h i s run are considerably b e t t e r than i n run RA-1. Again, antimony and tellurium were v o l a t i l i z e d during hydrofluorinat i o n , and tellurium and ruthenium were p a r t i a l l y removed from t h e salt during fluorination.

A s m a l l quantity of tellurium w a s c a r r i e d through

t h e off-gas scrubber and released t o the stack. In run RA-3 (two ORR elements cooled 80 days , Table 5.5), three ,

new species (91Y, 1311, and 1 4 0 B a ) and one additional isotope (141Ce) were present; of these, only the 1311 proved t o be of any concern. The iodine balance i n t h i s run w a s excellent; greater than 90% remained

i j

i n t h e salt, while about 9% w a s found i n the caustic used i n t h e HF system.

Traces were a l s o found i n t h e other two scrubbers and on t h e

stack sampler.

major processing s t e p s , and about 5% of t h e t o t a l found was released

As previously, the tellurium w a s v o l a t i l i z e d i n both

t o the stack; a s m a l l amount of ruthenium w a s a l s o released.

Most of

t h e antimony and some ruthenium and niobium were found i n t h e caustic used i n the HF system.

The material balances f o r all nuclides except

lo3Ru, 106Ru,and 127r129Te ranged from 67 t o 152%, which i s considered t o be excellent. !

In run RA-4 (one ORR element , cooled 25 days) , t h e presence of

greater than 2-1/4 kilocuries of 133Xe made t h e t o t a l a c t i v i t y from

the noble gases 100 t i m e s t h a t of t h e 85Kr; however, high d i l u t i o n

factors and high MPC values made rapid release safe.

In RA-4, t h e

t o t a l release required more than 10 h r ; i f it had been accomplished over a 3-hr period, the ground-level concentration would s t i l l have been less than 1% of the MPC. L

The significant f i s s i o n products f o r run

5.6.

RA-4 are shown i n Table

Three short-lived nuclides, absent i n runs RA-1,

present i n t h i s run:

99M0,

ll1Ag, and 132Te.

-2, and -3, were

The 132Te represented

Table 5.5. D i s t r i b u t i o n of S i g n i f i c a n t F i s s i o n Products i n Run RA-3

Fission Product

salt (curies

D i s t r i b u t i o n of Radioactivityarb i n Run Caustic Caustic Used i n Used i n Off-Gas m6 HF System F2 System Scrubber Product ( c u r i e s) (mci1 (mci1 (uti)

(%I

c0.2

8,636

7,973

108

(0.1

13,950

9,189

152

21,012

14,505

145

NSe

0.28 0.28

10.3 (99.9

Anal. /Calc.

113

0 .Ob( 'XI .01)

0.32( 1.2)

Total (curies)

6,036

co.1

l.l(l.4)

(mci)

6 ,836

0.28

152(0- 7 )

Released t o Stack

Calculated Radioactivity of Fuel Element (curies)

NSe

Q5c

136 (0.5 1

78.16

2,766

2.8

27.53

491

5.6

10.32

3(0.03)

14.1

0.56( 32.3)

21(1.21)

90( 5 .le)

1.74'

0.76 ( 8 -9

8.56

0.02

CO.1

411

342

120

0.02

(0.1

453

331

137

0.15

(0.1

4,059

2,727

0.33

co.1

7,773

8,964

271

8.5

0.64

101

aQuantities were determined by analysis. bNumbers i n parentheses a r e percentages of t h e t o t a l , a s determined by a n a l y s i s . 'As

t h e t o t a l of two isotopes.

$elow a n a l y t i c a l limits. e~~ = not sought.

% d u e was disregarded s i n c e it was l a t e r found t h a t niobium c o e x t r a c t s with antimony i n t h e a n a l y s i s . gFifty-five microcuries of t e l l u r i u m was found on t h e charcoal t r a p (FV-155); 52 p C i was found on t h e 400째C N i wool t r a p (FV-154). There was a t o t a l of <1 p C i of ruthenium on both t r a p s .

Table 5.6. Distribution of Significant Fission Products i n Run RA-4

Fission Product

Salt (curies )

Distribution of Radioactivitya~bi n Run Caustic Caustic Used i n Used i n Off-Gas w 6 W System Fz System Scrubber Product (curies) (curies) (mci) tmcij

Released o Stack lcuries)

Total (curies)

Calculated Radioactivity of Fuel Element (curies )

Anal. /calc

89~r

7,960(100 )

0.311

0.004

0.47

7.960

6306

126

91Y

9,965(100)

0.226

C<O.OOld

0.32

9,965

7928

126

95 ~ r

13,400(100 )

0.3U

<O.OOld

0.07

13,600

8333

161

95Nb

12.400(99.4

0.23

12,476

7523

166

71.8

148

15.9

3986

0.40

242

1.80

9 ~ o

lo3Ru 106Ru lllU

75.5(0.6)

65(90;5)

2.80( 3.90)

~(69.2)

0.032(0.20)

2.8(64.3) 22(100) d

0.019(0.44)

0.005 2.26(3.15

1660(2.31) d

4.75(29.9)

<0.1

1.5( 34.5)

g0.02 d

g0.003 d

4.35

0.006

0.5

2.08( 82.4 )

0.44( 17.4)

5.2(0.2)

2.52

<O.0Old

157

125s

g0.5

lZ7Te

lZ9Te

13.5

341

1311

<0.43d

O.SS(0.8)

0.25

go. O l d

110

2293

109(99.2

37 (64.4 1

0.586( 1.02)

12.8( 22.3)

0.15

137cs

232(99.9)

0.126(0.1)

<0.006'

0.32

140Ba

8,856( 1001

0.351

~0.04

141ce

7.050

144ce

5,617(100 1

0.076

0.13

32Te

.O O l d

e C<O.OOld

6.32 (110)

57.5*

(I)

6.5

4.8

232

152

153

8.243

8856

107

7,050

8559

5,617

4550

123

'&antities were determined b ; ~analysis. bNwnbers i n parentheses are percentages of t o t a l , as determined by analysis. 'Not detected by gama scan of charcoal i n stack sampler. %elow analytical limits. eSpeciflc analyses are not available; quantities are assumed t o be proportional t o the nuclide f o r which data are presented. h i s does not include about 1 curie ( t o t a l ) of 1 2 7 ~ 1 2 9 ~ 1 3 2t T h ae t is held up on t h e 400째C wool t r a p in the e x i t gas l i n e f o r t h e caustic scrubber used i n t h e fluorine system.

33 a 60% increase of the 127s129Te t o t a l , and the 127Sb a c t i v i t y was a

f a c t o r of 5 times the customary a c t i v i t y of 125Sb.

The presence of

99M0 presented a serious problem since special procedures had been T

necessary t o prevent s t a b l e molybdenum from following the UF6 product.

I n general, the results obtained i n RA-4 confirmed the trends observed i n the e a r l i e r runs; however, the data of RA-4 should be considered t o be more significant because of t h e appreciably higher a c t i v i t y levels-. (Note t h a t i n Table 5.6 some of t h e u n i t s a r e l a r g e r than i n preceding tables.

As previously, tellurium and ruthenium were v o l a t i l i z e d

principally during fluorination; antimony and iodine, along w i t h fract i o n a l percentages of niobium and cesium, were evolved during t h e hydrofluorination step.

Some antimony a l s o appeared i n the uF6 product

and i n the caustic used i n t h e fluorination system.

Although greater

than 90% of the 99M0 was found i n the salt, t h e remainder w a s well distributed.

Analyses showed t h a t greater t h a n 3%of the 99M0 w a s present

i n each of the two system scrubbers, 0.16% was present i n the off-gas scrubber, and 2.3% was found i n t h e uF6 product. i n the stack sampler.

None w a s detected

About 20 curies of tellurium and t r a c e s of 1311

and ruthenium were released through t h e stack, but t h e ground-level concentration w a s calculated t o be much l e s s than 5% of M P C , assuming a 40-min release time and a stack-to-ground

d i l u t i o n f a c t o r of

lo5

(see Sect. 7). With only one exception, no unusual o r unexpected events occurred

as a result of processing at t h e higher a c t i v i t y l e v e l of run RA-4

(about 1-1/2 times t h a t i n RA-3, even though twice as many f u e l elements

were processed i n RA-3).

Upon sparging the salt p r i o r t o withdrawing

t h e first samples from t h e fluorinator, there w a s a rapid incrkase i n the background radiation i n the v i c i n i t y of t h e chemical t r a p (FV-121). The increase continued u n t i l sparging was stopped; the radiation intens i t y then decreased i n a t y p i c a l manner ( i . e . , an exponential decay curve).

The r a t e of decrease indicated t h a t 1321(tlI2= 2.4 h r ) , t h e

daughter of 132Te, w a s the source of t h e a c t i v i t y . 1

allowed t o remain s t a t i c (i.e.

, only

The salt had been

low-rate nitrogen purges were

allowed t o flow) i n t h e hydrofluorinator f o r several hours while

temperatures were adjusted preparatory t o t r a n s f e r r i n g the salt t o t h e fluorinator.

Evidently, during t h i s period, a s i z a b l e amount of 1321

grew i n from t h e decw of 132Te, and largely remained i n t h e salt u n t i l fluorination.

Sparging i n t h e f l u o r i n a t o r caused t h e 1321 t o be c a r r i e d

through t h e 1 5 O o C NaF absorber and heated piping t o t h e ambient-temperat u r e t r a p (FV-121), where it deposited on t h e r e l a t i v e l y cold surface. The decay product was s t a b l e xenon, which could not be traced. Table 5.7 presents a comparison of t h e important data (from t h e standpoint of radiation background) from Tables 5.3-5 .6, expressed on

a percentage basis.

QUALITY OF UF6 PRODUCT, AND URANIUM INVENTORY

The uranium hexafluoride product was collected by desublimatio-n

as it w a s desorbed from the NaF p e l l e t s i n t h e movable-bed absorber

(FV-105).

As t h e bed w a s heated from 150 t o 4OO0C, t h e U F 6 ' 2 N a

complex decomposed.

The UF6 w a s c a r r i e d , via f l u o r i n e , through a porous

metal f i l t e r i n t o t h e product cylinder, which was cooled with a dry ice--trichloroethylene mixture.

A f t e r t h e product had been collected,

HF w a s removed from t h e cylinder by f l a s h i n g (7 min at O째C under vacuum); t h e cylinder w a s then weighed and sampled.

Table 6.1 gives t h e weights,

the uranium contents, and t h e i s o t o p i c analyses l e c t e d i n t h e four hot runs.

of t h e products col-

Results f o r an a d d i t i o n a l run (UA-3) a r e

included t o show an "unburned" i s o t o p i c composition.

Obviously, all

t h e uranium added t o t h e system i n a p a r t i c u l a r run does not appear i n t h e product f o r t h a t run since, i n t h i s seven-run campaign, about onet h i r d of t h e uranium w a s held up i n cold t r a p s , NaF t r a p s , s a l t h e e l s ,

etc.

This holdup l e d t o a blending, o r an overlap, of products from

run t o run.

The analysis of t h e uranium product, as reported, i s t h e

r e s u l t of a coulometric determination i n which a l l t h e uranium w a s assumed t o be 238U;

the values i n t h e table have been corrected f o r

t h e appropriate i s o t o p i c analysis.

A l l of t h e values a r e i n reasonable

agreement with( t h e t h e o r e t i c a l uranium content of u F 6 (67.36%) f o r t h e assays involved.

c *

Table 5.7.

Comparison of Fission Product Distributions i n Runs RA-1,

Fi ssion Product

RA-1

RA-2

89~r '5,

82 12

95Nb

45 55

113

70 83

2.8

126

RA-3 b

RA-4

126

152

161

145

166

0.7

0.6

1.5

0.8

2.6

2.8

0.4

7.2

5.6

1.8

RA-2

RA-1

%looc SlOOC

Sl0OC

%looc

98.5

99.2

RA-3

0.7

0.5

0.2

1.2

0.4

157

(

RA-4

Fa-1

RA-2

RA-3

RA- 4

100

99.3

3.9 1.4

Total FP Found in Fa System Caustic Total FP Found in Fuel

)

108

ll1AB

99.4

90.5

3.1 29.9

%looc

98.5

69.2

98.7

98.1

64.3

2.3

0.8

0.5

100

34.5 0.04

10.2

8.8

1 2 7 r 1 2 9 ~ ~ 1.9

5.4

1311

101

99.9 29

0.6

99.2

99.9

3.3

32.3

4.8

8.9

13*Te 120

153

140Ba

137

l4ke 144ce

RA-2

'Ru

137cs

RA-1

'MO

'*'Sb

RA-4

91Y 952,

RA-3

-2, -3. and -ha

Total FP Found i n Salt Total FP Found in Fuel

T o t a l FP Found i n Fuel Total FP Found i n HF stem Caustic Total FP Found ;n Fuel (Calculated Quantity of FP i n Fuel) (

138

82.4

~d 93.1

0.05

107

123

0.08

<0.4

0.10

0.8

2.7

0.03

99.9

%looc

99.9

Sl0OC

17.4

1.2

64.4

1.02 0.10

<20

57.6 91.1

99.2

%ata were obtained Apm Tables 5.3-5.6

0.8 22.3

100

%looc

100

and are expressed as percentages.

bLess than 0.01%. 'These values are approximate because only analyses of the s a l t were obtained; however, subsequent runs showed these species t o be essentially nonvolatile. s e s e values were meaningless since an e r r o r in the analytical procedures was discovered during the Fa-4 analysis.

Table 6.1.

U r a n i u m Input and Product Data

RA-1

Fa-2

R u n No. RA-3

RA-4

UA-3

Feed: -

Uranium i n fuel after burnup, g

323

306

324

159

30ha

300

470

410

100

300

Product : Weight gain of uF6 product cylinder, g

% u in

65.64

66.76

67.05

63.94

67.33

2 3 4 ~

0.98

1.16

1.17

1.15

0.82

235~

86.87

83.00

82.52

82.54

91.40

236u

3.22

5.93

6.09

5.89

0.32

238u

8.93

9.91

10.22

10.42

7.46

uF6

product

Isotopic analyses, %

As UF4; unirradiated.

6.1 Fission Product Decontamination Factors f o r the UF6 Product Most of t h e f i s s i o n products remained i n t h e melt and were discarded with t h e waste salt.

Portions of t h e f i s s i o n products t h a t were

v o l a t i l i z e d - d u r i n g fluorination were sorbed on t h e NaF p e l l e t s i n t h e movable-bed absorber (FV-105).

The s m a l l f r a c t i o n s of these radio-

nuclides t h a t were desorbed along with t h e uF6 usually constituted t h e contaminants i n t h e product cylinder.

(However, it i s possible t h a t

some of these contaminants were deposited i n e a r l i e r runs and were acquired from t h e connecting piping. )

The decontamination f a c t o r s

(DF's) f o r various f i s s i o n products i n t h e four hot runs a r e shown'in Table 6.2. The f i r s t of the four columns f o r each run l i s t s t h e a c t i v i t y f o r each nuclide i n the f u e l , as calculated by t h e CRUNCH code.

(Techne-

tium-99 i s omitted from t h e t a b l e and i s discussed i n Sect. 6.3.) This value, which i s given on a per-milligram-of-uranium

basis, was

obtained by dividing by t h e number of milligrams of uranium remaining i n t h e fuel as determined by the reactor operations burnup calculations.

(This i s a l s o t h e uranium content of t h e f u e l f o r accountability purposes. ) The second and t h i r d columns show t h e r e s u l t s of the analysis of the

mtj

product f o r radioactive contaminants.

Multiple samples of t h e uF6

product (as a l i q u i d ) were analyzed; t h e second column gives t h e a c t i v i t y level (on t h e date of analysis) t h a t was thought t o be most repres e n t a t i v e f o r each nuclide.

Sampling w a s done by heating and tipping

the product cylinder; hence t h e tabulated values do not r e f l e c t t h e

degree of p u r i f i c a t i o n t h a t would have been afforded by vapor sampling

of the uF6.

I n t h e t h i r d column, t h e a c t i v i t y i s placed on a per-

milligram-of-uranium

basis.

The DF's i n the first three runs were q u i t e s a t i s f a c t o r y , generally

lo7 t o

lo6 (lo6 t o l o 7 ) ( t o 1.8 x lo8) and

The 125Sb DF, which decreased t o , i n run RA-3, i s an obvious exception. The lo6Ru DF w a s low

ranging from

1O'O.

i n the first two runs, but i t s improvement i n RA-3

1.4

38 Table

6.2.

Decontamination Data for

uF6

Products i n R u n s RA-1, -2, -3, and -4

Activities of feeds w e r e calculated by CFiUIiCH code; activities of products were obtained by analysis.

4.14 x 1O1O

8 9 ~ r

90~r

2.59 x 109

275

0.42

6.2 x

109

3.27 x

4860

109

7.3

x 10~0

9.7 x

l o 6 1.63

109

6.6 x

lo6 1 . W x lo4

1.0x 107

>4.6

109

x 10l1

1.4 x

lo6

5.0 x

lo7

q.5

109

x loll

4.4 x io6 7.81 x 103

1.3

107

x 109

1.2 x 1010 5.70

x 109

<3.3

x 108

<5.0 x i o 6 a . 7 0 x 104

lo7

1.14 x lo8 1.78 x lo5

q400

a1.0

>4.0 x

109

5.46 x 1O1O

<9500

a4.1

>3.9

>3.5 x 109

6.30 x 1O1O

~9300

a3.8

1.5 x 109

9.94 x 1010

<go00

<13.3

4.5 x 108

2.76 x 109

io4

5.4 x

106

91Y 95 ~ r

2.14 x

lo8

~600

e0.91

4.80 x

lo8

<350

<O.

~5400

4.1

2.35 x 1 0 4

99~0

1 0 3 ~ ~

lo6Ru

me 1.15 x 109

. 5.90

x 105

1.3 x

900

lo6

3.23 x

lo9

3.4

510

105

6.3 x

lo6

6.41 x 107

5200

7.9

8.1 x

lo6

1.07 x lo8

<6500

<9.7

>1.1x 107

1 2 7 ~ 1 2 9 ~ ~ 1311

1.24

105

184

1.8 x 108

9.69 x 107

4.50

104

67.5

1.4 x lo6

1.85 x 109

1.75 x 104

25.9

7.1

107

x 109

e6.7

>2.2

x 106

x 1010

5.83 x

io?

a . 8 x 104

32Te 137cs

2.22 x 109

~350

>4.2 x 109

2.78 x

109

1.8

1200

1.5 x

109

140Ba 144h

%el

5.59 x

1010

<1.3

450

processed i n t h i s r)ur was cooled 1 9 months.

bFuel processed i n t h i s C Fuel processed i n t h i s %el processed i n t h i s e103Ru (tlI2= 4 1 days)

run was cooled 175 days.

run was cooled 80 W s . run w m cooled 25 days. and 141Ce (tlI2= 32 days) were not detected by gamma scan.

‘I

>4.3 x 1010

lo3 107

lo6 ~ 3 . 2 6x l o 3

>1.7 x

lo7

<. .*

-.<.

3:36 x io1O

lllk

125%

l o l o <1.9

2.34 x

6.14 x 1O1O

<6100

io9 io9

3.4 x

105

<520

‘8.52 x 103

510’

106

9.9 x 1 0 5 5.10 x 103

6.4

105

7.0 x i o 6 1.09 x 104

2.0

105

>1.7 x 10’

<7.9

>2.9 x 10’

x 1011 d.1 x 105 a . 2 0 x 103

<g.o

>6.8 x

x 1010

lo9

>1.1x 104

3.8 x

106

~13.6

>6.5 x 106

2.85 x

io6

4.59 x 103

>g.6 x 107 1.4 x 107

t h e f a c t t h a t lo3Ru w a s not detected lead t o t h e conclusion t h a t lo6Ru m a y have been acquired from incompletely decontaminated (from t h e pre-

vious campaign) process piping. Tellurium and iodine decontamination y although not high, w a s s a t i s f a c t o r y ; the p r i n c i p a l concern with these radionuclides i s release t o the atmosphere, which i s discussed i n Sect.

The DF f o r cerium, representing t h e rare e a r t h s , w a s consistently

high. In run

RA-4, t h e decontamination factors were somewhat lower than

i n t h e first t h r e e runs, generally ranging from 2 x lo5 t o lo8. The low antimony DF m a y have been caused by the t r a n s f e r of a l a r g e amount of antimony i n t o t h e fluorinator as a r e s u l t of the shorter hydrofluorination period employed i n

RA-4. The low "Mo

DF w a s not surprising

since the chemical properties of MoF6 are quite s i m i l a r t o those of UF6. The "molybdenum stripping" feature of t h e process was intended t o keep t h e bulk of t h e s t a b l e molybdenum from contaminating the UF6 product, but w a s not expected t o e f f e c t t h e complete removal required f o r radiochemical decontamination.

The "Mo

(tl,2 = 2.8 days) would not be pre-

sent i n fuel processed a f t e r a cooling period of 60 days; it would decay out of t h e product ( t o ~0.1%)30 days after processing.

Such a cooling

period might be desirable t o allow f o r decay of t h e unavoidable 237U. The "'Ag

i s believed t o be of no concern since none of the scrubber

solutions showed any appreciable concentrations of i t ; a low y i e l d and a high l i m i t of detection combined t o give an indeterminate DF f o r t h i s

nuclide. Generally speaking, the DF's obtained i n t h e f i r s t t h r e e runs were about one order of magnitude lower than those achieved i n t h e e a r l i e r zirconium program.'

W e believe t h a t t h i s i s due, at least i n p a r t , t o

fluorinating at 55OoC r a t h e r than at 500째C.

The l o w e r DF's i n run

m a y be due t o one o r more of t h e following factors.

RA-4

First, if the

sintered-metal f i l t e r located j u s t upstream of t h e product cylinder had f a i l e d , NaF f i n e s could have carried f i s s i o n products i n t o t h e UF6 cylinder.

Second, i n run

RA-4, t h e additional HF sparge carried out

a f t e r dissolution w a s complete w a s shortened; therefore, t h e molten

40 salt may have carried a higher percentage of the f i s s i o n products i n t o the fluorinator.

Finally (and t h i s factor seems most s i g n i f i c a n t ) , the

quantity of NaF p e l l e t s discharged from the movable-bed absorber may have been i n s u f f i c i e n t .

In t h e aluminum program, 3.8 kg of NaF (about

11% of the bed) was discharged during each run, as compared w i t h about

6.5 kg i n the zirconium program.

Insufficient bed discharge would

r e s u l t i n an upward migration and eventual breakthrough of some f i s s i o n products.

6.2 Primary Radioactive Contaminants i n the Product from Run RA-4 The gamma s c a n , f o r t h e product samples i n run

RA-4 ( t h e short-

cooled run) indicated t h a t the radioactivity of t h e product was almost e n t i r e l y due t o t h e presence of "Mo

and 237U.

The 237U content of t h e

product w a s equivalent t o a radioactivity l e v e l of 2.53 x 10l1 disintegrations per minute

per gram of uF6 at a cooling time of approximately

39 d a y s . The gross b e t a a c t i v i t y of t h e product i n run

RA-4

w a s 2.7 x 10l1

disintegrations per minute per gram of u F 6 , assuming a 10% geometry during analysis.

This value i s lo2 t o lo3 times t h e gross beta a c t i v i t y

of the product i n run RA-3 (3.66 x lo8 disintegrations per minute per gram of UF~). c

Radiation readings f o r the product cylinders i n runs RA-3 and

RA-4,

taken w i t h a c u t i e p i e at various times a f t e r the run w a s completed, were as follows:

R u n No.

RA-3 RA-4 RA-4 RA-4 RA-4

Approximate Cooling Time (days

4 5 28 (completion of run) 36 39 49

Cutie Pie Reading (r/hr)

0-5 50

6.2 4.2 2

h-,

The decrease i n t h e readings of t h e RA-4 product r e f l e c t s t h e decay

of both 99M0 and 237U.

6.3

Removal of Technetium, Neptunium, and Plutonium from t h e uF6 Product

The removal of technetium, neptunium, and plutonium from t h e uF6

product i s discussed separately i n t h i s section because of t h e incompleteness of t h e data.

This, i n t u r n , i s due t o the d i f f i c u l t y of

obtaining v a l i d analyses of highly radioactive material.

Most of t h e

d a t a f o r these t h r e e elements were discarded as meaningless; f o r example, one set of analyses indicated t h a t t h e 237Np content of the f u e l w a s 20 times as great as the t o t a l uranium content.

The recovery of plutonium i n t h e processing of highly enriched uranium f u e l s i s not economically f e a s i b l e .

Hence t h e only concern,

with respect t o plutonium, i n t h e runs discussed i n t h i s report i s assurance t h a t t h e plutonium content of t h e uF6 product i s within reasonable limits.

Past experience i n molten-salt v o l a t i l i t y processing

has shown excellent plutonium decontamination; t h e data from t h e aluminum program a r e given l a t e r i n t h i s section. I n general, technetium and neptunium f l u o r i d e s tend t o follow uF6 through t h e sorption-desorption cycle i n t h e NaF bed, although t h e data below suggest t h a t t h e cycle might be arranged t o achieve some decontamination.

Thus a separate p u r i f i c a t i o n s t e p , i n which TcF6 and NpFs

were sorbed on MgF2 p e l l e t s , w a s employed. Some technetium apparently remained i n t h e off-gas stream during sorption; f o r example, analysis of t h e NaF i n t h e chemical t r a p downstream from t h e sorber showed 3.6 ppm of technetium at t h e entrance and

0.5 ppm at about t h e middle of t h e t r a p a t t h e conclusion of runs RA-1, -2,

and -3.

However, most of the technetium w a s found on t h e MgF2.

I n l e t and o u t l e t samples of t h e sorbent from t h e MgF2 bed showed t h e following 99Tc concentrations:

RA-1,

-2, and -3 (combined), 1540 ppm

and not g r e a t e r than 0.15 ppm respectively; RA-4,

8.8 ppm and 1.7 ppm;

42 and T-18," 970 ppm and 2 ppm.

Results of t h e analyses of samples of

the composite bed f o r the e n t i r e s e r i e s indicated t h a t a t o t a l of 390 mg of 99Tc was present i n the 973-g bed.

three runs contained very l i t t l e 99Tc:

The products i n the first

4.6, 2.4, and 1.1 p a r t s of

technetium per million p a r t s of uranium respectively.

run RA-4, the product analyzed 440 ppm.

However, i n

This discrepancy has not, as

y e t , been explained. The r e s u l t s f o r neptunium were similar t o those f o r technetium. Some neptunium reached the chemical t r a p during runs RA-1,

-2, and -3;

the 237Np concentrations i n t h e NaF at the i n l e t and at the middle of

the t r a p were 23 ppm and 1 ppm respectively.

Analyses of samples of

MgF2 ( i n l e t and o u t l e t respectively) showed:

RA-1,

-2, and -3, 2310

ppm and 40 ppm; RA-4, 490 ppm and 150 ppm; and T-18,*

1140 ppm and 420

ppm. Analyses of samples of the composite bed indicate t h a t 785 mg of 237Np was present i n the bed. Analyses of the u F 6 products f o r runs RA-1 through

-4 showed no consistent trend:

140, l e s s than 12, 1670,

and 1860 p a r t s of 237Np per million p a r t s of uranium. Plutonium data are available from runs RA-3 and RA-4.

Data from

run RA-3 confirm the d i f f i c u l t y of fluorinating plutonium from a molten For the first fluorination i n run RA-3, the feed s a l t and t h e

salt.

waste salt showed e s s e n t i a l l y t h e same plutonium concentration (1.19 x

lo6

and 1.25 x

lo6

counts per minute per gram of s a l t , respectively);

the feed salt concentration w a s equivalent t o 1.2 x per gram of uranium.

lo9

counts per minute

For t h e second fluorination, the plutonium radio-

a c t i v i t y i n t h e feed salt w a s 1.26 x

lo6

counts per minute per gram of

salt (equivalent t o 1.6 x io9 counts per minute per gram of uranium),

but t h e analysis of t h e waste salt showed only 5.1 x lo5 counts per minutes per gram of s a l t .

Radiochemical analyses of four product (m6)

samples from run RA-4 f o r plutonium yielded t h e following results:

l o 4 , 5.1 x l o 5 , 1.7

3.1 x

gram of u F 6 .

lo4,

add 4.5 x

lo4

counts per minute per

If we discard t h e second value, t h e average i s 3.1 x

lo4

T-18 w a s a cleanup run (using barren s a l t ) made following run RA-4.

43 counts per minute per gram of UF6, o r 4.6 x 104 counts per minute per gram of uranium.

Based on a 51% counting geometry, 4.6 x lo4 i s equiva-

~ ~ of 0.63 ppm. l e n t t o a 2 3 9 concentration

6.4

Nonradioactive Cationic Contaminants

Contamination of t h e UF6 products (Table 6.3) with nonradioact5ve cations w a s higher than desirable.

Analyses of t h e products from indi-

vidual runs (even when multiple samples f r o m t h e product of a s i n g l e

run were used) were too inconsistent f o r v a l i d i n t e r p r e t a t i o n .

However,

r e s u l t s of t h e analysis of t h e composite product, consisting of a combination of these products in a form s u i t a b l e f o r shipment, appear quite reasonable.

The concentrations of most of t h e cationic contami-

nants, including C r , Cu, Fe, N i , and Z r , were l e s s than 50 p a r t s per million parts of uranium.

The boron and potassium contents were about 75

ppm and 500 ppm respectively.

Three other metals were present i n t h e

following (approximate) q u a n t i t i e s :

molybdenum, 0.1%; aluminum, 0.3%;

and sodium, 1%. The molybdenum content of t h e product w a s dependent on t h e conditions under which the movable-bed absorber w a s operated.

The

high sodium content l e d us t o suspect t h a t t h e sintered-metal f i l t e r located j u s t upstream of t h e product cylinder had f a i l e d , allowing NaF f i n e s t o be released.

The high aluminum content has not been completely

explained, but i s believed t o be nonrepresentative.

6.5

Cumulative Material Balances f o r S a l t and Uranium

A cumulative salt balance w a s maintained throughout each run i n

the aluminum-uranium campaign (Table 6.4).

This allowed us t o obtain

a uranium balance (uranium concentrations were reported as grams of uranium per gram of s a l t ) , and it a l s o provided a check on t h e i n s t r u mentation t h a t w a s used t o measure t h e l e v e l and density of molten salt. It was imperative t h a t t h e volume of s a l t i n t h e f l u o r i n a t o r be known

accurately i n order t o avoid o v e r f i l l i n g t h e waste salt can.

The

volume of waste salt t h a t w a s t r a n s f e r r e d w a s measured by difference i n t h e f l u o r i n a t o r , since t h e waste can contained no instrumentation.

Table 6.3.

Cation

UA-1

RA-1

Nonradioactive Cationic Impuritiesasb i n t h e w6 Products

RA-2 E

AVg.

RA-3

RA-4

200

1,900

1,600

1.700

1,300

1,500

1,300

2,850

3,100

3,500

560

4,600

8,100

5,800

4,100

5,125

3,700

9,100‘

lk,

210

300

1,540

420

300

400

357

UA-2

UA-3

1,200

800

900

1,500

570

2,000

13,350

7,800

6,300

4,900

4,400

1,000

ll.5

Composite

3.6

<O. 57

0.45

4.57

<50

<90

250

160

260

110

100

<30

0.50

105

130

<5O

200

300

230

1,300

200

300

502

378

600

454

565

660

300

900

1,000

270

770

120

600

l.200

740

700

7,000

880

,560

3,450

1,800

5,700

5,000

1,100

2,900

600

2.300

2,700

1,900

2,775

4,475

3,500

8,100

12,000

3.600

3.5

0.30

700

200

<30

3.3

<40

<TO

105

<20

18,880

13,280

‘9,500

9,670

13,500

16,690

Total

s62

V a l u e s a r e in p p .

b A l l d a t a except t h e values i n t h e l a s t column were taken from spectrographic analyses. analyses.

The d a t a i n t h e l a s t column were obtained from chemical

‘Sample appeared t o be contaminated.

ND = None detected.

C # .

C’

Table 6.4.

Run No.

Material Balances f o r t h e S a l t Charges Used i n VPP Runs

Initial Heel in

Barren Salt

AlF3 from

Initial Heel in

FV-1000 (kg)

Charge

Dissolution

(43)

(kg)

FV-1000 (kg)

Final Heel in

FV-100

After

Final Heel i n

(kg)

FV-1000 (kg)

DPla (kg)

FM-100 (kg)

32.1

N8F from FV-105

Total Input

(kg)

Waste salt (kg)

Cumulative Material

Recovery *

Balre

14.7

2.5

12.2

14.7

45.8

13.8

184.8

8.3

180.2

202.3

112.1

102.1

182.0

20.5

161.0

3.3

160.3

184.1

101.2

101.8

2.5

161.4

19.3

141.8

0.7

143.6

163.6

101.4

101.7

0.7

2.5

182.3

182.0

1.0

183.5

184.5

101.2

101.6

1.0

2.5

152 * 5

155.7

0.3

157.9

158.2

103.7

102.0

FVSb

32.1

DA-1

149.9

24.3

2.5

3.7

180.4

DA-2

13.8

132.7

24.6

8.3

2.6

UA-1

20.5

114.0

21.1

3.3

UA-2

19.3

139.2

20.6

UA-3

135-9

13.1

RA-1

190.3

27.4

0.3

3.0

221.0

48.9

172,6

28.6

147.0

224.5

101.6

101.9

RA-2c

48.9 130.3

190.3

27.0

1.9

294.8 148.3

130.3 6.5

171.1 136.0

16.1 4.5

155.0 133.4

301.4 144.4

102.2

28.6 16.1

97.4

102.1 101.8

6.5 73.5

209.6

25.4

73.5 15.4

163.3 99.2

100.9

1.0

102.2

236.8 118.6

96.3

4.0

246.0 116.1

124.7

4.5 38.6

38.6

102.2

101.1

15.4

114.3

13.1

1.0

4.2

148.0

9.6

136.3

1.0

139 5

150.1

101.4

101.1

1408.3

196.6

26.9

1631.8

9.6

1.0

1639.5

1650.1

RA-3c FA-4 Total

%PT = dissolver product transfer. bFVS = freeze valve salt charging. 'RU~

was divided i n t o two parts.

101.1

-r

46 Table 6.4 shows the salt inputs f o r each run, as w e l l as the t o t a l input.

I n i t i a l heels i n t h e dissolver (FV-1000) and the fluorinator

(FV-100) a r e the quantities that were indicated by instruments on the vessels; the quantities of barren salt and AlF3 were based on known weights; and t h e NaF t h a t was added t o FV-100 from t h e absorber (FV-105)

w a s measured by FV-100 instrumentation.

The recovery f o r each run i s

the t o t a l of t h e f i n a l heels (as measured by vessel instrumentation) and t h e waste salt (as measured by 'hoefore and a f t e r t r a n s f e r readings" on FV-100 instruments). The column labeled "FV-100 after DF'T" i s included t o indicate t h e degree of r e l i a b i l i t y of vessel instrumentation. value f o r each run

Theoretically, t h e

should be equal t o t h e wesght of t h e i n i t i a l salt

charge minus the sum of t h e weights of t h e NaF from FV-105 and t h e f i n a l heel i n FV-1000.

Ekperience has shown t h a t measurements of changes

i n vessel contents are reasonably r e l i a b l e when they a r e made using a single s e t of instruments.

The data i n Table 6.4 indicate discrepancies

between instruments. A very poor material balance was obtained f o r t h e freeze valve salt

charging (FVS).

Since only a s m a l l quantity of material w a s charged and

since instrument heels and l i n e volumes would account f o r some salt, a low material balance w a s expected.

However, t h e balance f o r t h e subse-

quent run compensated f o r the low balance; thus t h e weights of t h e f i n a l heel a f t e r FVS and the i n i t i a l heel f o r run DA-1 are probably incorrect. Each of the two runs RA-2 and RA-3 consisted of a complete dissolut i o n , a p a r t i a l t r a n s f e r of salt from FV-1000 t o FV-100, fluorination, t r a n s f e r of waste s a l t , t r a n s f e r of remaining salt from FV-1000 t o FV-100, fluorination, discard of NaF p e l l e t s from FV-105, and t r a n s f e r of waste

salt.

Hence, a salt balance i s shown f o r each fluorination.

An o v e r a l l material balance of 101.1% i s within t h e anticipated accuracy of the measurements invalved.

Therefore, the r e l i a b i l i t y of

t h e instrumentation w a s adequate f o r process control, and experience i n t h e first f e w runs generated a high degree of confidence i n t h e measurements

The uranium content of the salt w a s calculated from the s a l t balance data presented i n Table 6.4 and from analyses of salt samples. This information was used, along with weights and analyses of effluents

(including t h e product) and accountability data f o r the material charged, t o obtain t h e uranium material balances shown i n Table 6.5. RELEASE OF FISSION PRODUCTS TO THE ENVIRONMENT

Various quantities of f i s s i o n products were released through the 3020 stack t o the atmosphere during the fluorination step i n t h e hot

runs (see Table 7.1). Noble gases were released during dissolution, but they were not detected a t ground l e v e l . (Run RA-4)

, less

released.

than

In t h e l e a s t desirable case

1 m C i of lo6Ru and l e s s than 1 0 m C i of 1311 were

A t o t a l of about 20 C i of tellurium w a s released; of t h i s ,

6 C i w a s 78-hr 132Te, which decays t o 2.3-hr 1321and, i n turn, t o s t a b l e xenon.

Calculations t o determine the percentage of t h e maximum

permissible concentration (MPC) represented by these releases were made by assuming a constant r a t e of release during the fluorination period. The most unfavorable case was 1311 (<O.l% of MPC); however, even i f t h e e n t i r e quantity of 1311 had been released during a period as short as 1 min, the combined fraction of MPC f o r all three f i s s i o n products (129Te,

1311, and lo6Ru) would s t i l l have been below

5%. We conclude

t h a t f u e l t h a t has been cooled f o r as long as 25 days can be processed by t h e molten-salt fluoride-volatility process with no hazard r e l a t i v e t o the release of f i s s i o n products. The quantity of tellurium (20 C i ) t h a t was released during run

RA-4 represented only about 5% of the calculated quantity i n the f u e l element. (i.e.

The major f r a c t i o n of the tellurium that w a s v o l a t i l i z e d

, less

than one-half of the tellurium i n the f u e l ) was removed by

the KOH scrubber solution.

An attempt w a s made t o remove tellurium i n

a t r a p located downstream from t h e fluorine scrubber.

This t r a p con-

s i s t e d of nickel mesh heaked t o bOO째C, a gas cooler, and an activatedcharcoal. bed maintained near ambient temperature ( <5OoC).

Although

Table 6.5.

Uranium Balances for VPP Runs

Uranium Initial Total Uranium Found i n Holdup i n i n FV-100 and Hydrofluorinator System FV-100 Product Product Initial Uranium Total Final by S a l t Collection Collection

Product "Cold Trapped" i n Collection Cylinder

D i s t r i b u t i o n of Nonrecoverable Uranium Losses KOH Total Waste NeutralKOH in

Final Holdup i n Product Collection

UA-1

592

264

328

325 67.06

218

91.98

373

0.11

110

UA-2

264

304

568

239

329

110

439

440

67.24

296

91.31

1835

240

0.63

141

UA-3

239

304

543

391

152

141

293

300

67.33

202

91.40

253

171

0.28

RA-1

391

323

714

533

181

272

300

65.64

197

86.88

235

119

0.20

RA-2a

533

306

839

734

105

180

225

0.24

180

RA-2b

734

539

195

180

375

470

1.17

RA-3a

539

324

863

701

162

221

155

0.76

220

RA-3b

701

625

220

296

410 67.05

206

2.13

RA-4

625

159

784

600

184

203

loo

0.41

138

Cleanout

600

138

66.76

314

83.00

2234 1079

63.97

275

82.49

1410

82.52

746

6OTb

124'

% o t a l of 7 g unaccounted for (99.7% material balance). bCollected as

UO2F2 s o l u t i o n , e t c .

'Grams of uranium i n salt flushes.

This uranium could have been recovered by f l u o r i n a t i o n , but it was not considered worthwhile t o do so.

Table 7.1.

Fission Products Released t o the Atmosphere from the VFP ' During t h e Aluminum Campaign (Quantities expressed i n curies)

Fission Products Released from Dissolver

- _.._ ----

127m25

*r n LIUDiCe vaDicil

Present

Run No.

Detected

Present

Detected

Fission Products Released f r o m Fluorinator 129m1 3 12 7 10 3P.. 2c mu Present Detected Present Detected Present Detected

RA-1

47.3 (85Kr)

<0.01

0.007

<0.001

RA-2

60 (85Kr)

518

0.03

0.01

100

171

*O .09a

8.5

341

2Ob

RA- 3

0.5

748

~0.001

2766

~0.01

3986

106n.. I,U Present

Detected

167

445

W.003

491

so. 002

242

<O.0OlC

RA-4

24 (85Kr) 2261 (133Xe)

2293

........................................................................................................................................................... 0

WCf, v~i/cm3

%MF'Cg*h during RA-4

2.8

x 10-6

(see d) 10-3

10-9 (see e)

8 x 10-8

~ 6 . 5x

(see d)

(see

a . 0 x 10-2

BTot al tellurium. b127Te c103m

+ +

lZ9Te + 132Te; all isotopes assumed t o be l Z 9 T e i n t h e calculation t o determine t h e percent MF'C released. 106~~.

%orst case

f o r insoluble material..

eWorst case

f o r soluble material.

fSource: Maximum Permissible Body Burdens and Maximum Permissible Concentrations Of Radionuclides i n A i r and i n Water f o r Occupational Exposure, National Bureau of Standards Handbook 69, U.S. Department of Commerce, 1959. The MPC's given a r e based on a 40-hr week. g h determined at ground level. hBases bbr calculation: (1)r a t e of air flow through t h e stack w a s 43,000 cfm; (2) atmospheric d i l u t i o n of air discharged from t h e 3020 stack was lo5; ( 3 ) f i s s i o n products were released at a constant rate during the e n t i r e fluorination period (40 min). Quantities of f i s s i o n products were determined a n a l y t i c a l l y i n samples from charcoal t r a p s used t o sorb f i s s i o n products from a sidestream of stack gases.

appreciable amounts of tellurium were held by the t r a p , a s i g n i f i c a n t f r a c t i o n was allowed t o pass through t o t h e off-gas stack.

Some of t h e

trapped material w a s subsequently released a f t e r evolution from the fluorinator had ceased; thus, the t r a p helped extend the release period. It appears t h a t , although t h e quantity of tellurium t h a t w a s dispersed d i d not constitute a hazard, f u r t h e r developmental work i n t h i s area

would be advisable. A l l of the effluent streams i n run

Of the 2293 C i of 1 3 1 1 present a t the t i m e of dis-

t h e f a t e of iodine. solution, only 110

RA-4 were analyzed t o determine

w a s accounted f o r by analyses.

Since laboratory

experience has indicated t h a t there i s a low probability of finding t h e iodine i n a fused s a l t , we believe t h a t t h e remainder of t h e iodine was e i t h e r "plated out" i n the vessels or w a s discharged i n t h e waste salt. O f the 110 C i accounted f o r , 109 C i was found i n the caustic solution

used f o r scrubbing the dissolver off-gas and 0.8 C i w a s found i n t h e caustic used f o r scrubbing t h e fluorinator off-gas.

The quantities of

iodine found i n the other streams were i n s i g n i f i c a n t ; e s s e n t i a l l y none was released t o the atmosphere.

We conclude t h a t molten-salt fluoride-

v o l a t i l i t y processing of nuclear f u e l s presents no iodine release hazard, regardless of t h e length of t h e cooling period between irradiat i o n and processing.

ACKNOWLEDGMENTS

R. E. Brooksbank, Chief of the P i l o t Plant Section of the Chemical

Technology Division, had overall responsibility f o r t h i s p i l o t plant. R. P. Milford was reponsible f o r coordinating the V o l a t i l i t y Project.

Other supervisors of t h e V o l a t i l i t y P i l o t Plant who made major contributions were:

R. S. Lowrie, S. Mmn, R. J. Shannon, and E. L.

Youngblood. Many people i n other p a r t s of the Laboratory provided aid and assistance t h a t made the operation and maintenance of t h i s p i l o t plant a success; chief among these were: G. E. Pierce and R. P. Beard of t h e

Plant and Equipment Division, W. J. Greter of the Instrumentation and Controls Division, E. I. Wyatt and C. E. Lamb of t h e Analytical Chemistry Division, 0. J. Smith of t h e Inspection Engineering Department, and C. H. Miller and W. A. McLoud of the Health Physics and Safety Division.

The assistance of these, and of many other people at ORNL and a t other AEC i n s t a l l a t i o n s , i s gratefully acknowledged.

We a l s o take t h i s opportunity t o express appreciation t o Martha G. Stewart f o r her invaluable e d i t o r i a l assistance.

9. REFERENCES

Chem. Technol. Div.

Ann. Progr. Rept. June 30, 1962, ORNII-3314,

39-45; Chem. Technol. Div. Ann. Progr. Rept. May 31, 1964, ORNL3627, PP* 29-40.

pp.'

2. R. G. Nicol and W. T. McDuffee, Molten-Salt Fluoride Volatility Pilot Plant: Equipment Performance During Processing of Aluminum-Clad Fuel Elements, ORNL-4251 (July 1968). 3.

S. Mann and E. L. Youngblood, Decontamination of the ORNL MoltenSalt Fluoride Volatility Pilot Plant After Processing Irradiated Zirconium-Uranium Alloy Fuel, 0"L-3891 (February 1966).

4. E. L. Youngblood, R. P. Milford, R. G. Nicol, and J. B. Ruch, Corrosion of the Volatility Pilot Plant INOR-8Hydrofluorinator and Nickel 201 Fluorinator During Forty Fuel-ProcessinR Runs with Zirconium-Uranium Alloy, OFNL-3623 (March 1965).

R. E. Thoma, B. J. Sturm, and E. H. Guinn, Molten-Salt Solvents for Fluoride Volatility Processing of Aluminum-Matrix Nuclear Fuel Elements. ORNL-3594 ( A m s t 1964).

6. M. P. Lietzke and H. C. Claiborne, CRUNCH

A n IBM-704 Code for Calculating N Successive First-Order Reactions, 0RN1t-2958 (Oct.

24, 1960). 7.

P. D. Miller et al., Corrosion Resistance of Nickel-Base Alloys Under H y u , BMI-x-215 (Jan. 7, 1963).

8. P. D. Miller et al., Corrosion Resistance of Nickel-Base Alloys Under Hydrofluorinator Conditions with Aluminum Dissolving Part 11 . BMI-x-260 (act 18. 1963). - - a ~

Corrosion Investigation of HyMu 80, INOR-8, 9. P. D. Miller et al., ~~

and "L" Nickel Under Fluorinator Conditions for Processing Aluminum UsinR the Molten-Salt Fluoride Volatility Method, BMI-X-316 (Oct. 30, 1964).

10.

Chem. Technol. Div. Ann. Progr. Rept. May 31, 1966, ORNL-3945, p.

63.

10. APPENDIXES

54 10.1 Appendix A:

Fission Product Content

of I r r a d i a t e d Fuel Elements The r a d i o a c t i v i t y l e v e l s of t h e f i s s i o n products expected t o be

present i n t h e spent f u e l elements t h a t were processed i n t h e four hot runs of t h e uranium-aluminum campaign were calculated using t h e CRUNCH code.

6 The r e s u l t s a r e l i s t e d i n Table A-1.

55 Table A-1.

Calculated R a d i o a c t i v i t y Levels of F i s s i o n Products i n U-A1 Fuel Elements i n R U ~ SRA-1, -2,

-3, and

-4

start

of Processinga

Crilculated Radioactivit? Level i n Run at S t a r t o f Processinn (curies ) I

e5Kr

89~r 90~r 91Y 95

~ r

9 5Nb

RA-2

RA-3

RA-4

47.3

59.5

54.1

24.2

lo6Ru

0.09

0.03

0.19

7.77

10.48

8.56

28 Y

8.43

1.55

0.70

0.24

3131

7973

7928

0.38

10.80

13.84

10.76

31.1

3874

9189

8333

0.69

13.36

15.95

11.31

69.8

7342

14505

7523

1.56

25.32

25.18

10.21

148

2.8

0.5 167

0.06

0.02

0.06

748

2766

3986

445

491

242

9.3

14.7

1.7

6.5

14.1

51.4

100

33.8

171

0.01

8.5

13 3 ~ e

0.30

136cs

0.34 323

383 4.2

14OBa

l4l~e

761

7.6

3 ~ r

2338

7703 0.51

'47Nd

1081

1730

1.26

0.01

2.58

4.80

5.41

1.0 y

3.73

1.53

0.85

0.33 0.02

7.6 S2.7 Y 3.9

48.2

0.21

0.05

0.02

0.01

0.04 0.04

0.18

0.17

0.07

0.12

0.30

0.46

0.01

3.11

2293

8.05

232

3.2

0.32

5-27

3.07

0.01

2261

7.9

1.32

0.59

0.21

12.8

0.01

0.57

11.19

8559

2.62

7-05

11.62

442

8378

13.7

0.03

0.77

11.38

8964

4550

290

26.57

15.56

6.18

3126

11.3

1676

734

2.6 y

342

152

33 Y

331

8243

4059

149b

1.3

0.20

2.12 x 105

30.9

341

32Te

151h

0.21

17.0

1311

147%

1.06

179

129Te

144ce

10.27 y

403

127%

137cs

RA-4

450

377

0.25

127Te

RA-3

6306

111& 125%

RA-2

6036

0.03

3Ru

of T o t a l

RA-1

2252

8.5

9~ M O

9Tc

RA-1

1.28

7.7

2.25

0.73

73 Y

7.22

52.28

24.17

5.97

0.13

4.24

2.91

1.00 0.01

0.03

%e first n u c l i d e i n each decay chain c o n t r i b u t i n g a s i g n i f i c a n t f r a c t i o n of t h e t o t a l a c t i v i t y at t h e start of processing i s t a b u l a t e d . A t r u e t o t a l a c t i v i t y would a l s o have t o include t h e a c t i v i t y of t h e s h o r t - l i v e d daughters t h a t are i n s e c u l a r equilibrium with t h e n u c l i d e s l i s t e d .

bA blank i n d i c a t e s <0.01 c u r i e o r <0.01%. 'The h a l f - l i f e is expressed i n days unless followed by y , which designates y e a r s .

10.2

Appendix B:

Dissclution of Fuel Elements

The aluminum dissolution rates are plotted against run times in Figs. B-1 through B-10.

The HF flow rate, the average temperature,

and the salt compositions for each run are shown.

57 ORNL Dwg 69-7701

Fig. B-1.

Aluminum Dissolution Rate as a Function of Fuel Element

Dissolution Time for Run DA-1. O R N L Dwg 70-12432

Fig. B-2.

Aluminum Dissolution Rate as a Function of Fuel Element

Dissolution Time for Run DA-2.

ORNL Dwg 69-7702

DISSOLUTION TIME (hr)

Fig. B-3.

Aluminum Dissolution Rate as a Function of Fuel Element

Dissolution Time for Run UA-1.

Fig. B-4.

Aluminum Dissolution Rate as a Function of Fuel Element

Dissolution Time for Run UA-2.

59 ORNL Dwg 69-7703

O.? e L

20.6 1

w t-0.5 -

5 0.4 F

& 0.3 v)

L!!

0.2

5z 5

0.1

3 J

O O

DISSOLUTION TIME (hr)

Fig. B-5.

Aluminum Dissolution Rate as a Function of Fuel Element

Dissolution Time for Run UA-3. O R N L Dwg 70-12431

DISSOLUTION TIME (hr)

Fig. B-6.

Aluminum Dissolution Rate as a Function of Fuel Element

Dissolution Time for Run RA-1.

60 ORNL Dwg 69-7704

DISSOLUTION TIME (hr)

Fig. B-7.

Aluminum Dissolution Rate as a Function of Fuel Element

Dissolution Time for Run RA-2.

O R N L Dwg 70-12434

DISSOLUTION TIME (hr)

Fig. B-8.

Aluminum Dissolution Rate as a Function of Fuel Element

Dissolution Time for Run RA-3.

61 O R N L Dwg 69-7705

DISSOLUTION TIME ( h r )

Fig. B-9.

Aluminum Dissolution Rate as a Function of Fuel Element

Dissolution Time for Run RA-4. ORNL Dwg 70-12440 0.7 n

o, 0.6

CONDITIONS SALT (mole O/& INITIAL FINA

I ALl MlNUM DISSOLU’ION RA-E

3s?L

5 0.5 a

HFFLOW

KF 6 1,3 ZrFa 23.0 2t.O AlFa 15.5 22.9 Avg. Temp. 5 8 5 O C

I40

(20 n

too ‘E

RATE

80 w t-

-I- 0.4~

a a

$ 0.3

n v)

0.2

END OF DISSOLUTION-

0.1

3 J

fiw

0 -

- 0

10.3 Appendix C:

Decontamination of

P i l o t Plant Equipment Decontamination methods similar t o those described f o r decontaminating t h e plant a f t e r t h e zirconium program w a s completed 3 were used t o reduce radiation l e v e l s i n excess of 5000 r / h r t o l e v e l s s u f f i c i e n t l y low t o permit t h e plant t o be dismantled.

Radiation dosage t o indivi-

duals who did the mechanical work d i d not exceed t h e quarterly 1.3-r allowance (see data i n Table E-1).

F i n a l backgrounds ( p r i o r t o equip-

ment r e m o d . ) were 1 t o 5 r / h r i n t h e majority of c e l l 1 locations; a maximum reading of 60 r / h r was obtained adjacent t o t h e HF i n l e t l i n e ,

t o the dissolver.

l e s s than 1 r / h r .

Backgrounds i n other areas of the plant were generally

The decontamination sequence included flushing w i t h a molten s a l t ,

followed by treatment with t h r e e types of aqueous solutions f o r removing salt f i l m , metal s c a l e , and deposits of radioactive nuclides.

The

compositions of t h e t h r e e solutions were, respectively: Ammonium oxalate, 0.3 t o 0.35 M; o r Al(NO3)3*HN03, 0.1 t o 0.01 Aluminum n i t r a t e , 0.1 -

Sodium hydroxide--hydrogen 5-1-1 wt

peroxide--sodium

t a r t r a t e , about

This procedure i s s u i t a b l e f o r use w i t h r a d i a t i o n l e v e l s a t least

as high as those encountered i n the operating period c i t e d , and it does

not r e s u l t i n excessive personnel exposure during t h e decontamination and subsequent equipment removal.

10.4

Appendix D:

Corrosion of Vessels

Corrosion s t u d i e s i n t h e V o l a t i l i t y programs w e r e d i r e c t e d p r i i

marily at t h e dissolver (hydrofluorinator) and t h e f l u o r i n a t o r f o r t h r e e reasons.

F i r s t , these two vessels were located i n a congested

no-access" mea t h a t w a s accessible only after extensive decontamina-

t i o n , whereas many of t h e other vessels were located i n limited-access

areas.

Second, these vessels, which were a p a r t of t h e head-end por-

t i o n of t h e p l a n t , were e s s e n t i a l t o t h e operation of the p l a n t , and t h e i r replacement would require a major shutdown ( i n addition t o t h e necessary decontamination).

Third, corrosion observed on t h e other

vessels and components w a s s l i g h t , probably because t h e conditions under which they were operated ( e s p e c i a l l y temperature and contact with corrosive chemicals ) were less d r a s t i c . Earlier Studies a t BMI.

Studies made at BMI ( B a t t e l l e Memorial

I n s t i t u t e ) of t h e dissolver under run conditions indicated a metal

loss of 0.2 mil/month f o r INOR-8during aluminum processing as compared with 2.5 t o 5 mils/month during zirconium p r o ~ e s s i n g . ~No i n t e r g r a n u l a r a t t a c k w a s observed.

L a t e r s t u d i e s indicated an

INOR-8

corrosion rate of approximately 10 mils/month during aluminum processing because of intergranular attack.8

The t o t a l t i m e of HF exposure during

t h e aluminum runs w a s 250 h r . Studies at BMI indicated a corrosion rate of approximately 100 mils/month f o r t h e "L" n i c k e l f l u o r i n a t o r during aluminum a l l o y processing as compared with approximately 30 mils/month during zirconium processing.

The higher rate w a s a t least p a r t i a l l y a t t r i b u t e d t o t h e

higher operating temperature

[6oo0c r a t h e r than 5OO0C (ref. g ) ] t h a t

w a s used i n t h e aluminum processing. Corrosion of t h e Dissolver.

The t o t a l corrosion of t h e dissolver

during t h e 10 aluminum runs w a s 5 mils, as previously reported. lo This value w a s obtained by using pulse-echo and Vidigage techniques. 1

I ,

corrosion d a t a for the dissolver are reported elsewhere.

Fluorinator.

( a ) Vessel.

- Corrosion

data f o r t h e f l u o r i n a t o r

during zirconium processing a r e reported elsewhere.

(However, t h e

data f o r t h e l a s t 11 zirconium runs were not reported i n t h a t reference

because the fluorinator w a s not examined between t h e end of t h e z i r conium program and t h e start of the aluminum program.)

Corrosion

r e s u l t s obtained on examination of t h e f l u o r i n a t o r after t h e plant w a s dismantled following t h e aluminum program a r e reported i n Table D-1.

The values l i s t e d i n Table D-1 a r e f o r 50 runs (40 zirconium

runs and 1 0 aluminum runs) and all aqueous decontamination sequences. The region most vulnerable t o attack w a s t h e salt region, as noted p r e v i ~ u s l y . ~The maximum corrosion w a s l e s s than 3/4 mil per run; t h e Visual examination of t h e in-

average was less than 1 / 2 mil per run.

side of t h e f l u o r i n a t o r revealed no evidence of excessive attack. Comparison of the l o s s e s i n t o t a l w a l l thickness (Table D-1) with data obtained i n t h e e a r l i e r corrosion study

4 revealed

t h a t metal l o s s e s of

approximately 10 m i l s apparently occurred i n t h e bottom of t h e v e s s e l (but not i n t h e t o p ) during t h e aluminum runs. (b)

Corrosion Rods. -The

corrosion rods t h a t were placed i n

t h e f l u o r i n a t o r i n l a t e 1962 were removed when t h e plant was dismantled

following t h e aluminum program.

Results of v i s u a l inspection of t h e

rods a r e summarized i n Table D-2.

The rods were not examined metal-

lographically because of t h e termination of t h e V o l a t i l i t y program i n mid-1967. Other Vessels and Components. the d i s s o l v e r ) .

- Visual

( a ) HFV-2207-1 (HF i n l e t l i n e t o

inspection of t h e 3-ft length of INOR-8l i n e ,

including t h e elbow, showed only minor scratches.

No evidence of

leakage was noted when t h i s section of pipe was pressurized t o 35 psig. This portion of the l i n e w a s examined because it had f a i l e d during

e a r l y v o l a t i l i t y processing.

A t t h a t time, however, t h e l i n e was

made of Inconel instead of INOR-8. (b)

Fluorine Supply Tanks.

- Inspection

of t h e i n s i d e of

tank NB-1432 on June 7, 1962, revealed no corrosion; no leakage occurred

Table D-1.

Bulk Metal Losses from t h e Nickel 201 Fluorinator During F i f t y

and Associated

Aqueous Decontaminations i n t h e VPP i

Exposure times:

Section of t h e Fluorinator M e asurede

90.7 h r of fluorineCsd; 2962 h r of molten salt d

Total Wall Thickness LOSS ( m i l s ) MU. Avg

Corrosion R a t e M i l s per month M i l s per hour of Molten S a l t of F2 Exposure Exposure Max. Avg MU. Avg

Top 16-in. -dim section

3.5

0.12

0 A39

2-7

0.85

Top cone

8.4

0.14

0.093

3.2

2.0

Neck

18.6

0.26

0.205

5.8

4.51

Bottom 16-in. dim s e c t i o n

19.2

0.30

0.212

6.6

4.66

Bottom cone

21.0

0.32

0.232

7.0

5.10

%arty

zirconium runs and t e n aluminum runs.

bSee ORNL-3623 (ref. 41, Table 12. C

Fluorine exposure t i m e does not include v e s s e l exposure during desorption. dvalues given include those presented i n Table 1 2 , 0 ~ ~ ~ 3 6(ref. 2 3 4). A breakdown of exposure t i m e s i s as follows: Fp Exposure Time (hr)

(hr)

29 (value from 0 ~ ~ ~ 3 6 2 3

57.6

1922

11 ( f i n a l runs i n Z r program)

18.1

604

10 (runs i n Al program)

15.0

436

50 ( t o t a l runs>

Molten S a l t Exposure Time

No. of Runs

2962

e Measurements were made at 1-in. i n t e r v a l s i n t h e south quadrant of t h e vessel.

Table D-2.

Condition of Corrosion Rods on Removal from Fluorinator After t h e Aluminum Runs

Material

Results of Visual Operation

INOR-8a

Brown film; some corrosion i n salt region.

"L" Nickela

Brown film; some corrosion at middle and bottom.

HyMu-80

Thin brown f i l m over full length; some loose m a t e r i a l on t o p 4 i n . ; uniform diameter.

Specimen lb3'

Brown f i l m ; s l i g h t l o s s i n width a t bottom; warped.

Specimen 2

Same as f o r specimen 1.

Ni-Mg

Brown f i l m ; corrosion s i m i l a r from t o p t o bottom.

aI n s t a l l e d on November 13, 1962. b I n s t a l l e d on December 11, 1962. C

Weld t e s t u n i t s f a b r i c a t e d from "L" n i c k e l and INOR-8, using t h e following weld materials : I~CO-82 INOR-8 1nco-61 and "L" nickel.

b 5

during a 75-psig pneumatic t e s t .

The 1/2-in.

gage o u t l e t on t h e rear

head appeared t o be c a v i t a t i n g i n t h e heated-affected zone of t h e weld. The f i l l e d weld around t h i s o u t l e t 9n t h e tank e x t e r i o r w a s thought t o i

be adequate t o t a k e care of t h e condition on a temporary b a s i s .

All

s t o p valves i n t h e manifold s e c t i o n were checked and reworked i f necessary.

"he e x t e r n a l surfaces were found t o be corroding; painting

of t h e s e surfaces was recommended. Inspection of t h e i n s i d e of tank NB-1433 on July 28, 1966, revealed no corrosion; no leakage w a s observed when t h e tank w a s pressurized t o

75 psig.

I n l e t and o u t l e t stubs t h a t had been welded t o t h e r e a r head

showed no corrosion.

A l l s t o p valves i n t h e manifold section were

checked and overhauled.

External surfaces of t h e u n i t were s a t i s f a c t o r y ;

t h e u n i t w a s repainted. (c)

Other Vessels. - N o v i s i b l e corrosion w a s detected visu-

a l l y i n (1)t h e movable-bed absorber (FV-1051,

( 2 ) t h e flash cooler

(FV-1001), (3) t h e HF condenser (FV-2001), o r (4) t h e f u e l element

charging chute on t h e d i s s o l v e r (FV-1002).

Ir,

68 10.5

Appendix E:

Radiation Safety

Penetrating r a d i a t i o n from m a t e r i a l s being processed required t h a t each piece of equipment containing t h e m a t e r i a l be heavily shielded t o prevent exposure of operating personnel.

Access t o t h e process equip-

ment w a s c a r e f u l l y controlled at all times during normal operation and while maintenance o r decontamination operations were i n progress.

The

physical form of t h e m a t e r i a l involved (dust p a r t i c l e s , l i q u i d s , o r gases) governed t h e type of p r o t e c t i v e Clothing and r e s p i r a t o r y equipment t h a t was used i n t h e s e shielded areas. The average r a t e and t h e maximum r a t e of personnel exposure t o r a d i a t i o n i n t h e aluminum program were about t h e same as those encountered i n t h e zirconium program.

I n each case, exposure rates were highest

during p l a n t decontamination procedures because temporary, unshielded piping w a s connected t o "dead end" process l i n e s i n normal work areas f o r recycle of solutions.

Even under t h e s e conditions, however, t h e

m a x i m u m exposure t o any individual did not exceed 75% of t h e recommended 3 maximum permissible dose t o body organs f o r a calendar quarter. Unusual occurrences* were less frequent i n t h e aluminum program than i n t h e zirconium program, l a r g e l y because experience had been gained i n making design and operational changes.

Because we a n t i c i p a t e d

higher r a d i a t i o n l e v e l s from t h e shorter-cooled elements t h a t were t o be processed i n t h e aluminum campaign, l e a d shielding w a s i n s t a l l e d i n many areas t h a t formerly did not require i t ; c o r r e c t i v e actions were taken as problems developed. Personnel Exposure. -Radiation marily t o t h e presence of 12'Sb,

exposure t o personnel w a s due p r i -

9 5 Z r , and "Nb

i n t h e d i s s o l v e r off-gas

An unusual occurrence i s defined as one which may r e s u l t i n ( a ) personnel exposure i n excess of t h e maximum permissible, ( b ) cleanup c o s t s o r property l o s s i n excess of $5000, ( c ) an i n c i d e n t of publicr e l a t i o n s s i g n i f i c a n c e , o r ( d ) exposure of t h e o f f - s i t e population t o r a d i a t i o n i n excess of t h e maximum permissible.

system and t o the presence of 237U, c

99Tc, lo3Ru, lo6Ru, 99M0, and 237Np

i n t h e f l u o r i n a t o r off-gas and product c o l l e c t i o n systems. The average exposure r a t e remained fairly constant during the pro-

cessing of dummy and i r r a d i a t e d f u e l elements, but increased by about a f a c t o r of 2 during decontamination (Table El).

During the processing of i r r a d i a t e d fuel elements, t h e maximum dose received by an individual during a single day was 80 mrads. sure occurred at t h e conclusion of run

This expo-

RA-4 when an operator removed t h e

product receiver from t h e receiving s t a t i o n i n c e l l 2 and hauled it t o t h e sampling room i n a lead-lined drum.

The product was highly radio-

a c t i v e because of t h e unusually high 237U content.

The "cutie pie"

reading of t h e unshielded product cylinder w a s 25 r / h r at a distance of

1.5 i n . During t h e decontamination period, t h e maximum dose received by an individual i n a single day w a s 150 mrads.

This exposure occurred during

replacement of a damaged rubber gasket on t h e waste s a l t nozzle sealing jack.

Removal of t h e gasket was slow because the nut t h a t held t h e end

flange of t h e drain l i n e t i g h t l y against t h e gasket was located on t h e underside of the gasket holder and w a s not e a s i l y accessible. I n summary, the maximum r a d i a t i o n exposure received i n the aluminum campaign d i d not exceed about 75% of t h e ORNL l i m i t of 1.3 rem/quarter, o r 5 rem/year.

The average VPP personnel r a d i a t i o n exposure w a s about

50% of t h e maximum. Unusual Occurrences.

- N o unusual occurrences

were experienced

during t h e processing of dummy o r i r r a d i a t e d f u e l elements i n t h e aluminum program.

However, two occurred a t other times: one during prepara-

t i o n f o r s t a r t u p , and one during v e s s e l decontamination. J u s t before t h e first dummy f u e l element was processed, some previously used air-operated valves were dismantled and decontaminated i n a "hot" sink preparatory t o r e p a i r . i

and the n o s t r i l s of the operator i n charge were contaminated by a loose powder found i n s i d e t h e valves.

During t h i s procedure, t h e h a i r

However, subsequent investigation

Table E-1.

Radiation Exposure of Personnel During t h e

Processing of Aluminum-Clad Fuel Elements

Type of Run o r Operationa

Max. /Day

Exposure ( m a d s ) Max. /Quarter

Max. /Week

A v g . /Quarter

450

200

120

500

130

150

505

240

150

190

940

440

De cont amination

UA RA

- dummy elements containing aluminum only. - dummy aluminum elements "spiked" with u n i r r a d i a t e d UF4.

LITR and ORR f u e l elements cooled 18 months t o 25 days.

showed t h a t t h e operator had received no i n t e r n a l exposure and only negligible external exposure.

Analysis of t h e incident indicated t h a t

a v e n t i l a t e d hood w a s needed over t h e sink and t h a t a mask should be worn when contaminated equipment was being opened. After t h e aluminum s e r i e s had been completed, a radiochemical s p i l l (of material t h a t had not completely drained from a pipe) occurred during an attempt t o remove a valve bellows assembly from an inactive HF charging system pipe l i n e south of Building 3019.

The bellows as-

sembly w a s needed t o replace one t h a t had f a i l e d i n t h e combination c a u s t i c sampling and temporary decontamination solution recycle system i n c e l l 2.

When t h e valve w a s opened s l i g h t l y ( a s t e p i n t h e valve dis-

mantling procedure), some radioactive l i q u i d i n t h e pipe flowed past the operator, as signaled by a personal r a d i a t i o n monitor.

This l i q u i d

continued along t h e pipe t o a previously dismantled valve and overflowed t o t h e ground.

The blacktop area under t h e dismantled valve w a s decon-

taminated by flushing with water, chipping, and vacuum cleaning.

Inves-

t i g a t i o n showed t h a t t h e t h r e e people i n t h e v i c i n i t y at t h e time of t h e incident received no i n t e r n a l exposure and only negligible external exposure. Control of Exposure. -Radiation

exposure to personnel was controlled

by taking appropriate action following thorough, frequent checking of t h e work areas during t h e runs.

High radiation backgrounds were reduced

by shielding vessels and pipes w i t h lead p l a t e , by discarding contaminated solutions, and by backflushing f i l t e r s .

The number of c e l l e n t r i e s by

operating personnel was s a f e l y reduced by using revised operating procedures.

The amount of radioactive material t h a t escaped t o t h e atmosphere

w a s decreased by decreasing t h e purge rates and t h e duration of t h e purges. A 35-point radiation check of t h e area w a s made at l e a s t once per

8-hr s h i f t during each run t o determine any chmges i n the background.

Data a r e summarized f o r runs RA-1 through -4 i n Table E-2. Some of t h e measures taken t o reduce the r a d i a t i o n exposure are i

discussed below.

For example, the i n s t a l l a t i o n of 1/2-in.-thick

lead

shielding reduced t h e background r a d i a t i o n a t t h e c a u s t i c sampler i n

Radiation Backgroundsa in the Various VPP Work Areas During Runs RA-1.

Table E-2.

RA-1 Before After Run Run

Location

RA-2 Before RUI

-2, -3, and -4

RA-3 After Run

Before Run

RA-4

Max.

After Run

Before

Run

MU.

After Run

Cell 2. HF System FV-700-1C.

Bp f i l t e r

120

200

Caustic sampling station

Suction l i n e of FV-4201. caustic pump

35 22

65 65

35 410

ECV-1003-1,

EF catch tank drain

valve

100

25 28

8 5 5

58 15

45 32

2 120

6 49

9 6 5 40

105

9 13

6 2

18 3

15 3

5 3

25 8

35 6

6 34 4

24 3

W-4202. HF pump, remote head FV-l207. EF vaporizer

300 60

160 42

85 40 60 42

50 25

48 28

Cell 2. UFg System Heated duct a t entry from c e l l 1 W-120-A. MgFz bed

FV-723. product stream f i l t e r FV-220, cold trap FV-121-A. chemical t r a p

43 8 5 5 ll

6 5 13

200 MO

100

300 95

120 65.O0Ob 35

330 27

96 15

60 66 200

35 22 82

Penthouse. HF system

FV-7500. off-gas liquid trap Off-gas l i n e from W-1009. caustic neu-

tralizer

FV-9509, f l m e a m s t o r

Dissolver vent l i n e a t f i l t e r W-7501

2 1

Transmitter rack (back)

Penthouse Fz Systan

W-153. W-150. W-150. W-152. W-450.

liquid t r a p in off-gas l i n e

32 205 25 6

caustic scrubber. top caustic scrubber. bottom caustic surge tank caustic pump

155 16

Tellurium trap. W-154. nickel wool Tellurium trap, W-155. Off-Gas

charcoal

22 140 120

600

180

1,000

160

800

85 100

5.500 1.800 600

170 400

95 100

440

25 6

19 6

23 20

43 23 20 22

ll0 15 8

730 440

Scrubber System

FV-164, first stage of scrubber W-164. middle of scrubber FV-164. scrubber entrainment separator FV-765. f i l t e r pv-165. surge tank

%dues

(LIY

17 4

4 1 6 2

5 2

6 2

2 6 2

given in mr/b.

bAiter nitrogen-sparge of s a t in fluorinator (prior t o fluorination).

22 64

73 c e l l 2 t o 20% of t h e previous value.

The same thickness of lead re-

duced t h e background at t h e NaF t r a p (FV-121) t o about 25% of t h a t before i n s t a l l a t i o n .

This same thickness a l s o reduced t h e background

f o r t h e f l u o r i n a t o r off-gas scrubber (FV-150) t o 10-25% of the unshielded background.

Approximately 1 i n . of l e a d t h a t was placed over the i n l e t

l i n e f o r the c a u s t i c c i r c u l a t i n g pump (FV-4201) i n c e l l 2 reduced t h e background t o 6% of t h a t observed e a r l i e r . The dumping of f l u o r i n a t o r off-gas scrubber recycle c a u s t i c solut i o n reduced t h e background f o r t h e storage tank (FT-152) t o about 10% of t h a t p r i o r t o dumping.

A t the same time, the background f o r t h e

c a u s t i c c i r c u l a t i n g pump (FV-450) w a s reduced by about 40 t o 50%. Backflushing of t h e dissolver off-gas f i l t e r (FV-7001C), using waste l i q u i d â&#x201A;ŹIF, reduced i t s background t o about 25% of t h a t noted previously.

74 10.6

Appendix F:

Index of V o l a t i l i t y P i l o t Plant Log Books

The log books l i s t e d below, along with t h e run sheets and t h e recorder c h a r t s , comprise t h e primary record of t h e VPP operations described i n t h i s report.

The run sheets and t h e recorder charts w i l l

be destroyed two months a f t e r t h i s report i s issued.

The l o g books

w i l l be retained permanently a t O a k Ridge National Laboratory. ~

VPP Log

Laboratory Records Notebook

No.

Inclusive Dates

Subject Matter

1 t o 15

7/11/56-10/7/59

ARE Program

16 to 43

10/19/59-9/11/63

U-Zr Alloy Program

A-2964

9/12/63-1/22/64

Preparation f o r U-Al Program

A-2226

1/22/64-3/26/64

Preparation f o r U-Al Program

A-3389

A-

A-3422

3390

Run DA-1 (aluminum, no uranium)

5/1/64-6/2/64

DA-1

6/2/64-6/26/64

DA-2 and UA-1 (uranium and aluminum, nonirradiated )

A-3423

6/26/64-8/4/64

UA-1 and UA-2

A-3474

8/4/64-9/10/64

UA-3 and RA-1 ( i r r a d i a t e d )

A-3475

9/ll/64-10/19/64

RA-1 and RA-2

A-3476

10/19/64-ll/16/64

RA-2, RA-3, and RA-4

A-3477

11/16/64-12/14/64

RA-3 and RA-4 (shortcooled)

A-3478

Cleanout and Shutdown (cont. )

No.

Laboratory Records Notebook No.

A-3711

2/2/65-3/23/65

Cleanout and Shutdown

A-3714

3/24/65-5/27/65

Complete VPP Log Index

VPP Log i

Inclusive Dates

Subject Matter

A-6105

Sample Log

A-6106

Sample Log

ORNL-4 574

- Chemical

UC-10

INTERNAL DISTRIBUTION 1. Biology Library 2-4. Central Research Library 5. ORNL - Y-12 Technical Library Document Reference Section 6-25. Laboratory Records Department 26. Laboratory Records , OFiNL R. C 27 E. D. Arnold 28. J. E. Bigelow 29 R. E. Blanco 30. J. 0. Blomeke 31. R. E. Brooksbank 32. K. B. Brown 33. F. N. Browder 34. W. D. Burch 35 G. I. Cathers 36. W. L. Carter 37 J. M. Chandler 38. C. F. Coleman 39 D. J . Crouse 40. F. L. Culler 41. F. L. Daley 42. D. E. Ferguson 43. L. M. F e r r i s 44. E. J. Frederick 45 H. E. Goeller 46. A. T. Gresky 47 M. B. Herskovitz 48. R . W. Horton 49 L. J. King 5 O F. G. K i t t s 51 B. B. K l i m a 52 C. E. Lamb 53 R. E. Leuze 54 B. Lieberman 55 R. B. Lindauer 56 R. S. Lowrie 57 H. G. MacPherson

Separations Processes f o r Plutonium and U r a n i u m

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